ML20205S939

TS Proposed Change 211 to License DPR-28,revising Reactor Spiral Reloading Pattern Beginning Around Source Range Monitor

ML20205S939
Person / Time
Site:	Vermont Yankee
Issue date:	04/20/1999
From:	Wanczyk R VERMONT YANKEE NUCLEAR POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20205S942	List: ML20205S939 ML20205S943
References
BVY-99-58, NUDOCS 9904270094
Download: ML20205S939 (10)

Text

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VERMONT YANKEE-V NUCLEAR POWER CORPORATION

• 185 Old Ferry Road, Brattleboro, VT 05301-7002 -

(802) 257 5271' April 20,1999 BVY 99-58 U.S. Nuclear Regulatory Commission

- ATTN: Document Control Desk Washington,' DC 20555 '

Sabject: - Vermont Yankee Neelear Power Station -

License No. DPR-28 (Docket No. 50-271)

Technical Specifiestion Proposed Change No. 211 Spiral Core Loading Around a Source Range Monitor Pursuant to 10CFR50 ^0, Vermont Yankee (VY) hereby proposes to amend its Facility Operating License, DPR 28, by incorporating the attached proposed change into the VY Technical Specifications.

nis proposed change avises the reactor core spiral reloading pattem such that it begins around a Source Range Monitor (SRM.) ne offloading pattem is the reverse sequence. Although the normal practice for refueling at VY. is to perform core " shuffles" versus full offloads and reloads, this change is needed in the event a full core offload and reload is necessitated.

De VY Technical Specifications currently state: " ... the reactor will be spirally reloaded from the center cell outwards,' until the core is fully loaded." In response to NRC concems during the January 1989 refueling for Browns Ferry Unit 2, General Electric (GE) and EPRI issued NSAC 164L, Guidelines /

for BWR Reactivity Control During Refueling, which recommended modifying the pattern followed for reloading the reactor core. In accordance with the guidance, this revision proposes to change the VY Technical Specifications and the Technical Specifications Bases from the past practice of beginning core loading at the geometric center of the core to the recommended practice of beginning core loading

, hI around a single SRM. . Consistent with the philosophy of NUREG 1433, Standard Technical Specifications General Electric Plants, BWR/4, Revision 1, the Technical Specifications are revised to only stipulate use of the spiral pattern, with the specific details of the pattern provided in the Bases. De GE/EPRI recommendation for the offloading pattern is consistent with the current offloading pattern description in the Technical Specifications; that is, the reverse of the reloading pattern. Since this revision changes the reloading pattern, it also changes the offloading pattem.

Attachment I to this letter contains supporting information and the safety assessment of the proposed change. ' Attachment 2 contains the determination of no significant hazards consideration. Attachment 3 provides the mark-up version of the current Technical Specification pages and the Bases pages.

Attachment 4 is the retyped Technical Specification and Bases pages.

1 VY has reviewed the proposed Technical Specification change and the associated Bases change in

. accordance with 10CFR50.92 and concludes that the proposed change does not involve a significant hazards consideration.

9904270094 990420 r 7 I C i.i 7 PDR ADOCK 03000271 '

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, VERMONT YANKEE NUCLEAR POWER CORPORATION BVY 9058 \ Page 2 The Plant Operations Review Committee and the Nuclear Safety Audit and Review Committee have reviewed th'e proposed Technical Specification change and Bases change and concur with the above

' determinations. Pursuant to 10CFR50.91(b)(1), we have provided a copy of this proposed change and the associated no significant hazards consideration to the appropriate State of Vermont reprcr,entative.

VY has also reviewed the proposed change against the criteria of 10CFR51.22 for environmental considerations and concludes that the proposed change will not increase the types and amounts of cffluents that may be released offsite. Tnus, VY believes that the proposed change is eligible for categorical exclusion from the requirements for an environmertal impact statement in accordance with 10CFR51.22(c)(9).

We request that the Staffissue the subject license amendment no later than September 1999, in order to ensure implementation prior to the next scheduled refueling outage in October 1999.

If you have any questions on this transmittal, please contact Mr. Thomas B. Silko at (802) 258-4146.

