Text

Columbia - License Amendment Request for Changes to Technical Specifications Involving Core Operating Limits Report and Scram Time Testing
	ML082250680
Person / Time
Site:	Columbia
Issue date:	07/16/2008
From:	Gambhir S K; Energy Northwest
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	GO2-08-108
	Download: ML082250680 (52)
	v • d • e

Sudesh K. Gambhir E E G Vice President, Technical Services EN R YP.O. Box 968, PE04) NORTHW EST Richland, WA 99352-0968 Ph. 509.377.8313 I F. 509.377.2354 sgambhir@energy-northwest.com July 16, 2008 G02-08-108 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001

Subject:

COLUMBIA GENERATING STATION, DOCKET NO. 50-397 LICENSE AMENDMENT REQUEST FOR CHANGES TO TECHNICAL SPECIFICATIONS INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING

References:

1) BWR/4 Standard Technical Specifications (STS), NUREG-1 433, Revision 3.0 2) TSTF-222-A, Revision 1, Control Rod Scram Time Testing

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Energy Northwest hereby requests an amendment to the Columbia Generating Station (CGS) Operating License (NPF-21).

The proposed changes modify Technical Specifications (TS) 3.1.4, "Control Rod Scram Times," 3.2.2,"Minimum Critical Power Ratio (MCPR)," and 5.6.3, "Core Operating Limits Report (COLR)." The proposed changes are requested to support the transition to Global Nuclear Fuels -Americas (GNF) GEl 4 fuel during refueling outage R1 9 scheduled for the spring of 2009. The changes involve incorporating the analytical methodologies associated with operation of GNF fuel into the licensing basis. Specific changes for TS 3.1.4 include: 1) revising the Limiting Condition for Operations (LCO) to reflect GNF methodology consistent with Reference 1;2) changing scram time values as supported by GNF analysis; and 3) updating the surveillance frequencies as recommended by Reference 2.The change proposed for TS 3.2.2 includes the addition of a surveillance that reflects the GNF approach to calculating and monitoring this fuel thermal limit consistent with Reference

1. The changes proposed for TS 5.6.3 add the appropriate references to GNF analytical methods that will be used to determine core operating limits. The GEXL97 topical report to be added to the references requires Nuclear Regulatory Commission (NRC) review and approval and is provided in Attachment 4.The enclosure provides a description of the proposed change and the regulatory basis for the change. Attachment 1 provides the affected TS pages marked up to show the LICENSE AMENDMENT REQUEST FOR CHANGES TO TECHNICAL SPECIFICATIONS INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Page 2 proposed changes. Attachment 2 provides the affected TS pages in typed format to show the proposed changes. Attachment 3 provides the proposed TS Bases changes for information only. Upon approval of the requested amendment, these TS Bases changes will be implemented concurrently with the TS change in accordance with the CGS TS Bases Control Program. Attachment 4, NEDC-33419P, "GEXL97 Correlation Applicable to ATRIUM-10 Fuel,." Proprietary Version is provided for NRC review.Included in Attachment 4 is the GNF Affidavit for Request to Withhold Information.

Attachment 5 includes the Non-Proprietary Version, NEDO-33419, suitable for external release.Energy Northwest requests approval of the proposed amendment by May 2, 2009 to support loading of GNF fuel in R19. Once approved, the amendment shall be implemented prior to Cycle 20 operation.

There are no new regulatory commitments being made with this submittal.

Should you have any questions or require additional information regarding this matter, please contact Mr. MC Humphreys, Licensing Supervisor, at (509) 377-4025.I declare under penalty of perjury that the foregoing is true and correct. Executed on the date of this letter.pectfully, SK Gambhir Vice President, Technical Services

Enclosure:

Evaluation of the proposed changes Attachments:

1. Proposed Technical Specifications Changes (mark-up)2. Proposed Technical Specifications Changes (retyped)3. Proposed Technical Specifications Bases Changes (mark-up)4. NEDC-33419P, "GEXL97 Correlation Applicable to ATRIUM-1 0 Fuel," Proprietary Version 5. NEDO-33419, "GEXL97 Correlation Applicable to ATRIUM-10 Fuel," Non-Proprietary Version cc: EE Collins, Jr. -NRC RIV CF Lyon -NRC NRR NRC Senior Resident Inspector/988C RN Sherman -BPA/1 399 WA Horin -Winston & Strawn JO Luce -EFSEC RR Cowley -WDOH LICENSE AMENDMENT REQUEST FOR CHANGES TO TECHNICAL SPECIFICATIONS INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Enclosure Page 1 of 15 Evaluation of Proposed Changes

1.0 DESCRIPTION

This letter is a request to amend Operating License NPF-21 for Columbia Generating Station.The proposed changes modify Technical Specifications (TS) 3.1.4, "Control Rod Scram Times," 3.2.2, "Minimum Critical Power Ratio (MCPR)," and 5.6.3, "Core Operating Limits Report (COLR)." The proposed changes are requested to support the transition to Global Nuclear Fuels -Americas (GNF) GE14 fuel during refueling outage R19 scheduled for the spring of 2009. The changes involve incorporating the analytical methodologies associated with operation of GNF fuel into the licensing basis. Specific changes for TS 3.1.4 include adopting the approach of the Boiling Water Reactor (BWR) BWR/4 Standard Technical Specifications (STS), NUREG-1433, Revision 3.0 by: 1. revising the LCO to reflect GNF methodology;

2. changing scram time values as supported by GNF analysis; and 3. updating the surveillance frequencies as recommended by Technical Specification Task Force (TSTF) traveler TSTF-222-A, Control Rod Scram Times.The change proposed for TS 3.2.2 includes the addition of a surveillance that reflects the GNF approach to calculating and monitoring this fuel thermal limit consistent with the STS. The changes proposed for TS 5.6.3 add the appropriate references to GNF analytical methods that will be used to determine core operating limits.

2.0 PROPOSED CHANGE

2.1 TS 3.1.4 changes 2.1.1 Changes to LCO 3.1.4 and Table 3.1.4-1 Control Rod Scram Times Energy Northwest proposes adopting the licensing basis for GNF methodology, which is consistent with the STS via the following:

1) Simplify the LCO and associated CONDITIONS and REQUIRED ACTIONS to match that of the STS. These changes consist specifically of:

LICENSE AMENDMENT REQUEST FOR CHANGES TO TECHNICAL SPECIFICATIONS INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Enclosure Page 2 of 15-replacing current LCO 3.1.4 statement discussing average scram times in two-by-two arrays with two requirements, as follows: a. No more than 13 OPERABLE control rods shall be "slow," in accordance with Table 3.1.4-1, and b. No more than 2 OPERABLE control rods that are "slow" shall occupy adjacent locations.

-replacing ACTIONS section with the following:

CONDITION REQUIRED -COMPLETION ACTION TIME A. Requirements of the A.1 Be in Mode 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months LCO not met 2) Change the NOTE above Table 3.1.4-1, Control Rod Scram Times to:-add Note 1 -OPERABLE control rods with scram times not within the limits of this Table are considered "slow," and-designate existing information as Note 2.3) Change scram time limits to reflect the GNF analysis supported BWR/5 Scram Time versus Notch Position values.4) Correct a typographical error in note (a) to change "as" to "at".The associated Technical Specification Bases will be revised to reflect the above described changes.2.1.2 Revise Frequency of SR 3.1.4.1 and SR 3.1.4.4 (TSTF-222-A)

Energy Northwest proposes to incorporate the changes specified by TSTF-222-A, Rev. 1 which modifies STS, NUREG-1433, to clarify the frequency of performing control rod scram time testing subsequent to performance of an outage that involved the movement of fuel. The current wording of SR 3.1.4.1 could be interpreted that all control rods need to be scram time tested even if the shutdown was for a brief amount of time and only a limited amount of fuel was moved in the reactor (e.g., if only one bundle is moved in a mid-cycle fuel replacement).

This change clarifies the intent of the TS.It is proposed to revise CGS TS Section 3.1.4 to remove the surveillance test requirement to scram time test all control rods after each refueling outage. Only those control rods that reside in core cells that were affected by the refueling outage will need to be scram time tested after the refueling outage prior to reaching 40% rated thermal power (RTP). To affect this change, the frequency statements in SR 3.1.4.1 and SR 3.1.4.4 will be revised.

LICENSE AMENDMENT REQUEST FOR CHANGES TO TECHNICAL SPECIFICATIONS INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Enclosure Page 3 of 15 The associated TS Bases will also be revised to reflect these changes. The Bases will be clarified, in accordance with the approved TSTF, to discuss that control rods located in core cells in which one or more fuel bundles were moved will be scram time tested. Also, a clarifying statement will be added that explains that for normal refueling outages, all control rods would likely be affected and require testing.2.2 Addition of new surveillance requirement SR 3.2.2.2 An additional surveillance requirement is proposed with SR 3.2.2.2 to require MCPR operating limits to be determined subsequent to scram time testing required by SRs 3.1.4.1, 3.1.4.2, and 3.1.4.4. This surveillance will ensure that the specific scram speed distribution remains consistent with the GNF transient analysis, referred to as "Option B methodology" in Reference 7.1, that credits the conservatism in the actual scram speed performance.

