View source for ELV-00758, Application for Amends to Licenses NPF-68 & NPF-81,revising Rod Insertion Limits of Figure 3.1-3 to Show Fully Withdrawn Position as 222 Steps Instead of 228 Steps

{{Adams
| number = ML20246K221
| issue date = 08/25/1989
| title = Application for Amends to Licenses NPF-68 & NPF-81,revising Rod Insertion Limits of Figure 3.1-3 to Show Fully Withdrawn Position as 222 Steps Instead of 228 Steps
| author name = Hairston W
| author affiliation = GEORGIA POWER CO.
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000424, 05000425
| license number = 
| contact person = 
| document report number = ELV-00758, ELV-758, NUDOCS 8909050410
| package number = ML20246K226
| document type = OPERATING LICENSES-APPLIATION TO AMEND-RENEW EXISTING, TEXT-LICENSE APPLICATIONS &amp; PERMITS
| page count = 11
| project = 
| stage = Request
}}

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ELV-00758
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August 25, 1989 U. S. Nuclear Regulator;, Commission -
ATTN: Document Control Desk Washington, D. C. 20555 L
V0GTLE ELECTRIC GENERATING PLANT NRC DOCKETS 50-424, 52-425 OPERATING LICENSES NPF-68, hEF-81 REVISION TO TECHNICAL' SPECIFICATIONS CONTROL ROD MATERIAL ^AND ROD ~ INSERTION" LIMITS Gentlemen:
In accordance with provision of.10 CFR 50.90 and 10 CFR 50.59, Georgia Power Company (GPC) hereby proposes to amend the Vogtle Electric Generating Plant
                            '(VEGP) Units 1 and 2 Technical Specifications, Appendix A to Operating Licenses NPF-68 and NPF-81.
The proposed changes to the Technical . Specifications revise the rod insertion limits of figure 3.1-3 to show the fully withdrawn position as 222 steps
                           'instead of 228 steps. The purpose of the change in.the insertion limit is to allow a periodic change, within the range of 222 to 231 steps withdrawn, of the parked full' out position of the control rods which will minimize the effects of control rod wear caused by fretting against t,per internal control rod guide- surfaces. Specification 3.1.3.4 will also be revised to assure that control rod drop time measurements are made from the physical fully withdrawn position (231 steps withdrawn). In addition, a revision to 7echnical Speciff-L                           cation 5.3.2 is being proposed that will indicate that control rods may utilize an absorber material of either hafnium or silver-indium-cadmium. The use of either type of absorber material was described in the VEGP Final Stfety Analysis Report (FSAR) and in the NRC's Safety Evaluation Report for VEGP.
However, Section 5.3.2 of the Technical Specifications only described hafnium since that was the only absorber material in use at the time that the VEGP Units 1 'and 2 licenses were issued. Current plans are to replace control rods with absorber material of hafnium with control rods that use silver-indium-cadmium, beginning at the next refueling outage which is scheduled for VEGP Unit 1 in the Spring of 1990.
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GPC plans to begin the use of silver-indium-cadmium absorber material for the control rods and the revised rod insertion limits, starting with Cycle 3 of Unit:1 (Spring of 1990). These changes have been evaluated for Cycle 1 of Unit:2 and will be included in the reload evaluation for future cycles of both units. Therefore, GPC requests that the proposed Technical Specification revisions be applicable to both Units 1 and 2 at the beginning of Cycle 3 for VEGP Unit 1.
Bool 8909050410 890825 PDR        ADOCK 05000424                                                      I \
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                     '-                                                                                              i Georgia Power d U. S. Nuclear Regulatory Commission -
ELV-00758 Page Two l
Enclosure 1 provides a detailed description of the proposed changes and the circumstances necessitating the change request.
Enclosure 2 provides the bases for a determination that the proposed changes do not involve significant hazards considerations.                                            l J
Enclosure 3 provides instructions for incorporating the proposed changes into                  .
the Technical Specifications. The proposed revised pages for the combined                     i VEGP Units 1 anc P ?echnical Specifications are provided as Enclosure 3.