Sincerely, VERMO YANKEE NUCLEAR POWER CORPORATION

. . M Robert J. War zy 86 Director of S y and Regulatory Affairs, m%,

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STATEOF VERMONT %g.k.,,

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Then personally appeared before me, Robert J. Wanczyk, who, being duly 4 EUC sworn, d i

~ 4, Jtate apt he is jD Safety and Regulatory Affairs of Vermont Yankee Nuclear Power Corporation, that he fduly authermed'fo exeooy l and file the foregoing document in the name and on the behalf of Vermont Yankee Nuci 'f6w Co !pid )

that the statements therein are tme to the best of his knowiedge and belief. Q{5y;F

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Sally A. Sandstrum, Notary Public My Commission Expires February 10,2003 Attachmer.ts  :

cc: USNRC Region 1 Administrator USNRC ResidentInspector- VYNPS l

USNRC Project Manager- VYNPS Vermont Department of Public Service  ;

l VI.nMON1 YANhl.I:. Nt'ot.I.An Pown.n Colu onAlsoN Docket No. 50-271 l BVY 99-58

. Attachment i Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 211 Spiral Core Loading Around a Source Range Monitor Supporting Information and Safety Assessment of Proposed Change l

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, VLHMON1 YAMEl! NtT:11AH POWF.H COHPORA1HlN BVY 99-58/ Attachment I / Page 1 -

INTRODUCTION

' ne issue of the adequacy of neutron flux monitoring by the Source Range Monitors (SRMs) during core reloading activities was raised by the NRC during the January 1989 refueling for Browns Feny Unit 2.

l In response, General Electric Nuclear Energy issued RICSIL No. 039', which stated in 'part:

"ne USNRC recently reviewed the refueling practices at a BWR located in the United States during a full core reload following an extended shutdown. De USNRC queuioned the adequacy of core neutron flux monitoring during a spiral reloading from the. center of the core. Because the Source Range Monitors (SRMs) initially were separated by water from the region of the core in which fuel was being loaded, the SRMs were not effective in monitoring changes in neutron flux as the fuel was being loaded.

Following its review, the USNRC concluded that adequate monitoring exists only after sufficient fuel is loaded for the SRMs to be in contact with the fueled region."

Subsequently, GE prepared and EPRI issued NSAC 164L, Guidelines for BWR Reactivity Control During Refueling, dated April 19922 This document summarizes on page 5-12 under the heading

" Recommendations", subheading " Neutron Flux Monitoring - Fuel Omond/ Reload", sc-tion (15):

" Initiate fuel reloading adjacent to an SRM or FLC connected to the SRM circuitry. Omonding sequences should be the reverse of the loading sequences. Loading sequences which bring all four SRMs on scale as soon as practicable are recommended." (Note that "FLC" refers to a fuel loading chamber or

" dunking" chamber.)

5 ne BWR/4 Standard Technical Specifications address core reload /omond methodology as it applies to the requirements for SRM instrumentation. Spec'fication 33.1.2, Source Range Monitor (SRM)

Instrumentation requires two channels of SRM instrumentation in MODE 5, but modifies that requirement.with the folidwing Note (b): "Only one SRM channel is required to be OPERABLE during spiral omoed or reload when the fueled region includes only that SRM detector," ne associated Bases for LCO 3J.1.2 states: "In MODE 5, during a spiral omond or reload, an SRM outside the fueled region will no longer be required to be OPERABLE, since it is not capable of monitoring neutron flux in the fueled region of the core. Thus, CORE ALTERATIONS are allowed in a quadrant with no OPERABLE SRM in an adjacent quadrant provided the Table 33.1.2-1, footnote (b), requirement that the bundles being spiral reloaded or spiral offloaded are all in a single fueled region containing at least one OPERABLE SRM is met. Spiral reloading and omonding encompass reloading or omonding a cell on the edge of a continuous fueled region (the cell can be reloaded or omonded in any sequence)." Note that the Standard Technical Specifications also refer to " compliance with an approved [ spiral] reload .

I sequence" as a: requirement for utilizing Special Operations LCO 3.10.6, Multiple Control Rod Withdrawal-Refueling.

The specific details of the core loading pattem do not meet the criteria of 10 CFR 5036(c)(2)(ii) for retention in the Technical Specifications. De requirement for spiral loading and omonding is proposed to be ~ retained in the Technical Specifications, as are the associated SRM requirements. However,  !

. consistent with the Standard Technical Specifications philosophy of placing details of requirement i

Rapid Information Communication Services Infonnation Letter, Full Core Reloading Procedures, issued by General Electric February 10,1989.

8 NSAC 164L, Gaidelines for BWR Reactivity Control During Refueling, prepared by General Electric Company, issued by the Nuclear Safety Analysis Center division of the Electric Power Research Institute, dated April 1992.

NUREG 1433, Standard Technical Specifications General Electric Plants, BWR/4, Revision 1, dated April 7, 1995.'

