ML20129J538

Safety Evaluation Supporting Amends 116 & 101 to Licenses NPF-11 & NPF-17,respectively

ML20129J538
Person / Time
Site:	LaSalle
Issue date:	10/29/1996
From:	NRC (Affiliation Not Assigned)
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ML20129J536	List: ML20129J534 ML20129J538
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NUDOCS 9611060316
Download: ML20129J538 (10)

Text

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UNITED STATES g

j NUCLEAR REGULATORY COMMISSION 2

WAF;41NGTON, D.C. 20 6 0001

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION i

l RELATED TO AMENDMENT NO.116 TO FACILITY OPERATING LICENSE NO. NPF-11 AND l

AMENDMENT NO.101 TO FACILITY OPERATING LICENSE NO. NPF-18 f

COPMONWEALTH EDISON COMPANY LASALLE COUNTY STATION. UNITS 1 AND 2 DOCKET N05. 50-373 AND 50-374

1.0 INTRODUCTION

i By letter dated April 8,1996, as supplemented by letter dated October 14, 1996, Commonwealth Edison Company (Comed, the licensee) proposed changes to

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the Technical Specifications to reflect the transition from General Electric (GE) fuel to Siemens Power Corporation (SPC) fuel. LaSalle County Station, i

Units 1 and 2 currently operate with GE fuel and methodologies. Beginning with Unit 2 Cycle 8 and Unit 1 Cycle 9, SPC fuel will be loaded with co-l resident GE fuel. The ATRIUM-9B fuel design and SPC methodologies used to analyze fuel performance during normal and abnormal operating conditions were approved by the NRC. Technical Specification requirements related to the fuel thermal limits are affected as a result of the change in fuel design and methodologies. The current TS contain terminology and methodologies which are specific to GE fuel.

Because both GE and SPC fuel will be present in the core, the licensee proposes to remove all vendor specific references from the TS requirements. The proposed TS will include discussions of both GE and SPC methodologies in the TS Bases. Other changes, unrelated to the transition to SPC fuel, were also proposed as line-item improvements from the Improved Standard Technical Specifications (NUREG-1434). These include relocation of the traversing in-core probe TS to the Core Operating Limits Reports (COLR) and revision of the fuel description in section 5.0.

The October 14, 1996, submittal provided additional clarifying information that did not change the initial proposed no significant hazards consideration determination.

2.0 EVALUATION 2.1 Linear Heat Generation Rate (LHGR)

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 proposes to delete the definitions of FLPD and MFLPD from the Definitions section of the TS.

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. i FLPD and MFLPD are GE terms. SPC uses Fuel Design Limiting Ratio to monitor LHGR. The proposed change makes the TS vendor independent by deleting the definitions for FLPD and MFLPD and placing them in the COLR. The definition of LHGR will include a vendor independent statement that LHGR is monitored by the ratio of LHGR to its fuel specific limit, as specified in the COLR. There is no change to the limiting condition for operation (LCO) for LHGR which currently raquires that LHGR be maintained less than or equal to the LHGR limit specified in the COLR. These changes are acceptable. The proposed addition to the definition of LHGR was inadvertently not included in the i

proposed Unit 1 TS pages submitted by the licensee. This change was included in the proposed Unit 2 TS pages. The licensee has stated that the change is applicable to both units.

2.2 Critical Power Ratio (CPR)

LaSalle currently uses the GE critical power correlation called GEXL to calculate the CPR for the GE fuel bundles. The licensee proposes to revise the definition of CPR in section 1.9 of the TS by deleting reference to the GEXL correlation which is GE specific and replacing it with the term " approved CPR correlation". GEXL will no longer be used once SPC fuel is loaded.

Instead, the SPC critical power correlation (ANFB) will be the correlation of record. The change to a non vendor-specific term does not change any TS requirements and is acceptable.

TS 3.2.3, Minimum Critical Power Ratio, requires that the MCPR be equal to or greater than the MCPR limit specified in the COLR. The MCPR operating limit specified in the COLR may be scram time dependent. TS 4.2.3 specifies scram times to be used to determine the applicable MCPR limit. The current TS requires that an average scram time of 0.86 seconds be used prior to performance of the initial scram time measurements or an average scram time may be used within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the conclusion of each scram time surveillance.

The current TS is based on GE methodology which uses a 20% scram insertion point as the reference for scram time surveillance. The TS maximum average scram time from the 20% scram insertion point is 0.86 seconds. The licensee proposes to revise the required scram times used to determine MCPR operating limits to be consistent with SPC methodology.

SPC methods use the 5%, 20%,

50%, and 90% scram insertion points to determine the MCPR operating limit.