In accordance with 10 CFR 50.91, the designated state official will be sent a                 !
copy of this letter and all enclosures.
4 Mr. W. G. Hairston, III states that he is a Senior Vice President of Georgia Power Company and is authorized to execute this oath on behalf of Georgia Power Company and that, to the best of his knowledge and belief, the facts set forth in this letter and enclosures are true.
GEORGIA POWER COMPANY By:      tJW. >$.                          finNH w G. Hai rston, III l
Sworn to and subscribed before me this MS dayofOgad,1989.
UAA       Re Y hotary pblic JdiCOMEON EXNE CEC.15,1992 l.
j WGH,III/HWM/gm
!                      Enclosures l
xc               (see next page)
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ELV-00758                                                                                                                             i Page~Three:-                 -
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                                                                                                                                                                -l xc w/ enclosures'::                                                                                                             'l Georgia: Power Company.                                                                                                               !
                               'Mr. C. K. McCoy,                                                                                                                  '
Mr.. G. Bockhold, Jr.                                                                                                             R
: Mr. R. M. Odom -
                              .Mr. P. D. Rushton
: NORMS.
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                               . Southern Company Services.                                                                                                           )
                              .Mr. L..B. Long.                                                                                                                  Lj
:-                          = Mr. J. L. Leason '                                                                                                                    l Mr.~' B.' C. Armstrong .                                                                                                         'l U.'  S. Nuclear Regulatory Commission
                              .Mr. S.. D. Ebneter, Regional Administrator L Mr. J. B. Hopkins,! Licensing Project' Manager, NRR Mr. J. F. Rogge, Senior Resident Inspector, Vogtle.
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S l                                                   ENCLOSURE 1 l
l V0GTLE ELECTRIC GENERATING PLANT NRC DOCKETS 50-424, 50-425 OPERATING LICENSES NPF-68, NPF-81 REVISION TO TECHNICAL SPECIFICATIONS CON 1ROL R00 MATERIAL AND R0D INSERTION LIMITS BASIS FOR PROPOSED CHANGE Proposed Changes:
The following changes are included in this submittal:
: 1. In Specification 3.1.3.4 for control rod drop time the word " physical" is inserted into the Limiting Condition for Operation (LCO) such that it states "The individual shutdown and control rod drop time from the physical fully withdrawn position..."                                                                                       ,
: 2. In Specification 3.1.3.5 for shutdown rod insertion limits the LC0 is restated as " All shutdown rods shall be withdrawn to a position greater than or equal to 222 steps" instead of "All shutdown rods shall be fully wi thdrawn" .
In the ACTION statement the phrase "not fully withdrawn" is replaced by
                   " inserted to a position less than 222 steps," and the phrase " Fully                                                      )
withdraw the rcd, or" is replaced by " withdraw the rod to a position                                                      j greater than or equal to 222 steps, or"                                                                                    j j
: 3. In Specification 4.1.3.*;, the Sur veillance Requirement for Shutdown Rod Insertion Limit, the phrase "Eacn shutdown rod shall be determined to b2 fully withdrawn:" is repisced t'y "Each shutdown rod shall be determined to be withdrawn to a position y*ater than or equal to 222 steps:"
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: 4. Figure 3.1-3 " Rod Bank Insertion Limits Versus Thermal Power" is replaced                                                  '
with a Figure that is identical except that it shows the fully withdrawn position as 222 instead of 228.
: 5. In Surveillance Requirement 4.10.1.1 the phrase "each control rod either partially or fully withdrawn" is changed to "each control rod not fully inserted".