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. Vamom YAmn Nranw Powrn CosmoeuuoN LVY 99-58 / Attachment I / Page 2 implementation in the Bases, the proposed VY amendment removes the following spiral reloading descriptive phrases from Technical Specification 3.12.E.3.b: I e "De core may be spirally reloaded to either the original configuration or a different configuration in the reverse sequence of that used to unload, ..." is changed to "De core may be spirally reloaded, i

e' "

..., the reactor will be spirally reloaded from the center cell outwards until the core is fully loaded."

is changed to "the reactor will be spirally reloaded until the core is fully loaded."

De Bazs for Technical Specification 3.12.E is changed to incorporate details of the methodology by replacing the current description with the following:

• " Spiral reloading and unloading encompass reloading or unloading a cell on the edge of a continuous fueled region (the cell can be reloaded or unloaded in any sequence.) ne pattern begins (for reloading) and ends (for unloading) around a single SRM. De spiral reloading pattern is the reverse of the unloading pattern, with the exception that two diagonally adjacent bundles, which have previously accumulated exposure in-core, are placed next to each of the four SRMs before the actual spiral reloading begins. The spiral reload can be to either the original configuration or a different configuration."

To achieve the optimum core loading for the cycle, it is sometimes necessary to change the bundles that were initially loaded adjacent to the SRMs as the spiral reload pattern encompasses them. Such bundle changes are a part of the spiral reload pattern. He following purely administrative clarifying change is proposed for Technical Specification 3.12.E.3.b, as well as Technical Specification 3.12.B, Core Monitoring, subsection 4 and the Bases for Technical Specification 3.12.B:

• De words "their designated" are removed from the phrase "two (2) diagonally adjacent fuel assemblies, which have previously accumulated exposure in the reactor, shall be leaded into their designated core positions next to each of the four (4) SRMs to obtain the required 3 cps." His is required prior to beginning spiral reloading of the core. "Their designated" can be interpreted to  !

i mean that the bundles initially loaded adjacent to the SRMs to ensure their OPERABILITY must be the bundles that will occupy those core positions for the next cycle. He bundles are simply being used to ensure proper SRM response; it is overly restrictive to disallow changing the bundles that occupy those positions when the spiral reload pattern encompasses the associated SRM.

SAFETY ASSESSMENT l

Revising the core reloading and offloading pattern to spiral around an SRM is not a safety concern for  !

the following reasons: l (1) NSAC 164L was generated in response to an NRC concern regarding adequate flux monitoring during reload.

(2) NSAC 164L assessed the various reload options, and recommended the spiral reload around an SRM methodology.

(3) Spiral loading around an SRM is consistent with the Standard Technical Specifications.

(4) Spiral loading around an SRM is conservative relative to the pattern stipulated by the current Technical Specifications in that it provides better flux monitoring.

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, VI'.14 HON 1 YANht.3' .Nitt.A.AH PowliH Coluanst VitoN BVY 99-58 / Attachment 1/ Page 3 (5) ne design basis accident associated with refueling is the Refueling Accident; i.e., the accidental

. dropping of a fuel bundle onto the top of the core. Here is no assumption as to the core loading pattern in the analysis of this accident. ,

(6) ne analyzed abnormal operational transients associated with refueling are: 1) the Control Rod Removal Error During Refueling, and 2) the Fuel Assembly Insertion Error During Refueling. Dere is no assumption as to the core loading pattern in the analyses of these transients. He Fuel Assembly Insertion Error During Refueling transient involves mislocated and rotated fuel assembly ,

loading errors. However,a change in the approved core loading pattern has no impact on the probability of mislocating or rotating a bundle while following that pattem.

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. Y HMoNT YANhi;Ii Ni ci. '.Ait l'ows:n CostrustAlsoN Docket No. 50-271 BVY 99-58

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Attachment 2 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 211 Spiral Core Loading Around a Source Range Monitor Determination of No Significant Hazards Consideration l

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. YliHMONT YANhlili NCCI.liAH POWiiH COHl'OHNilON BVY 99-58 / Attachment 2 / Page 1 Pursuant to 10CFR50.92, Vermont Yankee (VY) has reviewed the proposed change and concludes that the, change does not involve a significant hazards consideration since the proposed change satisfies the criteria in 10CFR50.92(c). .

1,' . He oneration of Vermont Yaakaa Nuclear Power Statian in accordance with the oronacad

===d==* will not involve a sinaifie=at incr=v in the orobability or maaaces of an accident previously evaluated- .

i VY has detennined that the proposed change to reload the reactor core in a spiral pattern beginning around a Source Range Monitor (SRM) does not involve a significant increase in the probability or consequences of an accident previously evaluated. He design basis accident associated with refueling is the Refueling Accident; i.e., the accidental dropping of a fuel bundle onto the top of the core. There is no assumption as to the core loading pattem in the analysis of this accident. De analyzed abnormal operational transients associated with refueling are: 1) the Control Rod Removal Error During Refueling, and 2) the Fuel Assembly Insertion Error During Refueling. Here is no assumption as to the ,

core loading pattern in the analyses of these transients. The Fuel Assembly Insertion Error During  !