The licensee proposes to revise TS 4.2.3 to delete the requirement to use 0.86 seconds prior to performance of the initial scram time measurements and instead use Technical Specification Scram Speeds (TSSS). TS 3.1.3.3 specifies the required average scram insertion times for the 5%, 20%, 50%, and 90% scram insertion points. The use of four scram times provides additional conservatism and is acceptable. In addition, rather than using the average scram insertion times from the surveillances, the proposed TS use the nominal scram speed (NSS) insertion times specified in the COLR if the surveillance results meet the NSS insertion times. SPC methods for analyzing the scram time dependence of the MCPR operating limit utilize cycle-specific nominal values. The process for determining the MCPR operating limits has been replaced with the SPC methods, including use of the NSS which is maintained in the COLR. The proposed change appropriately reflects the NRC-approved SPC 1

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. methodology and does not change the current requirement that MCPR meet the limits specified in the COLR. Therefore, the proposed change is acceptable.

i Current TS 6.6.A.6.a.2 requires that core operating limits be established and documented in the COLR for MCPR before each reload cycle. The licensee proposes to delete the requirement that MCPR limits be based on a 20% scram insertion time and instead require that scram time dependent MCPR limits be established. As discussed above, the GE methodology which bases MCPR i

operating limits on the 20% scram insertion time is being replaced with SPC methodology which uses four scram insertion points. The change to TS 6.6.A.6.a.2 reflects this change and is acceptable.

The licensee also proposes to add to TS 6.6.A.6.a.2 the requirement to include the effects of m alyzed equipment out of service in the COLR. The re.quirement is being added to TS 6.6.A.6.a.2 because SPC methodology performs tne out of service analyses on a cycle-specific basis, the results of which are documented in the COLR before each reload. The proposed change accurately reflects the revised SPC methodology and is acceptable. The details of these analyses had previously been relocated from the bases section of the TS to the COLR by letter dated May 23, 1995.

Current TS 3.4.1.1 " Recirculation Loops" Action a.l.b requires that, with one recirculation loop in operation, within four hours, the MCPR safety limit must be increased by 0.01 to 1.08 per TS 2.1.2.

The licensee proposes to delete the specific requirement to increase MCPR to 1.08.

TS 2.1.2 requires that MCPR shall not be less than 1.08 with single recirculation loop operation.

t Therefore, the inclusion of the MCPR limit in TS 3.4.1.1 is repetitious and unnecessary and its deletion is acceptable.

2.3 Averaae Planar Linear Heat Generation Rate (APLHGR)

TS 3.2.1, Average Planar Linear Heat Generation Rate, requires that the APLHGR not exceed the limits specified in the COLR. This specification assures that the peak cladding temperature following the postulated design basis LOCA will not exceed the limit specified in 10 CFR 50.46. Current TS 3.4.1.1, Recirculation Loops, action a, requires certain actions to be taken when only one reactor coolant system recirculation loop is in operation. The licensee proposes to add action a.l.e to require that the APLHGR limit be reduced by the applicable single loop operation factor specified in the COLR. This requirement is specific to SPC fuel. The GE methodology does not require a specific single loop operation (SLO) factor to be applied because GE methodology, as documented in the COLR, uses multipliers based on core flow (SLO results in a reduction of 30 - 40% in total core flow) which have the same effect on APLHGR limits as the use of a SLO factor. The proposed requirement conservatively ensures that the peak cladding temperature for SLO is bound by the peak cladding temperature for two-loop operation and is acceptable.

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2.4 Traversina In-core Probe (TIP) i l

TS 3.3.7.7 requires that the traversing in-core probe (TIP) system be operable for the purpose of calibrating LPRM detectors. The TIP system allows calibration of LPRM signals by correlating TIP signals to LPRM signals as the TIP is positioned in various radial and axial locations in the core. The current TS allows the use of substituted TIP data from symmetric channels if the control rod pattern is symmetric. SPC methods use a statistical check of TIP symmetry, which is an assumed parameter in their analysis methods.

Therefore, this method needs to be incorporated into the LaSalle Core Operating Limits Report (COLR). The licensee proposes to relocate the requirements of TS 3/4.3.7.7 from the TS to the COLR.

Section 182a of the Atomic Energy Act (the "Act") requires applicants for nuclear power plant operating licenses to state TS to be included as part of i.he 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 conditions 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" (" Final Policy Statement"), 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 " technical specifications are to be reserved for those matters as to which the imposition of rigid conditions or limitations upon reactor operation is deemed necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety."