: 6. In the bases for Specifications 3/4.2.2 and 3/4.2.3, part "b" is replaced with " Control rod banks are sequenced with a constant tip-to-tip distance between banks as defined by Figure 3.1-3."                                                                                 )
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: 7. In Specification 5.3.2 the description of the absorber composition is revised by inserting the phrase "and/or 80% silver,15% indium, and 5%
cadmium".                                                                                                                  j i
Basis                                                                                                                           j Recent operating experience at other nuclear plants has identified control rod wear as a potential problem. This wear is a result of the fretting of control El-1

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ENCLOSURE 1 (CONT'NUED)
CONTROL R0D MATERIAL AND R0D JNSERTION LIMITS BASIS FOR PROPOSED CHA.NGE rods against upper internal control rod guide surfaces.                                                Because the control rods operate for extensive periods of time in the withdrawn position, the point of wear is local and is determined by the withdrawn parked position. By allowing for a range of parked positions, the potential effects of fretting induced wear can be reduced. The current control rod insertion limits require that fully withdrawn rods be at leart 228 steps withdrawn. The physical limit to rod withdrawal is 231 steps. This proposed change reduces the control rod insertion limit to 222 steps withdrawn. This change will revise the all rods out (AR0) parked position such that the control rods can be periodically repositioned between 222 and 231 steps withdrawn.                Prior to reducing the rod insertion limits it is necessary to review the potential effects on the LOCA and Non-LOCA Accidents to assure that the analyses of these accidents will remain valid for the new limits. This has been done and the results are discussed in Enclosure 2. Since this change to the Technical Specifications would allow the control rods to be placed at a variety of positions between 222 and 231 steps withdrawn, the Technical Specifications were reviewed to assure that- the tenn " fully withdrawn" as used for control rods would not be confusing. As a result of this review the phrase " physical fully withdrawn" was chosen for the control rod drop time specification in order to assure that control rod drop times are measured from the most withdrawn position possibl e. In specification 4.10.1.1 the wording was revised to eliminate the use of the phrase " fully withdrawn" without changing the requirements of the specification.
An additional change has been included with this Technical Specification change request that concerns control rods but is unrelated to control rod position. This change is to that portion of Section 5.3.2 of the Technical Specifications that describes the neutron abscrber material contained in the control rods. The design of VEGP Units 1 and 2 as described in the FSAR allows for the use of either hafnium or silver-indium-cadmium absorber material. This aspect of the design was also noted in the NRC staff's Safety Evaluation Report for VEGP Units 1 and 2. However, since only hafnium was used in the initial core design of VEGP Units 1 and 2, it was the only absorber material mentioned in that section of the Technical Specifications.
GPC expects to replace the current hafnium control rods with rods that use silver-indium-cadmium absorber material at the next refueling outage for VEGP Unit 1 which is scheduled for the Spring of 1990. Similar absorber material will be used in replacement control rods for VEGP Unit 2 at its first refueling outage. Therefore, this change which is being requested for the Technical Specifications is consist. nt with the current design presented in the FSAR and as previously evaluated by the NRC staff. The use of this absorber material was also reviewed by Westinghouse to determine that the previous considerations of the FSAR remain valid. The results of this evaluation are also documented in Enclosure 2.
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l                                                                                  ENCLOSURE 2 V0GTLE ELECTRIC GENERATING PLANT NRC DOCKETS 50-424, 50-425 OPERATING LICENSES NPF-68, NPF-81                                                                             i REVISION TO TECHNICAL SPECIFICATION CONTROL R0D MATERIAL AND R0D INSERTION LIMITS 10 CFR-50.92 EVALUATION Pursuant to 10 CFR 50.92, Georgia Power Company-(GPC) has evaluated the attached proposed amendment to the VEGP Units 1 and 2 Technical Specifications and has determined that operation of the facility in accordance with the proposed amendment would not involve significant hazards considerations. The basis for this determination follows:

===Background===
The current control rod insertion limits as defined in Technical Specification Figure' 3.1-3 identifies the maximum RCCA insertion limit for control banks A, B and C at 100% power as 228 steps withdrawn. The Figure also implies a constant tip-to-tip distance between control banks. The maximum number of steps that a RCCA can physically be withdrawn is 231 steps. During extended periods of operation the control rods may be parked in an all rods out (ARO)'
position of between 228 and 231 steps. The practice at VEGP is currently to park the control rods at 228 steps withdrawn. Past operational history at other plants has indicated that operation with control rods withdrawn to the same position for long periods of time can result in excessive localized control rod wear due to fretting against the upper internals guide surface as a result of flow induced vibration. In order to minimize the potential effects of such wear, GPC proposes to reposition control rods within a range from 222 to 231 steps withdrawn. This requires a change to the RCCA insertion limits, which is included in this proposed change to the Technical Specifications. Even though the amount of overlap between control banks will vary with the AR0 position, the tip-to-tip distance between banks will not change as a result of repositioning.