Refueling transient involves mislocated and rotated fuel assembly loading errors. However, a change in the approved core loading pattern has no impact on the probability of mislocating or rotating a bundle while following that pattern. Furthermore, the proposed change implements a core loading pattem that provides improved flux monitoring as compared to the pattern prescribed by the current Technical Specifications. When loading the core in accordance with the proposed change, the SRM indication will be indicative of the true flux of the loaded fuel, as the creation of flux traps (moderator filled cavities surrounded on all sides by fuel) is precluded.

De Technical Specification Bases are under the purview of 10CFR50.59. As such, subsequent changes j made via 10CFR50.59 to the information relocated to the Bases are not allowed to increase the probability or consequences of an accident previously evaluated. Derefore, relocating the details of the core loading pattern to the Bases does not involve a significant increase in the ' probability or consequences of an accident previously evaluated.

De SRMs a' nd the core loading pattern are not initiators of any accident previously evaluated. As such, the subject changes cannot affect the probability of an accident previously evaluated. He core loading 1 pattern is not assumed in the mitigation of any accident. Since the proposed change provides improved flux monitoring by the SRMs, operators will have more accurate indication and SRM automatic trip functions will actuate more accurately. As such, any event mitigation function provided by the SRMs is enhanced by this change. Herefore, the associated changes do not involve a significant increase in the consequences of an accident previously evaluated.

2. He oneration of Vermont Yankee Nuclear Power Station in accci ?=.am with the procosed

==aad==' will not create the nossibility of a new or different kind of accidant from any accident previously evaluated

- VY has determined that the proposed change does not create the possibility of a new or different kind of accident from any occident previously evaluated. VY proposes to change the core reloading and offloading patterns to start and stop, respectively, at an SRM versus the geometric center of the core as prescribed by current Technical Specifications. This ensures that flux monitoring instrumentation is i always OPERABLE in the fueled region of the vessel. Here is no separation of the monitoring device from the fuel by cavities of water as is the case with the pattern prescribed by the current Technical Specifications. As such, flux monito.ing is enhaarad during core reloading and offloading. His change u

V BVY 99-58 / Attachment 2'/Page 2 r-is conservative relative to the current requirements. Therefore, no new categories or types of accidents are created.

Additionally, the Technical Specification Bases are under the purview of 10CFR50.59. As such, subsequent changes made via 10CFR50.59 to the information relocated to the Bases are not allowed to create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. Therefore, relocating the details of the core loading pattern to the Bases does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. The oneration of Vermont YanM Nuclear Power Station in ac=.Ma .m with the oronosed

===ded will not involve a sinnificant ration in a marnin of safety.

VY has determined that the proposed change does not involve a significant reduction in a margin of safety. Loading around the geometric center of the core as prescribed by the current Technical Specifications results in cells of moderator separating the fuel from the instrumentation monitoring its flux. This change requires the flux monitoring instrumentation to be in the fueled region, and, in so doing, provides for more accurate monitoring of core flux during core reloading and offloading. As such, the operators will have more accurate indication and SRM automatic trip functions will actuate when the actual flux reaches the trip setpoints. This corrects non-conservatisms that result from cells of moderator separating the fuel from the instrumentation. *herefore, this change will not result in a significant reduction in a margin of safety.

Additionally, the details of the loading pattern are relocated from the Technical Specifications to the Bases. Since any future changes to the Bases will be evaluated per the requirements of 10CFR50.59, no reduction in a margin of safety will be allowed. Therefore, relocating the core loading pattern details to i the Bases does not involve a significant reduction in a margin of safety. j i

Summary No Significant Hazards Consideration On the basis of the above, VY has determined that operation of the facility in accordance with the proposed change does not involve a significant hazards consideration as defined in 10CFR50.92(c), in that it: (1) doer not involve a significant increase in the probability or consequences of an accident previously evaluated; (2) does not create the possibility of a new or different kind of accident from any accident previously evaluated; and (3) does not involve a significant reduction in a margin of safety.

In making this determination, Vermont Yankee has also reviewed the NRC examples of license amendments considered not likely to involve significant hazards considerations as provided in the fm' al adoption of 10CFR50.92 published in the Federal Renister. Volume 51, No. 44, dated March 6,1986.

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Docket No. 50-271 BVY 99-58 Attachment 3 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 211 Spiral Core Loading Around a Source Range Monitor Marked-up Version of the Current Technical Specifications and Bases l

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