Consistent with this approach, the Final Policy Statement identified four criteria to be used in determining whether a particular matter is required to be included in a Limiting Condition for Operation (LCO), as follows:

(1)

Installed instrumentation that is used to detect, and indicate in the 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 transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; (3) a 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 LCO 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 do not fall within or satisfy these criteria may be relocated to other, licensee-controlled documents. The Commission amended 10 CFR 50.36 to codify and incorporate these four criteria (60 FR 36953).

The four criteria are discussed below:

Criteria 1:

Installed instrumentation used to detect, and indicath in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

The TIP is used as a calibration tool for the LPRMs.

It is not used for detecting and indicating degradation of the primary pressure boundary.

Any leakage of the TIP tubing in the reactor pressure boundary would be indicated in the control room similar to any other primary leak (e.g.,

through increased drywell pressure or increased sump flow rates).

Criteria 2:

A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The TIP system is used only as a calibration tool for the LPRMs and is not a condition of a Design Basis Accident or Transient analysis.

Its function as a calibration tool for the LPRMs results in uncertainties which are included in the core monitoring methods.

Criteria 3: A structure, system, or component that is part of the 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.

The TIP system's direct accident / transient function is the containment isolation function of the TIPS when they are penetrating primary containment. However, this system function is not related to the calibration function covered by the subject TS. The calibration function of the TIP which is required by the current TS is not a portion of the primary success path of safety sequence analysis.

Criteria 4: A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

The licensee did not identify any probabilistic risk assessment concerns i

with the TIP system. The TIP system calibrates the LPRMs which are not l

safety related and are not required to actuate safety systems, i

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The TIP operability requirements and action statements will be relocated to the COLR which is required to be submitted to the NRC per TS 6.6.A.6.

Any changes in the design or procedures for the TIP system will be evaluated for unreviewed safety questions per 10 CFR 50.59.

Because this change meets the four criteria for relocating LCOs from the TS and changes to the relocated requirements will be adequately controlled, the relocation of TS 3/4.3.7.7 to the COLR is acceptable.

2.5 Reactivity Anomaly i

i TS 3.1.2 requires that the reactivity equivalence of the difference between the actual rod density and the predicted rod density be less than or equal to 1 percent delta k/k. This limit ensures that plant operation is maintained within the assumptions of the safety analyses. The licensee proposes to replace t% words " rod density" with " critical control rod configuration" in l

the LCO and surveillance requirement. The current methodology utilizes a correlation between a change in rod density and core reactivity. The licensee is modifying its methodology to use a core monitoring code to monitor predicted Keff vs. actual Keff. This. monitoring system uses critical control rod configuration as an input. The new methodology is a more direct measurement method and provides a more accurate estimate of the anomaly than the current method. Because the proposed change to the TS only revises the method of measuring the difference between predicted and monitored core reactivity and does not change the required limit, the change is acceptable.

2.6 Fuel Bundle Description TS 5.3.1, Fuel Assemblies, provides a description of the fuel assemblies. The licensee proposes to expand this description consistent with Improved Standard Technical Specifications, NUREG-1434, and to better reflect the ATRIUM-9B design.

The revised description includes discussion of the use of water rods i

or water boxes which is consistent with the SPC fuel design. The proposed TS includes the statement that a limited number of lead test assemblies that have not completed representative testing may be place in non-limiting core regions, consistent with NUREG-1434. The proposed change accurately describes the SPC fuel design, is consistent with NUREG-1434, and does not affect any 4

l current TS requirements. Therefore, the proposed change is acceptable.

The proposed TS also states that limited substitutions of Zircalloy or ZIRLO or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. This change is consistent with NUREG-1434, and incorporates the guidance of Generic Letter 90-02, Supplement 1, " Alternative Requirements for Fuel Assemblies in the Design Features Section of Technical Specifications." The GL provided flexibility in the repair of fuel assemblies containing damaged and leaking fuel rods by reconstituting the assemblies. The number and location of filler rod substitutions are limited to configurations for which applicable NRC 1

approved codes and methods are valid and that have been shown by test or analyses to comply with all fuel safety design bases. To satisfy generic fuel

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design criteria as described in the Standard Review Plan, the filler rods

4 require thermal-hydraulic, neutronic, and mechanical analyses to demonstrate that inclusion of the filler rods in fuel assemblies with the specific configurations and core locations chosen for a specific fuel cycle is acceptable with respect to overall fuel performance and safety considerations.

l The proposed TS change complies with the gu* dance in GL 90-02, Supplement 1, will permit safe core configurations, and is acceptable.

l 2.7 Miscellaneous Chanaes i

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l Changes are made to the index pages to reflect deleted definitions and i

l deletion of the TIP TS as discussed above.