In addition, GPC has decided to replace the control rods that are currently used in VEGP Units 1 and u with control rods of the same design but utilizing silver-indium-cadmium as the absorber material instead of hafnium. The use of either type of control rod absorber material has previously been reviewed by the NRC for use at VEGP but it was not included in the Technical Specifications. Therefore, this proposed change to the Technical Specifications also includes a revision to the Design Features section to indicate the use of either hafnium (Hf) or silver-indium-cadmium (Ag-In-Cd) as an absorber material in the control cods.
The effects of these revisions to the control rod insertion limits and the use of silver-indium-cadmium absorber material on the accidents that have been analyzed for VEGP Units 1 and 2 have been evaluated. The results of these evaluations are presented in the analysis that follows.
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ENCLOSURE 2 (CONTINUED)
CONTROL R0D MATERIAL AND R0D INSERTION LIMITS 10 CFR 50.92 EVALUATION Analysis The current Technical Specifications allow individual control rods to be parked at any ARO position of 228 to 231 steps withdrawn. The proposed revision would change this to a range of 222 to 231 steps withdrawn and allow the use of Ag-In-Cd control rods. In order to evaluate the effects of these Technical Specification changes, the effects of the proposed changes on the potentially affected safety analyses were evaluated and are discussed below.
1.0      Non-1EA Accident Analyses The RCCA's are modeled in the non-LOCA inalyps in terms of rod worth, total trip reactivity, drop time and reactivity insertion versus time. It has been    ,
determined, via a generic evaluation, that Ag-In-Cd rods used in place of Hf    '
rods will not cause any changes to the safety analysis input parameters used for Vogtle 1 and 2 and that such parameters are appropriate for either rod material. The generic evaluation consisted of performing two parallel core designs, varying only the control rod material. Results confirmed that parameters such as control rod worths and peaking factors remained essentially the same. This conclusion also applies to any combination of hafnium and Ag-In-Cd rods in the core. The two different RCCA types are neutronically equivalent. It is therefore concluded that the assumptions to the non-LOCA analyses do not change dependent upon the control rod material and there is no I           affect on these analyses as a result of this change. The non-LOCA results presented in the FSAR remain bounding.
The data used in the non-LOCA safety analyses was also reviewed with respect to the change in rod insertion limits. The input parameters to the safety analyses remain unchanged and are not affected by the proposed parked position range between 222 and 231 steps withdrawn.
Specifically, the RCCA worth, total trip reactivity, shutdown margin, RCCA drop time and reactivity versus time curve remain valid. In addition, there are no changes to the physics parameters assumed in the safety analyses.
These results apply to both Hf and Ag-In-Cd control rods. Therefore the safety analyses results, as presented in the FSAR remain bounding.
The rod insertion limits Technical Specification change will be implemented during the Unit 1 Cycle 2 to 3 outage (Spring 1990). Thus, the revised Radial Peaking Factor Limit Report for Cycle 3 will incorporate the effects of the revised rod insertion ifmits. Revised radial peaking factors for the remainder of Unit 2 Cycle 1 will be identified via a mid-cycle revision to the Radial Peaking Factor Limit Report in accordance with Technical Specification
: 6. 8.1. 6.