This additional surveillance requirement is consistent with the proposed change in licensing basis to GNF methodology and is reflected in the STS.The Technical Specification Bases will also be revised to reflect the additional surveillance requirement for determination of MCPR operating limits.2.3 Changes to COLR TS 5.6.3.b lists the analytical methods used by CGS to determine core operating limits.With the transition to GNF methodology the following references will be added: 20. NEDC-33419P, "GEXL97 Correlation Applicable to ATRIUM-10 Fuel," Global Nuclear Fuel 21. NEDE-2401 1-P-A and NEDE-2401 1-P-A-US, "General Electric Standard Application for Reactor Fuel (GESTAR II) and Supplement for United States," Global Nuclear Fuel

3.0 BACKGROUND

3.1 TS 3.1.4 changes The scram function of the Control Rod Drive (CRD) System inserts negative reactivity during abnormal operational transients to ensure specified acceptable fuel design limits are not exceeded.

The control rods are scrammed using hydraulic pressure exerted on the CRD piston.The Design Basis Accident (DBA) and transient analyses assume that all of the control rods scram at a specified insertion rate. The resulting negative scram reactivity forms the basis for the determination of plant thermal limits (for example, MCPR). The current licensing basis at CGS allows for other distributions of scram times (e.g. several control rods scramming slower than the average time, with several rods scramming faster than LICENSE AMENDMENT REQUEST FOR CHANGES TO TECHNICAL SPECIFICATIONS INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Enclosure Page 4 of 15 the average time) to also be credited for providing sufficient scram reactivity.

Surveillance of each individual control rod's scram time averaged into associated two-by-two arrays ensures the scram reactivity assumed in the DBA and transient analyses can be met.The proposed change in licensing basis is to adopt the simplified STS approach that allows for a distribution of scram insertion times (some slower and some faster than the average time) based on individual control rod scram times instead of two-by-two arrays.The updated licensing basis will continue to require surveillances of each individual control rod scram time to ensure the scram reactivity assumed in the DBA and transient analyses can be met.3.1.1 Changes to LCO 3.1.4 and Table 3.1.4-1 Control Rod Scram Times Energy Northwest converted to the improved STS in 1997. Energy Northwest deviated from the STS for LCO 3.1.4 and retained much of the previous licensing basis as that was the approach supported by the methodology of the fuel vendor under contract at the time of the conversion.

The licensing basis retained was that of an "average" scram time based on two-by-two arrays, and listed multiple REQUIRED ACTIONS to characterize the number and spacing of allowed "slow" control rods. The BWR/4 STS, NUREG-1433, follows the GNF methodology and lists in the STS BASES the references that provide the foundation for the proposed changes (References 7.1 and 7.2).The following is an excerpt from the BWR/4 STS BASES for LCO 3.1.4: The scram times specified in Table 3.1.4-1 ensure that the scram reactivity assumed in the DBA and transient analysis is met. To account for single failure and "slow" scramming control rods, the scram times specified in Table 3.1.4-1 are faster than those assumed in the design basis analysis.

The scram times have a margin that allows up to approximately 7% of the control rods to have scram times exceeding the specified limits (i.e. "slow" control rods) assuming a single stuck control rod and an additional control rod failing to scram per the single failure criteria.Table 3.1.4-1 is modified by two Notes which state that control rods with scram times not within the limits of the table are considered "slow" and that control rods with scram times greater than 7 seconds are considered inoperable as required by SR 3.1.3.4.Updating the TS to match the STS for this LCO will ensure that the licensing basis properly reflects the GNF analytical methodology and ensure that specified acceptable fuel design limits are not exceeded during abnormal operational transients.

LICENSE AMENDMENT REQUEST FOR CHANGES TO TECHNICAL SPECIFICATIONS INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Enclosure Page 5 of 15 3.1.2 Revise Frequency of SR 3.1.4.1 and SR 3.1.4.4 (TSTF-222-A, Rev. 1)SR 3.1.4.1 requires that the scram time of each control rod be verified to be within the limits of Table 3.1.4-1 with reactor steam dome pressure 800 psig prior to exceeding 40% RTP after each refueling and after each reactor shutdown of greater than or equal to 120 days. Maximum scram insertion times occur at a reactor steam dome pressure of approximately 800 psig because of the competing effects of reactor steam dome pressure and stored accumulator energy. Therefore, demonstration of adequate scram times at reactor steam dome pressure _> 800 psig ensures that the measured scram times will be within the specified limits at higher pressures.

Verifying proper scram times before exceeding 40% RTP ensures that scram time testing is performed within a reasonable time after a shutdown duration of greater than or equal to 120 days.SR 3.1.4.4 requires that the scram time of each control rod be verified to be within the limits of Table 3.1.4-1 with reactor steam dome pressure 800 psig prior to exceeding 40% RTP after fuel movement within the affected core cell and work on control rods or the CRD system that could affect scram time. This testing ensures that, prior to withdrawing a control rod for continued operation, the control rod scram performance is acceptable for operating reactor pressure conditions.

When only fuel movement occurs, then only those control rods associated with the core cells affected by the fuel movement are required to be scram time tested.As precedence, the following Safety Evaluation was approved by the NRC for TSTF-222:-NRC Safety Evaluation for Peach Bottom, 5/10/2006, Technical Specification Amendments 259 and 262.3.2 Add new surveillance SR 3.2.2.2 The operating limit MCPR is established to ensure that no fuel damage results during anticipated operational occurrences (AOOs). The GNF transient analysis that calculates the operating limit MCPR credits the conservatism in scram speed performance by utilizing an interpolation of realistic scram times vice the allowed times specified in LCO 3.1.4, "Control Rod Scram Times." In order to credit the realistic scram times the transient analysis must assume an effective scram speed distribution, which may change slightly during the fuel cycle. The assumptions of the transient analysis can be validated by monitoring the actual scram speed distribution compared with the assumed distribution, a value designated as T. This surveillance requirement will monitor the expected small changes in T and ensure that the specific scram speed distribution remains consistent with that used in the transient analysis.

This analytical approach has been reviewed and approved by the NRC in Reference 7.1 and is consistent with the BWR/4 STS (NUREG-1433, Rev. 3.0).The BASES will be updated to reflect this new surveillance requirement.

LICENSE AMENDMENT REQUEST FOR CHANGES TO TECHNICAL SPECIFICATIONS INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Enclosure Page 6 of 15 3.3 Changes to COLR Core operating limits are established each operating cycle in accordance with TS 3.2,"Power Distribution," TS 3.3.1.3, "Oscillation Power Range Monitor (OPRM)Instrumentation" and TS 5.6.3, "Core Operating Limits Report (COLR)." These operating limits ensure that the specified acceptable fuel design limits are not exceeded during any conditions of normal operation or in the event of any AOO.The format of the references in TS 5.6.3.b follows the BWR/4 STS (NUREG-1433, Revision 3.0) 5.6.5.b which states:[Identify the Topical Report(s) by number and title or identify the staff Safety Evaluation Report for a plant specific methodology by NRC letter and date. The COLR will contain the complete identification for each of the Technical Specification referenced topical reports used.to prepare the COLR (i.e., report number, title, revision, date, and any supplements).]

The methods used to determine the operating limits are those previously found acceptable by the NRC and listed in TS Section 5.6.3.b. The analytical methods currently listed support the determination of core operating limits by using those methods applicable to the fuel supplied by AREVA and Westinghouse.

CGS currently operates with AREVA supplied ATRIUM-1 0 fuel and Westinghouse supplied SVEA-96 fuel.The references in CGS TS 5.6.3.b are used to determine the core operating limits for TS 5.6.3.a, which points to TS 3.2.1 -Average Planar Linear Heat Generation Rate (APLHGR), TS 3.2.2 -Minimum Critical Power Ratio (MCPR), TS 3.2.3 -Linear Heat Generation Rate (LHGR), and LCO 3.3.1.3 -Oscillation Power Range Monitor (OPRM)Instrumentation.

The proposed GNF references support determination of GE14 and AREVA ATRIUM-10 core operating limits listed in TS 5.6.3.a. The existing AREVA references in TS 5.6.3.b support determination of ATRIUM-10 APLHGR and LHGR limits in the R19 reload for use in TS 5.6.3.a.1, "The APLHGR for Specification 3.2.1" and TS 5.6.3.a.3, "The LHGR for Specification 3.2.3," respectively.

AREVA and Wes~tinghouse references are retained to support preparation of a COLR if failed fuel replacement is required.