Numerous changes are made to the TS Bases to reflect the changes discussed above.

In addition, a typographical error is corrected on page B 2-9 to agree i

with the UFSAR. The current bases page states that the IRM system terminates l

a low power control rod withdrawal error transient at 1% rated thermal power.

The correct power level is 21%. The basis for the correction is the analysis in section 15.4.1.2 of the Updated Final Safety Analysis Report. Therefore, l

this change is acceptable.

l The licensee proposes to revise TS 6.6.A.6.b to include references to the i

following topical reports which are used to determine the core operating limits.

(1)

ANFB Critical Power Correlation, ANF-1125(P)(A) and Supplements 1 i

and 2, Advanced Nuclear Fuels Corporation, April 1990.

(2)

Letter, Ashok C. Thadani (NRQ to R.A. Copeland (SPC), " Acceptance for Referencing of ULTRAFLOW Spacer on 9x9-IX/X BWR Fuel l

Design," July 28, 1993.

(3)

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, XN-NF-524(P)(A) Revision 2, and Supplement 1 Revision 2, Supplement 2, Advanced Nuclear Fuels Corporation, November 1990.

(4)

COTRANSA 2: A Computer Program for Boiling Water Reactor Transient Analysis, ANF-913(P)(A), Volume 1, Revision 1 and Volume 1 Supplements 2, 3 and 4, Advanced Nuclear Fuels Corporation, August 1990.

l (5)

HUXY: A Generalized Multirod Heatup Code with 10 CFR 50, Appendix K Heatup Option, ANF-CC-33(P)(A), Supplement 1 Revision 1; and Supplement 2, Advanced Nuclear Fuels Corporation, August 1986 and January 1991, respectively.

(6)

Advanced Nuclear Fuel Methodology for Boiling Water Reactors, XN-NF-80-19(P)(A), Volume 1, Supplement 3, Supplement 3 Appendix F,

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and Supplement 4, Advanced Nuclear Fuels Corporation, November 1990.

i (7)

Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads, XN-NF-80-19(P)(A), Volume j

4, Revision 1, Exxon Nuclear Company, June 1986.

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(8)

Exxon Nuclear Methodology for Boiling Water Reactors THERMEX:

Thermal Limits Methodology Summary Description, XN-NF-80-19(P)(A),

i Volume 3, Revision 2, Exxon Nuclear Company, January 1987.

J (9)

Generic Mechanical Derign for Exxon Nuclear Jet Pump BWR Reload Fuel, XN-NF-85-67(P)(A) Revision 1, Exxon Nuclear Company, September 1986.

(10) Advanced Nuclear Fuels Corporation Generic Mechanical Design for i

Advanced Nuclear Fuels Corporation 9x9-IX and 9x9-9X BWR Reload Fuel, ANF-89-014(P)(A), Revision 1 and Supplements 1 and 2, October 1991.

l (11)

Volume 1 - STAIF - A Computer Program for BWR Stability Analysis in the Frequency Domain, Volume 2 - STAIF - A Computer Program for BWR Stability Analysis in the Frequency Domain, Code Qualification i

Report, EMF-CC-074(P)(A), Siemens Power Corporation, July 1994.

(12)

RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, XN-NF-81-58(P)(A), Revision 2 Supplements 1 and 2, Exxon Nuclear Company, March 1984.

(13) XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic i

Core Analysis, XN-NF-84-105(P)(A), Volume 1 and Volume 1 Supplements 1 and 2; Volume 1 Supplement 4, Advanced Nuclear Fuels Corporation, February 1987 and June 1988, respectively.

(14) Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model, ANF-91-048(P)(A), Advanced i

Nuclear Fuels Corporation, January 1993.

(15)

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, Richland, WA 99352, March 1983.

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F o (16) 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.

(17) Generic Mechanical Design Criteria for BWR Fuel Designs, ANF 98(P)(A), Revision 1 and Revision 1 Supplement 1, Advanced Nuclear Fuels Corporation, May 1995.

(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.

These additional topical reports are those used in SPC methodology and have been approved by the NRC for use at LaSalle. 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, and assist safe operation of the facility.

3.0 STATE CONSULT,$ HON In accordance with the Commission's regulations, the Illinois State official l

was notified of the proposed issuance cf the amendments. The State official had no comments.

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 effluants that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.

The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (61 FR 25699). The amendment also revises reporting requirements or record keeping requirements. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and c(10).

Pursuant to l

10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in ccnnection with the issuance of the amendments.

5.0 CONCLUSION

The Commission has concluded based on the consideration: discussed above, that:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations,

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and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor:

D. Skay Date: October 29, 1996 j

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