2.0      LOCA and LOCA-Related Analyses Small Break LOCA The small break LOCA analysis for the Vogtle Units is described in FSAR         <
Section 15.6.5. The current FSAR small break LOCA analysis was performed E2-2                                    l
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ENCLOSURE 2 (CONTINUED)
CONTROL R0D MATERIAL AND ROD INSERTION LIMITS 10 CFR 50.92' EVALUATION with the WFLASH and LOCTA Evaluation Model. This Evaluation Model assunes' the reactor core is brought to a subcritical condTon by the trip reactivity of
                   -control rods. Since the rod drop time and. trip reactivity requirements assumed in the safety analyses remain valid for this change in rod material
                  ' and for the change in the rod insertion limits, there will be no effect on the current FSAR small. break LOCA analysis.
Large Break LOCA The large break LOCA analysis for the Vogtle Units is described in FSAR
                   . Section 15.6.5. The current FSAR large break LOCA analysis was performed with the 1981 Evaluation Model . This large break LOCA Evaluation Model does not take-credit for the negative reactivity introduced by the control rods.
During a large break LOCA, the reactor is brought to a subcritical condition by the presence of voids in the core. Since credit was not taken for the control rods there will be no effect on the current FSAR large break LOCA analysis.
Post-LOCA Long-Term Core Cooling The Westinghouse Evaluation Model commitment is that the reactor will remain shutdown by borated ECCS water from the RCS/ sump following a LOCA. Since credit-for the control rods is not taken for a large break LOCA, the borated ECCS water provided by the accumulators and the RWST must have a boron concentration, when' mixed with other water sources, which will result in the reactor core remaining subcritical, assuming all control rods out. Since there is no credit taken for control rods, there will be no change in the calculated RCS/ sump boron concentration after a postulated LOCA, and there is no adverse effect on the post-LOCA long-term core cooling requirement as presented in the FSAR. It is conservative to assume no credit for control rod insertion since this results in-a more reactive condition upon which the evaluation is performed.
Hot-Leg Switchover to Prevent Boron Precipitation Discussion of the hot leg switchover time is found in the Vogtle FSAR Section 6.3.2.8, Section 6.3.3.3, and Section 15.6.5.2. The hot leg swit9over time is determined for inclusion in emergency procedures to ensure no boron precipitation in the reactor vessel following boiling in the core. This time is' dependent on power level, and the RCS, RWST, and accumulator water volumes and boron concentrations. Changing the insertion limit or changing the control rod meterial does not affect either the power level, or the maximum boron concentrations assumed for the RCS, RWST and accumulators in the hot leg
                    .switchover calculation. Therefore, the post-LOCA hot leg switchover time presented in the FSAR remains valid.
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4 ENCLOSURE 2 (CONTINUED)
CONTROL R0D MATERIAL AND R0D INSERTION LIMITS 10 CFR 50.92 EVALUATION Rod Ejection Mass Releases The rod ejection mass releases for dose calculations for the Yogtle Units is presented in FSAR Table 15.4.8-2. The current FSAR rod ejection mass release calculation for the Vogtle Units was performed with the WFLASH Evaluation Model. Since the rod drop time does not change, the change in the control rod material and rod insertion limits does not create an adverse effect on the rod ejection mass releases as presented in the FSAR.
LOCA Hydraulic Forcing' Functions The blowdown hydraulic forces resulting from a LOCA are considered in Section 3.9(N).l.4.6 of the Yogtle FSAR. Based on the current LOCA analyses, the shortest time calculated for tripping the control rods is about 0.5 seconds after a postulated LOCA. The peak loads generated on the reactor vessel and internals as result of a postulated LOCA occur typically between 10 to 50 milliseconds (.01 to .05 seconds) after the break and these loads have subsided well before 500 milliseconds (.50 seconds). Since the hydraulic forces as result of a LOCA have peaked and subsided before the time the control rods are calculated to trip, the change of the control rod material and change in rod insertion limits will have no effect on the structural analysis presented in the Vogtle FSAR.
3.0      Containment Integrity                                                                                    i The change in control rod material and rod insertion limits has no effect on the mass and energy release assumptions used in the containment integrity analyses. Therefore, the results presented in the FSAR remain bounding for these changes.