Failed fuel replacement assemblies are typically obtained from the discharged fuel in the spent fuel pool which includes ATRIUM-10 and Westinghouse SVEA-96 fuel.The new topical report, NEDC-33419P, "GEXL97 Correlation Applicable to ATRIUM-10 Fuel," will allow Energy Northwest to use GNF methodology for determination of fuel assembly critical power of AREVA ATRIUM-10 fuel. This correlation is directly applicable to CGS. NEDC-33419P is provided as Attachment 4.Some of the information in topical report NEDC-33419P, "GEXL97 Correlation Applicable to ATRIUM-10 Fuel," is PROPRIETARY to GNF. GNF requests that the PROPRIETARY information be withheld from public disclosure in accordance with 10 LICENSE AMENDMENT REQUEST FOR CHANGES TO TECHNICAL SPECIFICATIONS INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Enclosure Page 7 of 15 CFR 9.17(a)(4) and 10 CFR 2.390 (a)(4). An affidavit by the information owner, GNF, supporting the request for non-disclosure is included in Attachment

4. A non-proprietary version of the topical report is provided in Attachment

5.4.0 TECHNICAL

ANALYSIS 4.1 TS 3.1.4 changes 4.1.1 Changes to LCO 3.1.4 and Table 3.1.4-1 Control Rod Scram Times Energy Northwest is proposing to change the existing LCO 3.1.4, which was based on the analytical methods of a previous fuel vendor, to the approach specified in the STS, which is based on GNF analytical methods, Reference 7.1. The specific changes include changing the allowed scram time values in Table 3.1.4-1 to BWR/5 specific values, of which three values are more conservative (i.e. require faster scram times), and one is relaxed (i.e. allows for a slower scram time). The relaxed position corresponds to the time to reach notch position 45 (5% insertion value) which is proposed to change from the current value of 0.43 seconds to 0.528 seconds. The number of slow control rods is also proposed to be changed to match the GNF analysis from the current value of 8 control rods to a proposed value of 13 which is equivalent to 7% of the total control rod inventory of 185.The purpose of the control rod scram insertion time LCO is to ensure that the negative scram reactivity corresponding to that used in the licensing basis calculations is supported by individual control rod drive scram performance distributions allowed by the TS. The current CGS TS accomplish the above purpose by placing requirements on maximum individual control rod drive insertion times (7.0 second requirement) and average scram insertion times (two-by-two arrays). Reference

7.2 discusses

the approach of utilizing individual scram times vice "average" scram times to determine if the analytical basis (scram reactivity) is met. Reference 7.2 also discusses the bases for BWR/5 scram times and how the analytical models allow for a distribution of scram insertion speeds including "slow" rods. The bases for the relaxation of the 5% insertion time reqluirement for BWR/2-5 and the justification for allowing up to 7% of the total number of control rods to be"slow" is also discussed in Reference 7.2. The approach discussed in Reference 7.2 was adopted into the BWR/4 STS and resulted in simplifying the scram time LCO for plants utilizing the GNF methodology.

Energy Northwest has reviewed the conclusions of Reference 7.2, the resulting changes reflected into the BWR/4 STS, and has determined that this approach is directly applicable to CGS with the transition to GNF fuel and methodology as analyzed via Reference 7.1.4.1.2 Revise Frequency of SR 3.1.4.1 and SR 3.1.4.4 (TSTF-222-A)

These revisions were proposed in TSTF-222-A, Revision 1, and approved by the NRC as reflected in a letter to the Nuclear Energy Institute (NEI) dated May 12, LICENSE AMENDMENT REQUEST FOR CHANGES TO TECHNICAL SPECIFICATIONS INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Enclosure Page 8 of 15 1999. These revisions have been incorporated into the latest approved version of BWR/4 STS issued by the NRC (NUREG-1433, Revision 3.0).The first sentence of TSTF-222-A states its purpose: "Clarify that post-fueling control rod scram time testing only applies to control rods affected by movement of fuel." The remainder of the TSTF-222-A justification discusses how SR 3.1.4.1 effectively requires all rods to be scram time tested even if only one fuel assembly is moved, and how changes proposed in the TSTF would resolve this misinterpretation.

Current CGS SR 3.1.4.1 requires that the scram time of each control rod be verified to be within Table 3.1.4-1 limits prior to exceeding 40% RTP after a refueling or after a shutdown of 120 days or greater. As revised by this request, SR 3.1.4.1 will require scram time testing of each control rod prior to exceeding 40% RTP after a shutdown of 120 days or greater. Refueling is addressed by the following requested revision to SR 3.1.4.4.The current CGS SR 3.1.4.4 requires that the scram time of each affected control rod be verified to be within Table 3.1.4-1 requirements prior to exceeding 40% RTP after work on the control rod or CRD System that could affect scram time and after fuel movement within the reactor pressure vessel. As revised by this request, SR 3.1.4.4 will require scram time testing of each affected control rod prior to exceeding 40% RTP after fuel movement within the affected core cell and after work on the control rod or CRD System that could affect scram time. In a typical, routine refueling outage, all core cells are likely to be affected as a result of some fuel movement, e.g. a spent fuel assembly is replaced with a fresh assembly, a fuel assembly is relocated from one cell to another, or a fuel assembly is reoriented within a core cell.If a core cell is not affected by (1) movement of one of the four fuel assemblies in the cell, (2) replacement of the control rod in that cell, or (3) maintenance on the control rod drive system for the rod in that cell, the scram time of the control rod in that core cell is not expected to be impacted.

As a result there would be no need to conduct scram time testing on that control rod. Furthermore, it is expected that the periodic scram time testing of a representative sample (10% of the control rods), as required by SR 3.1.4.2, will identify any long term phenomenon that could result in degradation of scram time.Revising the second Frequency of SR 3.1.4.4 to require scram time testing after fuel movement "within the affected core cells" clarifies that only those control rods in core cells in which fuel was moved or replaced or control rod maintenance was performed are required to be scram time tested. It is expected that all core cells will be affected in this manner during a routine refueling outage, and therefore, that scram time testing will likely be required on all control rods.

LICENSE AMENDMENT REQUEST FOR CHANGES TO TECHNICAL SPECIFICATIONS INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Enclosure Page 9 of 15 To make the CGS TS consistent with the current version of BWR/4 STS (NUREG-1433, Rev. 3.0) requires changes to the Frequencies of both SR 3.1.4.1 and SR 3.1.4.4. SR 3.1.4.1 changes include deleting the first part of the Frequency and revising the second part. SR 3.1.4.4 changes involve reversing the order of the two parts of the Frequency.

These changes are expected to be of benefit in the conduct of outages in which only a limited number of fuel cells are affected by avoiding the need to perform scram time testing on control rods in core cells that were not affected by fuel moves, control rod replacements, or control rod drive maintenance.

4.2 Add new surveillance SR 3.2.2.2 MCPR is a ratio of the fuel assembly power that would result in the onset of boiling transition to the actual fuel assembly power. The MCPR Safety Limit (SL) is set such that 99.9% of the fuel rods avoid boiling transition if the limit is not violated.

The operating limit MCPR is established to ensure that no fuel damage results during AOOs.To ensure that the MCPR SL is not exceeded during any transient event that occurs with moderate frequency, limiting transients have been analyzed to determine the largest reduction in critical power ratio (CPR). The limiting transient yields the largest change in CPR (ACPR). When the largest ACPR is added to the MCPR SL, the required operating limit MCPR (OLMCPR) is obtained.

One of the inputs into the transient calculations is the scram time of the control rods.The proposed addition of SR 3.2.2.2 adds a requirement to determine the MCPR operating limit after performance of various scram time surveillances.

The Bases of the STS for SR 3.2.2.2 states, "Because the transient analysis takes credit for conservatism in the scram speed performance, it must be demonstrated that the specific scram speed distribution is consistent with that used in the transient analysis." GNF methodology employs this approach to calculating the OLMCPR.After a scram time test is performed via SR 3.1.4.1, 3.1.4,2, or 3.1.4.4, the data from these tests is used to generate the actual scram speed distribution which is then compared with the assumed distribution used by the transient analysis.

This comparison, designated as value T, is then used as an input ior interpolating between the allowed scram time values designated in LCO 3.1.4 and the realistic scram time values to determine OLMCPR.By performing this surveillance within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of obtaining the necessary actual scram time input data, the effective scram speed distribution can be monitored to ensure that the expected minor changes that occur to T during a fuel cycle or after maintenance is performed that could affect scram times remain consistent with the transient analysis.This approach is consistent with the current version of BWR/4 STS (NUREG-1433, Rev.3.0) and Bases. When CGS implemented the STS via amendment 149 in 1997, this surveillance was not adopted because the fuel vendor under contract at that time did not utilize this analytical approach.

With the introduction of GNF fuel in R1 9, and LICENSE AMENDMENT REQUEST FOR CHANGES TO TECHNICAL SPECIFICATIONS INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Enclosure Page 10 of 15 transition to GNF analytical methodology, adoption of this surveillance requirement is now warranted.

4.3 Changes

to COLR TS Safety Limits (SLs) ensure that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and AOOs. The MCPR fuel cladding integrity SL ensures that during normal operation and during AGOs, at least 99.9% of the fuel rods in the core do not experience transition boiling.The margin between calculated boiling transition and the MCPR SL is based on a detailed statistical procedure that considers the uncertainties in monitoring the core operating state. One specific uncertainty included in the SL is the uncertainty inherent in the critical power correlation. .The fuel vendor's critical power correlations are based on data which provide a high degree of assurance that the critical power, as evaluated by the correlation, is within a small percentage of the actual critical power being estimated, The GEXL correlation is an NRC approved GNF method of accurately predicting the occurrence of boiling transition in Boiling Water Reactor (BWR) fuel. The GEXL correlation is necessary for determining MCPR operating limits resulting from transient analysis, the MCPR safety limit analysis, and the core operating performance and design.Energy Northwest plans to insert GE14 fuel in the reactor during the upcoming refueling outage and will begin using GNF's safety analysis methodologies, including GNF's critical power correlation methods during the subsequent operating cycle (cycle 20).The proposal to revise the listing in the COLR to reflect GNF methodology includes listing NEDE-2401 1-P-A and NEDE-2401 1-P-A-US, "General Electric Standard Application for Reactor Fuel (GESTAR II) and Supplement for United States," as a method of determining core limits. GESTAR describes the use of GEXL or GEXL-PLUS as approved methods for critical power correlations; however it does not describe a specific approved method of performing critical power correlations for ATRIUM-10 fuel.Therefore, Energy Northwest is requesting NRC approval to add the GEXL97 reference to TS 5.6.3.b as a correlation method to be used for ATRIUM-10 fuel. In the Safety Analysis process, the GEXL97 correlation is to be applied to the ATRIUM-10 fuel in the mixed core while the appropriate approved GEXL correlation will be applied to the GNF fuel (including the determination of an acceptable MCPR safety limit for the mixed core).Proprietary and non-proprietary versions of the GEXL97 topical report are provided as Attachments 4 and 5 respectively.

LICENSE AMENDMENT REQUEST FOR CHANGES TO TECHNICAL SPECIFICATIONS INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Enclosure Page 11 of 15 5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration Energy Northwest has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment" as discussed below.1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response:

No.The requested changes to the Scram Time LCO and associated scram times are based on ensuring that the analytical approach utilized by GNF is met. A scram time slower than required might result in an increase in the consequences of an accident.

The CGS proposed changes do not constitute an increase to any consequences to any accidents because the slower allowed scram times for the 5% insertion limit and the increase in the number of allowed "slow" control rods are bounded by the GNF analysis which demonstrates that all required limits are met.The frequency at which control rod scram time is verified does not affect any postulated precursors to an accident.

Revising the frequency for verifying scram time of the control rods is not expected to impact the scram time. Verifying that the scram time is acceptable will continue to be required prior to plant startup following fuel movement or work on the control rods or control rod drive system.Therefore, revising the frequency for verifying insertion time to clarify when it is required does not involve a significant increase in the probability or consequences of an accident.The addition of a new administrative surveillance to ensure that the GNF analytical bases continue to be met with SR 3.2.2.2 is an enhancement that requires CGS to confirm that operation of the plant remains within the analyses that supports safe operation.

The addition of Reference 7.1, which has been previously approved by the NRC, to the COLR represents an administrative type change required to support a transition to a different fuel vendor's analytical methods. The proposed change to TS 5.6.3.b also includes the addition of the GEXL97 correlation for CGS (Attachment 4). CGS plans to use the analytical methods of the new fuel vendor GNF for the analysis of the mixed core consisting of ATRIUM-10 and GE14 fuel bundle types. The GEXL97 correlation appropriately determines the critical power for ATRIUM-10 fuel. In addition, the GEXL97 application'range covers the range of expected operation of the ATRIUM-10 fuel during normal steady state and transient conditions in the CGS reload core.

LICENSE AMENDMENT REQUEST FOR CHANGES TO TECHNICAL SPECIFICATIONS INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Enclosure Page 12 of 15 The requested TS changes concern the use of analytical methods and do not involve any plant modifications or operational changes that could affect any postulated accident precursors or accident mitigation systems and do not introduce any new accident initiation mechanisms.

The proposed changes have no effect on the type or amount of radiation released, and have no effect on predicted offsite doses in the event of an accident.

Thus, the proposed changes do not affect the probability of an accident previously evaluated nor do they increase the radiological consequences of any accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response:

No.The proposed TS changes will not change the design function, reliability, performance, or operation of any plant systems, components, or structures.

It does not create the possibility of a new failure mechanism, malfunction, or accident initiator not considered in the design and licensing bases. Plant operation will continue to be within the core operating limits that are established using NRC approved methods that are applicable to the CGS design and the CGS fuel.Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?Response:

No.Sufficiently rapid insertion of control rods following certain accidents (scram time)will prevent fuel damage, and thereby maintain a margin of safety to fuel damage. The proposed changes to the TS ensure that adequate control rod testing continues to be maintained with implementation of this activity.

The administrative changes proposed involving control rod scram time testing continue to meet analytical requirements and hence do not involve a significant reduction in the margin of safety.The proposed changes also involve the addition of GNF methodology, Reference 7.1, and the GEXL97 correlation, Attachment 4, to the list of analytical methods in TS 5.6.3.b that can be used to determine core operating limits. Use of the GEXL97 correlation analytical method provides an equivalent level of protection as that currently provided.

The administrative change involving the GEXL97 correlation does not alter any method of analysis as described in the NRC approved versions of Reference 7.1. The proposed change does not modify the LICENSE AMENDMENT REQUEST FOR CHANGES TO TECHNICAL SPECIFICATIONS INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Enclosure Page 13 of 15 SLs or setpoints at which protective actions are initiated, and does not change the requirements governing operation or availability of safety equipment assumed to operate to preserve the margin of safety.Therefore, these proposed changes do not involve a significant reduction in the margin of safety.5.2 Applicable Requlatory Requirements/Criteria The proposed changes have been evaluated to determine whether applicable regulations and requirements continue to be met.Energy Northwest has determined that the proposed changes do not require any exemptions or relief from regulatory requirements, other than the TS, and do not affect conformance with any General Design Criterion (GDC) differently than described in the Final Safety Analysis Report (FSAR).TS 3.1.4 Control Rod Scram Times satisfies the following regulatory requirements:

10 CFR 50.36 (c)(2)(ii)(C)

Criterion 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, and...10 CFR 50 Appendix A, GDC 10 -Reactor Design The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

As discussed in Section 4.1 above, the changes being proposed ensure that required scram insertion rates for individual control rods and the number of allowed "slow" rods assumed by the GNF analyses, Reference 7.1, are met. Remaining within the analysis bases ensures that the scram function of the CRD System will control reactivity changes during abnormal operational transients and ensure specified acceptable fuel design limits are not exceeded, thus satisfying both regulatory requirements listed above.TS 3.2.2 Minimum Critical Power Ratio (MCPR) satisfies the following regulatory requirements:

10 CFR 50.36 (c)(2)(ii)(B)

Criterion 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.

LICENSE AMENDMENT REQUEST FOR CHANGES TO TECHNICAL SPECIFICATIONS INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Enclosure Page 14 of 15 The operating limit MCPR meets the above described regulatory requirement.

The operating limit MCPR is established to ensure that no fuel damage results during anticipated operational occurrences (AOOs) by ensuring that the MCPR safety limit (SL)is not exceeded.

To ensure that the MCPR SL is nct exceeded during any transient event that occurs with moderate frequency, limiting transients have been analyzed to determine the largest reduction in critical power ratio (CPR). The limiting transient yields the largest change in CPR (ACPR). When the largest ACPR is added to the MCPR SL, the required operating limit MCPR (OLMCPR) is obtained.

10 CFR 50.36(c)(2)(ii)(B)

Criterion 2 is therefore satisfied by ensuring that the MCPR in the core is always maintained greater than or equal to the operating limit MCPR during normal operation.

As discussed in section 4.2 above, the addition of SR 3.2.2.2 to the TS will ensure that the operating limit MCPR accounts for changes in scram time during a fuel cycle or after maintenance is performed that could affect scram times. This will ensure that if there is a change in scram time performance, the operating limit MCPR will be appropriately adjusted so that it continues to ensure that MCPR SL is not violated during AOOs.Maintaining the MCPR in the core greater than or equal to the adjusted operating limit MCPR will therefore ensure Criterion 2 will continue to be satisfied.

TS 5.6.3 Core Operating Limits Report (COLR) satisfies the following regulatory requirements:

10 CFR 50.36 (c)(5) -Administrative Controls Administrative Controls are the provisions relating to organizations and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.10 CFR 50.34, Contents of Applications; Technical Information This regulation requires that Safety Analysis Reports be submitted that analyze the design and performance of structures, systems, and components provided for the prevention of accidents and the mitigation of the consequences of accidents.

The COLR is required as a part of the reporting requirements specified in the CGS TS Administrative Controls section. The TS requires the core operating limits to be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and to be documented in the COLR. In addition, it requires the analytical methods used to determine the core operating limits to be approved by the NRC and listed in the Administrative Controls section of the TS. The proposed TS changes ensure that these requirements are met.As part of the core reload design process, reload safety evaluations are performed to ensure that the safety analyses remain bounding for the design cycle. To confirm that the analyses remain bounding, key inputs to the safety analyses such as the CPR are confirmed to be conservative with respect to the current design cycle. If key safety analysis parameters are not bounded, a re-analysis or re-evaluation of the affected LICENSE AMENDMENT REQUEST FOR CHANGES TO TECHNICAL SPECIFICATIONS INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Enclosure Page 15 of 15 transients or accidents is performed to ensure that the applicable acceptance criteria are satisfied.

The proposed TS are needed to perform reload safety analysis for the next cycle's core which will consist of fuels from two different fuel vendors (AREVA ATRIUM-10 and GNF GE14).In conclusion, based on the considerations discussed above, (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, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.6.0 ENVIRONMENTAL CONSIDERATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement.