4.0      Steam Generator Tube Rupture During a steam generator tube rupture, the loss of coolant due to the primary to secondary break flow results in a decrease in the RCS pressure and reactor trip occurs due to low pressurizer pressure. The control rod scram was modeled utilizing the total trip reactivity and the normalized reactivity versus time af ter trip data. Since neither of these are affected, the steam generator tube rupture analysis assumptions and results remain bounding for the change in control rod material as well as for the change to rod insertion limi ts.
: 5. 0 -   Mechanical Performance Evaluation The RCCA's are designed to fall via gravity into the reactor core when power is removed from the drive mechanisms. The time to fall into the dashpot region E2-4

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l ENCLOSURE 2 (CONTINUED)
CONTROL R0D MATERIAL AND R0D IN5ERTION LIMITS 10 CFR 50.92 EVALUATION of''the core is currently assumed to be 2.2 seconds. 'It' has been determined -
that this rod drop time assumption is valid for either rod material such that there will be no change to the value provided for use in the accident analyses.
Rod repositioning.1:; being implemented as a managtment scheme to mitigate 'the effects of wear and prolong the life of the control' rods to perform their functional ' requirements. These functional requirements include providing negative ' reactivity, maintaining geometric configuration for expected duty and maintaining continuous free movement. Mechanically, the repositioning will not ' affect tne operation of the control rod drive mechanisms and as such does
                   .not represent a design or operational issue. Therefore, the conclusions of any analyses which assume a 2.2 second rod drop time remain valid for either rod material and for the range of parked positions from 222 to 231 steps.
RESULTS-The safety analyses have been determined to be unaffected by the changes in-control rod material or rod insertion limit. Also, there are no adverse effects to any safety related fluid system performance, radiological consequences or equipment, qualification issues as a result of these modifications. Based on this and the additional information presented above.
The _ following has been concluded:
: 1.          The change in rod insertion limits and use of either rod material will not increase the probability of an accident.previously evaluated in the FSAR. The. evaluations presented above have determined that the rod drop time and 'other associated parameters assumed in the safety analyses have not changed. Additionally, there are no -9ew failure modes introduced that would increase the probability of any reactivity anomaly occurring.      Therefore, all current safety analyses will remain valid with these changes.
: 2.          The change in rod insertion ifmits and use of either rod mater _tal will not increase the consequences of an accident previously evaluated in
 ,                              the FSAR. It has been determined that all safety related design criteria have been met assuming these changes. These changes will have minimal impact on' core behavior and will not require any setpoint changes nor will the operability of the units be affected. Future reload safety evaluations will review the AR0 position to assure continued conformance with the design parameters.
: 3.          The change in rod insertion limits and use of either rod material will not create the possibility of a new or different kind of accident.from any accident previously evaluated in the FSAR. No new failure mechanisms or new limiting single failures have been identified as a result of these changes.
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ENCLOSURE 2 (CONTINUED)
CONTROL R0D MATERIAL AND R0D INSERTION LIMITS '
r 10'CFR 50.92' EVALUATION
: 4.       The change.in rod insertion' limits and the use of either rod material will not decrease the margin of safety in the bases to any Technical Specification. Since it has been determined that all accident analysis acceptance criteria and design and performance requirements continue to' be met, there is no reduction to any previously-defined margin of safety in the Technical Specifications.
p            It is expected that 'these revisions to -the. Technical Specifications will-l             become effective at the beginning of Cycle 3 for VEGP. Unit 1. In addition, since Unit 2 Cycle 1 will still be'in      3rogress when these revised Technical Specifications become effective, the c1anges will also be acceptable for immediate implementation on Unit 2 Cycle 1.
Conclusion Based on the preceding analysis, GPC has detennined that the proposed change to the Technical Specifications does not involve a significant increase in the probability or consequences of an accident previously evaluated, create the
_ possibility of a new or different kind of accident from any accident previously evaluated or involve a significant reduction in a margin of safet". :Therefore, GPC concludes that the proposed changes meet the requirements of 10 CFR 50.92(c) and do not involve a significant hazards consideration.
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