However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

7.1 NEDE-24011-P-A and NEDE-24011-P-A-US, "General Electric Standard Application for Reactor Fuel (GESTAR II) and Supplement for United States," Global Nuclear Fuel 7.2 Letter from R.F. Janecek (BWROG) to R.W. Starostecki (NRC), "BWR Owners Group Revised Reactivity Control System Technical Specifications," BWROG-8754, September 17, 1987 7.3 NUREG-1433, Revision 3.0, "Standard Technical Specifications General Electric Plants, BWR/4" LICENSE AMENDMENT REQUEST FOR CHANGES TO TECHNICAL SPECIFICATIONS INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Attachment 1 Proposed Technical Specifications Changes (mark-up)

Control Rod Scram Times 3.1 REACTIVITY CONTROL. SYSTEMS 'r 3.1.4 Control Rod Scram LCO .1.4 The verae scram time of all OPERABLE control rods in all two-by-two arrays shall not exceed the limit s of Table 3.1.4-1.APPLICABILITY:

MODES 1 and 2.ACTIONS-------------------------------------

NOTE---------------------------------

Separate Condition entry is allowed for each two-by-two a ay.CONDITION REQUIRED ACT N COMPLETION TIME A. One or more two-by-two A.1 Decl e each control Immediately arrays with average ro in the two-by-two scram time not within ray with a scram the limits of time slower than the Table 3.1.4-1. average scram time limits "slow." AND /A.2 Verify the total Immediately number of "slow" and inoperable control rods is eight.AND A.3 Verify each "slow" Immediately control rod meets the"slow" control rod separation criteria.(continued)

Columbia Generating Station 3.1.4-1 Amendment No. +44 1691

3.1 REACTIVITY

CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS

NOTE --------------------------------

During single control rod scram time Surveillances, the control rod drive (CRD) pumps shall be isolated from the associated scram accumulator.

SURVEILLANCE FREQUENCY SR 3.1.4.1 Verify each control rod scram time is within the limits Prior to exceeding of Table 3.1.4-1 with reactor steam dome pressure 40% RTP after_>[800] psig. each12reactOr SR 3.1.4.2 Verify, for a representative sample, each tested 120 days control rod scram time is within the limits of cumulative Table 3.1.4-1 with reactor steam dome pressure operation in> [800] psig. MODE 1 BWR/4 STS 3.1.4-1 Rev. 3.0, 03/31/04 I I I Control Rod Scram Times 3.1.4 SURVEILLANCE REQUIREMENTS

NOTE---------------------------------

During single control rod scram time Surveillances, the control rod drive (CRD) pumps shall be isolated from the associated scram accumulator.

..............................................................................

SURVEILLANCE SR 3.1.4.1 Verify each control rod scram time is within the limits of Table, 3.1.4-1 with reactor steam dome pressure > 800 psig.\C, kc-tW.' ~+I Columbia Generating Station 3.1.4-2 Amendment No. 4-44 1691 I I I Control Rod Scram Times 3.1.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.2 Verify, for a representative sample, each 200 days tested control rod scram time is within the cumulative limits of Table 3.1.4-1 with reactor steam operation in dome pressure > 800 psig. MODE 1 SR 3.1.4.3 Verify each affected control rod scram time Prior to is within the limits of Table 3.1.4-1 with declaring any reactor steam dome pressure.

control rod OPERABLE after work on control rod or CRD System that could affect scram time SR 3.1.4.4 Verify each affected control rod scram time Prior to is within the limits of Table 3.1.4-1 with exceeding reactor steam dome pressure > 800 psig. 40% RTP aft r work on con rol rod or CRD System th t could af ect scram tcte Prio to Lrxce gx ding 40%RTafter fuel toement within t e reactor pressure vessel I Columbia Generating Station 3.1.4-3 Amendment No. 149,169 194 Control Rod Scram Times 3.1.4* ~~able 3.1.4-1 ~~Control Rod Scram Times _ --Enter app ble Conditions and Required Actions of LCO 3.1.3, "Control Rod OPERABILITY," for con ith scram times > 7 seconds to notch position 5. These control rods are in accordance with SR 3.1.3.4, and are not considered "slow SCRAM TIMES(a)(b) (seconds)WHEN REACTOR STEAM DOME PRESSURE NOTCH POSITION_>

800 psig 45 04 ~~39 25 5 (a) Maximum scram time from fully withdrawn position, based on de-energization of scram pilot valve solenoids time zero.wh (b) Scram times as a function of reactor steam dome pressure, when< 800 psig, are within established limits.Columbia Generating Station 3.1.4-4 Amendment No. +4-9 1691 Control Rod Scram Times 3.1.4 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.4.3 Verify each affected control rod scram time is within Prior to declaring the limits of Table 3.1.4-1 with any reactor steam control rod dome pressure.

OPERABLE after work on control rod or CRD System that could affect scram time SR 3.1.4.4 Verify each affected control rod scram time is within Prior to exceeding the limits of Table 3.1.4-1 with reactor steam dome 0%'RTP after pressure > [800] psig. uel movement N Nithin the affected core cell AND Prior to exceeding 40% RTP after ork on control od or CRD System that could ffect scram time BWR/4 STS 3.1.4-2 Rev. 3.0, 03/31/04 Control Rod Scram Times 3.1.4 Table 3.1.4-1 (page 1 of 1)Control Rod Scram Times OEALcotoroswtscaties not within the limits of this Table are considered

2. Enter applicable Conditions and Required Actions of [CO 3.1.3, "Control Rod OPERABILITY," for control rods with scram times > 7 seconds to notch position~l These control rods are inoperable, in accordance with SR 3.1.3.4, and are not con-sidered"slow." SCRAM TIMES(a)(b)(seconds)WHEN REACTOR STEAM DOME PRESSURE NOTCH POSITION > [8001 psig[46] [0.44 ][36] [1.08 ][26] [1.831[061 [3.35](a) Maximum scram time from fuldy withdrawn position, based on de-energization of scram pilot valve solenoids at tim e zero.(b) Scram times as a function of reactor steam dome pressure, when < 800 established limits.psig are within BWR/4 STS 3.1.4-3 Rev. 3.0, 03/31/04 MCPR 3.2.2 3.2 POWER DISTRIBUTION LIMITS 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)LCO 3.2.2 APPLICABILITY:

All MCPRs shall be greater than or equal to the MCPR operating limits specified in the COLR.THERMAL POWER > 25% RTP.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any MCPR not within A.1 Restore MCPR(s) to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months limits, within limits.B. Required Action and B.1 Reduce THERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion to < 25% RTP.Time not met.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify all MCPRs are greater than or equal Once within to the limits specified in the COLR. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after> 25% RTP AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter A 40k -S R 3. Z Columbia Generating Station 3.2.2-1 Amendment No. 4-4-9 1691 MCPR 3.2.2 SURVEILLANCE REQUIREMENTS (continued)

BWR/4 STS 3.2.2-2 Rev. 3.0, 03/31/04 Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.3 CORE OPERATING LIMITS REPORT (COLR) (continued)

16. EMF-2292(P)(A), "ATRIUM" -10: Appendix K Spray Heat Transfer Coefficients," Siemens Power Corporation

17. EMF-CC-074(P)(A)

Volume 4, "BWR Stability Analysis-Assessment of STAIF with Input from MICROBURN-B2," Siemens Power Corporation

18. CENPD-300-P-A, "Reference Safety Report for Boiling Water Reactor Reload Fuel," ABB Combustion Engineering Nuclear Operations

19. NEDO-32465-A., "BWR Owners' Group Reactor Stability Detect and Suppress Solutions Licensing Basis ii o o d eload .p ns" C. ermined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.5.6.4 Post Accident Monitoring (PAM) Instrumentation Report When a report is required by Condition B or F of LCO 3.3.3.1,"Post Accident Monitoring (PAM)Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.Columbia Generating Station 5.6-4 Amendment No. 149,154,169,185 190

$.6.3-1 INSERT 1: 20. NEDC-33419P, "GEXL97 Correlation Applicable to ATRIUM-10 Fuel," Global Nuclear Fuel 21. NEDE-2401 1-P-A and NEDE-2401 1-P-A-US, "General Electric Standard Application for Reactor Fuel (GESTAR II) and Supplement for United States," Global Nuclear Fuel LICENSE AMENDMENT REQUEST FOR CHANGES TO TECHNICAL SPECIFICATIONS INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Attachment 2 Proposed Technical Specifications Changes (retyped)

Control Rod Scram Times 3.1.4 3.1 REACTIVITY CONTROL SYSTEMS 3.1.4 Control Rod Scram Times LCO 3.1.4 a. No more than 13 OPERABLE control rods shall be "slow," in accordance with Table 3.1.4-1, and b. No more than 2 OPERABLE control rods that are "slow" shall occupy adjacent locations.

APPLICABILITY:

MODES 1 and 2.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months LCO not met.Columbia Generating Station 3.1.4-1 Amendment No. 149.169 Control Rod Scram Times 3.1.4 SU.RVE.LL.ANE...

REOU .REMENT...................

NOTE-------------------------------------

During single control rod scram time Surveillances, the control rod drive (CRD) pumps shall be isolated from the associated scram accumulator.

SURVEILLANCE FREQUENCY SR 3.1.4.1 Verify each control rod scram time is Prior to within the limits of Table 3.1.4-1* with exceeding reactor steam dome pressure > 800 psig. 40% RTP after each reactor shutdown > 120 days SR 3.1.4.2 Verify, for a representative sample, each 200 days:tested control rod scram time is within the cumulative limits of Table 3.1.4-1 with reactor steam operation in ddme pressure > 800 psig. MODE 1 SR 3.1.4.3 Verify each affected control rod scram time Prior to is within the limits of Table 3.1.4-1 with declaring any reactor steam dome pressure.

control rod OPERABLE after work on control rod or CRD System that could affect scram time (continued)

Columbia Generating St~ation.3 .1 .4 -2.'Amendm6ntýNo'.

149,169,194 29mrT m Control Rod Scram Times 3.1.4 SURVEILLANCE REQUIREMENTS

.SURVEILLANCE FREQUENCY I.-SR 3.1.4.4 Verify, each affected control rod scram time is within the limits of Table.,3.1.4-1 with.--:-reia cto r:-steam dome,.pres~su.re X- 800 p-s ig " Prior. to exceeding 40% RTP after-fuel movement withi.n.J he......affected core Icel l'AND Prior to exceeding 40%RTP after work on control rod or CRD System that could affect scram time I Columbia Generat.ing .St'ation 3.1-.4-3 Amendment No.-149-169.194 Control Rod Scram Times 3.1.4..... Table 3.1.4:1 ...Control Rod Scram Times,----------------------------

NOTES ----------------------------------

1. OPERABLE control rods with scram times not within the limits of this Table are considered "slow." 2. Enter applicable Conditions and Required Actions of LCO 3.1.3, "Control Rod OPERABILITY," for control rods with scram times > 7 seconds to notch position 5. These control rods are inoperable, in accordance with SR 3.1.3.4, and are not considered "slow." SCRAM TIMES(a)(b) (seconds)WHEN REACTOR STEAM DOME PRESSURE NOTCH POSITION 800 psig 45 0.528 39 0.866.25 ... 1 1.917 , 5 3.437 (a) Maximum scram time from fully withdrawn position, based on de-energization of scram pilot valve solenoids at time zero.(b) Scram times as a function of reactor steam dome pressure, when< 800 psig, are within established limits.Columbia Generating Station 3.1.-4-4 Aýendmen't'No.

149,169 MCPR 3.2.2 3.2 POWER DISTRIBUTION LIMITS 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)LCO 3.2. 2IMCPR~s shall' b&:greater than "or equal dperatingl, .i mifs spe~ci..fied in the .COLR.to-the MCPR APPLICABILITY:

THERMAL POWER >X251%,RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any MCPR not within A..'-], R'est re`MCP'R.(s) to'.; 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months limits. within limits.I' 'i ' " ' I .' , ,- * ..., .B. Requi red'Act!ion a nda B. ' THERMAL' POWER ,'to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Com1,etion'" < 25% 'RTP Ti mi nb, mat SURVEILLANCE REQUIREMENTS

..:_______,__",__, SURVEILLANCE,'

FREQUENCY SR 3.2.2.1 Verify a],] MCPRs are greater than or equal Once within 12 to the limits specified in the COLR. hours after >.-: / .!:.. ;:..... ,- .. .:- .j !. ' .L ..., -,-:". -..; 25 '% ".R T P ..'24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter (continued)

Columbia Gen~erat:ing Station 3.2.2-1 Amendment No. 149,169

?9m r 6i)MCPR 3.2.2 SURVEILLANCE REQUIREMENTS

_SU RV0LLIANCE FREQUENCY SRt3: 3 .2 .2 Determine the MCPRimits. .Ocewithin 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each completion of SR 3.4.4..1 Once within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each completion of SR 3.1.4.2 AND Once within 72'hours after each completion of SR 3.1.4.4 r 4 ./I;~' -, 4 V~j 1' ~II .1. .'-Columbia Generatjntg, Station 31.2ý.2-2 Amendment No.

Reporti ngFW4,Pf?5.6 Reporting Requirements 5.6.3 CORE OPERATING LIMITS REPORT (COLR) (continued)

6. XN-NF-80-19(P)(A)

Volume 4, "Exxon Nuclear Methodology for Bojling Water..Rea~ctors.:

Application of the ENC Met o1l'.gy6tO,.BWR.Reqoads,,"'

Exxon"Nuclear" 'Company 7. EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors:

Evaluation and Validation of CASMO-4/MICROBURN-B2," Siemens Power Corporation

8. XN-NF-80-19(P)(A)

Volume 3, "Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description," Exxon Nuclear Company 9. XN-NF-84-105(P)(A)

Volume 1, "XCOBRA-T:

A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis," Exxon Nuclear Company 10. ANF-524(P)(A), "ANF Critical Power Methodology for Boiling Water Reactors," Advanced Nuclear Fuels Corporation

11. ANF-913(P)(A)

Volume 1, "COTRANSA2:

A Computer Program for Boiling Water Reactor Transient Analysis," Advanced Nuclear Fuels Corporation

12. ANF-1358(P)(A) "The Loss of Feedwater Heating Transient in Boiling Water Reactors," Advanced Nuclear Fuels Corporation

13. EMF-2209(P)(A), "SPCB Critical Power Correlation," Siemens Power Corporation

14. EMF-2245(P)(A), "Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel," Siemens Power Corporation

15. EMF-2361(P)(A), "EXEM BWR-2000 ECCS Evaluation Model," Framatome ANP Richland (continued)

Columbia Generating Station 5.6-3 Amendment No. 141,185 190 149,154,158,169,

[F -~>1 q 3 M Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.3 CORE OPERATING LIMITS REPORT (COLR) (continued) 16: EMF.-22.9,2(P)(A),.

."ATRJIUMTM

-10: Appendix K Spray Heat Tra'nsfer Coeffici~ents.,"'

S'iemens Power Corporation

17. EMF-CC-074(P)(A)

Volume 4, "BWR Stability Analysis-Assessment of STAIF with Input from MICROBURN-B2," Siemens Power Corporation

18. CENPD-300-P-A, "Reference Safety Report for Boiling Water Reactor Reload Fuel," ABB Combustion Engineering

'Nuclear' Operations 19.--NEDO-32465-,A, "BWR Owner-s' .G'roup Reactor.Stabil~ity Detect and Suppress Solution's Licensing Basis Methodologyand:,Reload Applications"...

20. NEDC-33419P, "GEXL97 Correla~tion Applicable to ATRIUM-.1.0 ...Fuel G.' o ba L.. Nu,cle.. .Fuel.21. NEDE.2401,1,-P-A.

and. NEDE-24011-PA.US, "General Electric Standard Appli]c.atJion for Realctor Fuel (GESTAR II) and Supplement for United States:," Giobai Nublear :Fuel c.The core operating limits shall b'e determined such that all appl i cabl e i mits' ( e . fu& the'eml re'charical 11ilmi t S"';core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.The COLR, including any midcycle.'rev.isions or supplements,.sha.ll. be.,provided.

upon .ssuance fo.r .each.

toQthe NRC.*-Post Accident Monitoring (PAM) Instrumentation Report 5.6.4 When 'a report is required by Condition B or F of LCO 3.3.3.1,"Post, Accident Monitoring (PAM) Instrumentation," a report shall besubmitted within the following 14 days. The report shall outl*ine the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.Columbia Generating Station 5.6-4 Amendment No.149,,154,169, 18S;199 LIC#StjAMENDMENT REQUEST FOR CHANGES TO TECHNICAL SPECIFICATIONS INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Attachment 3 rposd Techn ica Specifipcations Bases Changes (mark-up)I II (I r I,'.)., I, J, .*'r i LI I Ij -~, :, ,-'. ....... .. ",, ,, -.', I ,) I I' I LWAI s.

Control Rod Scram Times B 3.1.4 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.4 Control Rod Scram Times BASES BACKGROUND The scram function of the Control Rod Drive (CRD) System controls reactivity changes during abnormal operational transients to ensure that specified acceptable fuel design limits are not exceeded (Ref. 1). The control rods are scrammed by positive means, using hydraulic pressure exerted on the CRD piston.When a scram signal is initiated, control air is vented from the scram valves, allowing them to open by spring action.Opening the exhaust valves reduces the pressure above the main drive piston to atmospheric pressure, and opening the inlet valve applies the accumulator or reactor pressure to the bottom of the piston. Since the notches in the index tube are tapered on the lower edge, the collet fingers are forced open by cam action, allowing the index tube to move upward without restriction because of the high differential pressure across the piston. As the drive moves upward and accumulator pressure drops below the reactor pressure, a ball check valve opens, letting the reactor pressure complete the scram action. If the reactor pressure is low, such as during startup, the accumulator will fully insert the control rod within the required time without assi~stance from reactor pressure.APPLICABLE SAFETY ANALYSES The analytical methods and assumptions used in evaluating the control rod scram function are presented in References 2, 3, 4, 5, and 6. The Design Basis Accident (DBA) and transient analyses assume that all of the control rods scram at a specified insertion rate. The resulting negative scram reactivity forms the basis for the determination of plant thermal limits (e.g., the MCPR). Other distributions of scram times (e.g., several control rods scramming slower than the average time, with several control rods scramming faster than the average time) can also provide sufficient scram reactivity.

Surv illance of each individul control r m " it rh sicoit t w r 1 .-Ctra-r aalys-es ram reac m vit ytss tra nt analyses can be met.(continued)

Columbia Generating Station B 3 .1. 4- 1 Revision 241 Control Rod Scram Times B 3.1.4 BASES APPLICABLE SAFETY ANALYSES (continued)

The scram function of the CRD System protects the MCPR r--'7--1ýSafety Limit (SL) (see Bases for SL 2.1.1, "Reactor Core SLs," and LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)"'and the 1% cladding plastic strain fuel design limit (see Bases for LCO 3.2.3, 'LINEAR HEAT GENERATION RATE (LHGR)"), 'Ii-2.which ensure that no fuel damage will occur if these limits are not exceeded.

Above 800 psig, the scram function is designed to insert negative reactivity at a rate fast enough to prevent the actual MCPR from becoming less than the MCPR SL during the analyzed limiting power transient.

Below 800 psig, the scram function is assumed to perform during the control rod drop accident (Ref. 6) and, therefore, also provides protection against violating fuel damage limits during reactivity insertion accidents (see Bases for LCO 3.1.6, "Rod Pattern Control").

For the reactor vessel overpressure protection analysis (Ref. 4), the scram function, along with the safety/relief valves, ensure that the peak vessel pressure is maintained within the applicable ASME Code limits.Control rod scram times satisfy Criterion 3 of Reference 7.LCO The scram times specified in Table 3.1.4-1 are required to ensure that the scram reactivity assumed in the DBA and transient anal is is met. lhe cram times have a m gin to allow up toIght of the contr( rods to have scra times-I---]r that excee the specified lits (i.e., "slow" c/ trol rods in a two y-two array that o not meet the aver ge scram Sassuming a singoo aow by LCO 3.1.3, "Control Rod OPERABILITY")

and an additional control rod failing to scram per, the single failure criterion.

The scram times are specified as a function of reactor steam dome pressure to account for the pressure dependence of the scram times. The scram times are specified relative to measurements based on reed switch positions, which provide the control rod position indication.

The reed switch closes ("pickup")

when the index tube passes a specific location and then opens ("dropout")

as the index tube travels upward. Verification of the specified scram times in Table 3.1.4-1 is Add-el accomplished through measurement of the "drop out" times. z Table 3.1.4 1 is modified by/a Note, which ates that .control r/s with scram tims > 7 seconds e considered inoperab e as required by R 3.1.3.4.(continued)

Columbia Generating Station B 3.1.4-2 Revision 241 I I I Control Rod Scram Times B 3.1.4 BASES LCO (continued)

This LCO applies only to OPERABLE control-rods since inoperable control rods will be inserted and disarmed (LCO 3.1.3). Slow scramming control rods may be conservatively declared inoperable and not accounted for as"slow" control rods.APPLICABILITY In MODES 1 and 2, a scram is assumed to function during transients and accidents analyzed for these plant conditions.

These events are assumed to occur during startup and power operation; therefore, the scram function of the control rods is required during these MODES. In MODES 3 and 4, the control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied. This provides adequate requirements for control rod scram capability during these conditions.

Scram requirements in MODE 5 are contained in LCO 3.9.5,"Control Rod OPERABILITY-Refueling." ACTIONS El',-The ACTIONS Table is modified by a Note indiaver g am t separate Condition entry is allowed for each two-by-two array. This is acceptable since the scram times are g applicable fory a appliatio orray, and the Requireed ctions Actions.t for each Condition provide appropriate compensat Iaton for each two-by-two array not within the averagcram time limits. Complying with the Required Actionsay allow for continued operation and subsequent two-by-to arrays not within the average scram time limits govmned by subsequent Condition entry and application of asu sreiated Required reActions.

A.1, A.2, and A.3 With one or more two-by-t arrays not within the average scram time limits of Tee 3.1.4t1, the scram reactivity rate assumptions i( t safety analysis may not be met.oTaerfo eight co nrol rod iare t"ow"y-two array, with a I scram time sl hrge scram time limits must be declared "sl o m e i t ly o e s r the overall scram reactivity t smt h oa number _of "slow" and inope r a oto od utb mmediately verified to be<ý e ig h s e s re h t t e s f ty a n a ly s is a s s um p t io n sýr t(h aey nlsshssficient margin to assume otal of eight control rods are "slow," one is stuck, andinued)Columbia Generating Station B 3. 1.4-3 Revision 241 Control Rod Scram Times B 3.1.4 BASES ACTIONS A A 2 an'l d s Is A t n e 0 nanother fails to scr Therefore, ensuring the total u 1.a 3 s s eh a h r 1 e 0 r 0 nnumber of "slow"' inoperable is < eight is conservati c dr r Z A.1. A.2. and A.3 (conti ed)e n 0 eT a 1 m n0 r t h0 n b c since the inope ble control.rods are already fully 1 1.inserted).

ensure the local scram reactivity rat is met, each low" control rod must be immediately v ified to t 0 n s u 0 tr m meet the slow" control rod separation cri t eria. The "slow" contr rod separation criteria are met if: a the "slow" con ol rod is separated in all directions f m each "slow".c trol rod and each stuck control rod by y combination of.Ic1. c t0 ip wo or more fully inserted contr 0' rýds OPERABLE, withdrawn control rods that are not s w"; and b) two or less additional inoperable or "slow" control rods are in the same group as the "slow" control r c 0 n t ro.With the verifications describ d above performed e slow C'0 r 'satisfactorily, the scram re /ctivity rate assumptions in the safety analysis will be me and continued operation is allowed.Rpi'-When ay e quireU ActiX and associateý Compl etion 7ime is Inot tt, Ithe rate of negative reactivity insertion during a scram may not be within the assumptions of the safety analyses.

Therefore, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.SURVEILLANCE The four SRs of this LCO are modified by a Note stating that REQUIREMENTS during a single control rod scram time surveillance, the CRD pumps shall be isolated from the associated scram accumulator.

With the CRD pump isolated (i.e., charging valve closed), the influence of the CRD pump head does not affect the single control rod scram times. During a full core scram, the CRD pump head would be seen by all control rods and would have a negligible effect on the scram insertion times.(continued)

Columbia Generating Station B 3.1.4-4 Revision 241 Control Rod Scram Times B 3.1.4 BASES SURVEILLANCE In addition, the sc m times in Table 3.1 -1 are the REQUIREMENTS average of a two-two array. Therefo , a control rod (continued) scram time, as etermined by the foll ing SRs, must be factored inen the average scram ti for all applicabl]

two-by-two ar ys.SR 3.1.4.1 a The scram activity used in DBA and transient analyses is based on assumed control rod scram time. Measurement of the scram times with reactor steam dome pressure > 800 psig demonstrates acceptable sfz.am times for the transients analyzed in References 5A nd/ a P, J -D _ 4 ,$ it Maximum scram insertion times occur at a reactor pressure of approximately 800 psig because of the competing effects of reactor steam dome pressure and stored accumulator energy.Therefore, demonstration of adequate scram times at reactor steam dome pressure > 800 psig ensures that the scram times will be within the specified limits at higher pressures.

Limits are specified as a function of reactor pressure to account for the sensitivity of the scram insertion times with pressure and to allow a range of pressures over which scram time testing can be performed.

To ensure scram time\- testing is performed within a reasonable time followin a V -3. 1 I efiyllinq6r- ) shutdown > 120 days, .._controlirods are required to be tested before exceeding 40% RTP. This Frequency is acceptable, considering the additional surveillances performed for control rod OPERABILITY, the frequent verification of adequate accumulator pressure, and the required testing of control rods affected by work on control rods or the CRD System.SR 3.1.4.2 e oc-ccAt cklANJL 69 Additional testing of a sample of control rods is required to verify the continued performance of the scram function during the cycle. A representative sample contains at least 10% of the control rods. The sample remains representative if no more than 7.5% of the control rods in the sample tested are determined to be "slow." If more than 7.5% of the sample is declared to be "slow" per the criteria in Table 3.1.4-1, additional control rods are tested until this 7.5% criterion (i.e., 7.5% of the entire sample size) is (continued)

Columbia Generating Station B 3.1.4-5 Revision 42 Control Rod Scram Times B 3.1.4 BASES SURVEILLANCE REQUIREMENTS SR 3.1.4.2 (continued)

.71 74-D satisfied, or unti the total number of "slow" control rods (throughout the ore, from all Surveillances) exceeds the LCO limit .For planned testing, the control rods selected for the sample should be different for each test.Data from inadvertent scrams should be used whenever possible to avoid unnecessary testing at power, even if the control rods with data were previously tested in a sample.The 200 day Frequency is based on operating experience that has shown control rod scram times do not significantly change over an operating cycle. This Frequency is also reasonable, based on the additional Surveillances done on the CRDs at more frequent intervals in accordance with LCO 3.1.3 and LCO 3.1.5, "Control Rod Scram Accumulators." SR 3.1.4.3 When work that could affect the scram insertion time is performed on a control rod or the CRD System, testing must be done to demonstrate that each affected control rod retains adequate scram performance over the range of applicable reactor pressures from zero to the maximum permissible pressure.

The scram testing must be performed once before declaring the control rod OPERABLE.

The required scram time testing must demonstrate that the affected control rod is still within acceptable limits.The limits for reactor pressures

< 800 psig are found in the Licensee Controlled Specifications Manual (Ref. 8), and are established based on a high probability of meeting the acceptance criteria at reactor pressures

> 800 psig. Limits for > 800 psig are found in Table 3.1.4-1. If testing demonstrates the affected control rod does not meet these limits, but is within the 7-second limit of t No to Table 3.1.4-1 the control rod can be declared OPERABLE and"slow."___

Specific examples of work that could affect the scram times include (but are not limited to) the following:

removal of any CRD for maintenance or modification; replacement of a control rod; and maintenance or modification of a scram solenoid pilot valve, scram valve, accumulatonisolation valve, or check valves in the piping requirefor scram./ (continued)

Columbia Generating Station B 3. 1.4-6Revision 42 Control Rod Scram Times B 3.1.4 BASES SURVEILLANCE REQUIREMENTS SR 3.1.4.3 (continued)

The Frequency of once prior to declaring the affected control rod OPERABLE is acceptable because of the capability of testing the control rod over a range of operating conditions and the more frequent surveillances on other aspects of control rod OPERABILITY.

SR 3.1.4.4 When work that could affect the scram insertion time is performed on a control rod or CRD System, or when fuel movement within the reactor, pressure vessel occurs, testing must be done to demonstrate each affected control rod is still within the limits of Table 3.1.4-1 with the reactor steam dome pressure > 800 psig. Where work has been performed at high reactor pressure, the requirements of SR 3.1.4.3 and SR 3.1.4.4 will be satisfied with one test.For a control rod affected by work performed while shut down, however, a zero pressure and a high pressure test may be required.

This testing ensures that the control rod scram performance is acceptable for operating reactor pressure conditions prior to withdrawing the control rod for continued operation.

Alternatively, a test during hydrostatic pressure testing could also satisfy both i e lace-,,ro wnen on y rue! movem t occurs, en on ose control rods ssociated with e core cells aff ted by the , tk I'fuel moveme are required o be scram time sted.The Frequency -f once prior to ex eding 4 -% RTP i i jacceptable cause of the capab ity of testing e control ruu at LIdt Qu uaie moure surveil XAnces on other aspey/s of control ro%'OPERABILITY.

REFERENCES

1. 10 CFR 50, Appendix A, GDC 10.It rSc&T-2. FSAR, Section 4.3.2.5.3. FSAR, Section 4.6.1.1.2.5.3.

4. FSAR, Section 5.2.2.2.3.

5. FSAR, Section63 ot -nued)(continued)

)Columbia Generating Station B 3.1.4-7 Revision 241 I I I Control Rod Scram Times B 3.1.4 BASES R E F E R E N C E S 6 .F S A R , S e c t i o ...0 ..(continued)

E 7. 10 CFR 50.36(c)(2)(ii).

8. Licensee Controlled Specifications Manual.$"4..- V.At , ,N c ' V ." -..~ ~.e4~cbi~y

$ysewiJ Columbia Generating Station B 3. 1.4-8 Revision 241 MCPR B 3.2.2 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)BASES BACKGROUND MCPR is a ratio of the fuel assembly power that would result in the onset of boiling transition to the actual fuel assembly power. The MCPR Safety Limit (SL) is set such that 99.9% of the fuel rods avoid boiling transition if the limit is not violated (refer to the Bases for SL 2.1.1.2).

The operating limit MCPR is established to ensure that no fuel damage results during anticipated operational occurrences (AOOs). Although fuel damage does not necessarily occur if a fuel rod actually experiences boiling transition the critical power at which boiling (2e vcj transition s calculated to occur has been adopted as a fuel S- e i i teri on.The onset of transition boiling is a phenomenon that is readily detected during the testing of various fuel bundle designs. Based on these experimental data, correlations have been developed to predict critical bundle power (i.e., the bundle power level at the onset of transition boiling)for a given set of plant parameters (e.g., reactor vessel pressure, flow, and subcooling).

Because plant operating conditions and bundle power levels are monitored and determined relatively easily, monitoring the MCPR is a convenient way of ensuring that fuel failures due to inadequate cooling do not occur.I APPLICABLE SAFETY ANALYSES The analytical methods and assumptions used in evaluating the AOOs to establish the operating limit MCPR are prsen. d in the FSAR, Chapters 4, 6, and 15, and References 4 3, yh J To ensure that the MCPR SL is not exceeded during a" transient event that occurs with moderate frequency, limiting transients have been analyzed to determine the L largest reduction in critical power ratio (CPR). The ty e of transients evaluated are loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.

The limiting transient yields the largest change in CPR (ACPR). When the largest ACPR is added to the MCPR SL, the required operating limit MCPR is obtained.(continued)

Columbia Generating Station B 3.2.2-1 Revision 34 MCPR B 3.2.2 BASES APPLICABLE SAFETY ANALYSES (continued)

The MCPR operating limits derived from the transient analysis are dependent on the operating core flow and power state (MCPRf and MCPRP, respectively) to ensure adherence to fuel design limits during the worst transient occurs with mo as identifie F AR, -Flow dependent MCPR limits are determined by steady-state thermal hydrauli ds using the three-dimensional BWR simulator code n- ------ )and a multi-channel thermal-hydraulic cod. MCPRf curves are provided based on ma mum cr le flow runout transient for ASD e-e nden J mits (MCPRP) e determined by the tllee-dimensional BWR simulator code (frnc )and a multi-channel thermal-hydraulic code (Rfr, ..Due to the sensitivity of the transient onse o initial core flow levels at power leve ow those at which the turbine s and turbine control valve fast closure scram trips are bypassed, high and low flow MCPRp operating limits are provided for operating between 25% RTP and the previously mentioned bypass power level.The MCPR satisfies Criterion 2 of Reference 4.LCO The MCPR operating limits specified in the COLR are the result of the Design Basis Accident (DBA) and transient analysis.

MCPR operating limits that include the effects of analyzed equipment out-of-service are also included in the COLR. The MCPR operating limits are determined by the larger of the MCPRf and MCPRP limits.APPLICABILITY The MCPR operating limits are primarily derived from transient analyses that are assumed to occur at high power levels. Below 25% RTP, the reactor is operating at a slow recirculation pump speed and the moderator void ratio is small. Surveillance of thermal limits below 25% RTP is unnecessary due to the large inherent margin that ensures that the MCPR SL is not exceeded even if a limiting transient occurs.(continued)

Columbia Generating Station B 3.2.2-2 Revision 34 MCPR B 3.2.2 BASES (continued)

SURVEILLANCE SR 3.2.2.1 REQUIREMENTS The MCPR is required to be initially calculated within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER is > 25% RTP and then every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter.

It is compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is based on both engineering judgment and recognition of the slowness of changes in power distribution during normal operation.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance after THERMAL POWER reaches > 25% RTP is acceptable given the large inherent margin to operating limits at low power'. Y25 ,e v els.REFERENCES

1. ANF-524(P)(A)

Revision 2 and Supplements 1 and 2, "ANF Critical Power Methodology for-Boiling Water Reactors," Advanced Nuclear Fuels, November 1990.2. XN-NF-80-19(P)(A)

Volume 1 and Supplements 1 and 2,"Exxon Nuclear Methodology for Boiling Water Reactors-Neutronic Methods for Design and Analysis," Exxon Nuclear Company, March 1983.3. XN-NF-80-19(P)(A)

Volume 3 Revision 2, "Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description," January 1987.4. 10 CFR 50.36(c)(2)(ii).

Columbia Generating Station B 3.2.2-4 Revi-sion 34 03.1.4-1 INSERT 1:1 To ensure that local scram reactivity rates are maintained within acceptable limits, no more than two of the allowed "slow" control rods may occupy adjacent locations.

03.1.4-1 INSERT 2: Table 3.1.4-1 is modified by two Notes which state that control rods with scram times not within the limits of the Table are considered "slow" and that control rods with scram times > 7 seconds are considered inoperable as required by SR 3.1.3.4.03.1.4-1 INSERT 3:1 When fuel movement within the reactor pressure vessel occurs, only those control rods associated with the core cells affected by the fuel movement are required to be scram time tested. During a routine refueling outage, it is expected that all control rods will be affected.0B3.1.4-1 INSERT 4:1 The Frequency of once prior to exceeding 40% RTP is acceptable because of the capability to test the control rod over a range of operating conditions and the more frequent surveillances on other aspects of control rod OPERABILITY.

3.1.4-1 INSERT 5:-When the requirements of this LCO are not met, 1$3.1.4-3 INSERT 1:1 transient analysis is met (Ref. 6). To account for single failures and "slow" scramming control rods, the scram times specified in Table 3.1.4-1 are faster than those assumed in the design basis analysis.

The scram times have a margin that allows up to approximately 7% of the control rods (e.g., 185 x 7% = 13) to have scram times exceeding the specified limits (i.e., "slow" control rods) 13.2.2-1 INSERT 1:1 SR 3.2.2.2 Because the transient analysis takes credit for conservatism in the scram speed performance, it must be demonstrated that the specific scram speed distribution is consistent with that used in the transient analysis.

SR 3.2.2.2 determines the value of T, which is a measure of the actual scram speed distribution compared with the assumed distribution.

The MCPR operating limit is then determined based on an interpolation between the applicable limits for Option A (scram times of LCO 3.1.4, "Control Rod Scram Times") and Option B (realistic scram times) analyses.

The parameter -t must be determined once within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each set of scram time tests required by SR 3.1.4.1, SR 3.1.4.2, and SR 3.1.4.4 because the effective scram speed distribution may change during the cycle or after maintenance that could affect scram times. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is acceptable due to the relatively minor changes in -T expected during the fuel cycle.B3.2.2-21NSERT 1:1 5. NUREG-0562, June 1979.6. NEDE-2401 1-P-A and NEDE-2401 1-P-A-US, "General Electric Standard Application for Reactor Fuel (GESTAR II) and Supplement for United States," Global Nuclear Fuel.7. NEDO-30130-A, "Steady State Nuclear Methods," May 1985.8. NEDO-24154, "Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors," October 1978.

v t e Letter Sequence Request
TAC:MD9247, (Open)
Initiation Request Acceptance... Administration Withholding Request Acceptance, Withholding Request Acceptance Questions 25 November 2008 Answers Results Approval... Other: GO2-09-103, Report of Changes and Errors Pursuant to 10 CFR 50.46
MONTHYEAR2008-09-03 [Table View]

ML082250680

Text

Subject:

References:

Dear Sir or Madam:

Enclosure:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

5.4.0 TECHNICAL

7.2 discusses

4.3 Changes

7.0 REFERENCES

3.1 REACTIVITY

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