JSP-066-93, Application for Amend to License NPF-62,revising TS 5.3.1, Fuel Assemblies & 5.3.2, CR Assemblies to Utilize Wording in Suppl 1 to GL 90-02, Alternative Requirements for Fuel Assemblies in Design Features Section of Ts

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Application for Amend to License NPF-62,revising TS 5.3.1, Fuel Assemblies & 5.3.2, CR Assemblies to Utilize Wording in Suppl 1 to GL 90-02, Alternative Requirements for Fuel Assemblies in Design Features Section of Ts
ML20128K094
Person / Time
Site: Clinton Constellation icon.png
Issue date: 02/11/1993
From: Jamila Perry
ILLINOIS POWER CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20128K097 List:
References
GL-90-02, GL-90-2, JSP-066-93, JSP-66-93, U-602085, NUDOCS 9302180067
Download: ML20128K094 (11)
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. Ithno,s Powor Company CWcn Pcme' reaton

' P O Box 0re Cknton. lt 01721 Ts! 217 93542.4 Fan 217 935 40M J. Stephen Perry Seor W.e H evaunt ILLIN/ SIS POWER JSP-066-93 February 11, 1993 U 602085 L47 93(02-11)-LP 8E.100a 10CFR50.90 Docket No. 50 461 Document Control Desk Nuclear Regulatory Commission Washington, D.C. 20555 Subj ect: Clinton Power Station Proposed Amendment of Fncil,ity Oncratinr Licenpye No, NPF 6,,1

Dear Sir:

Pursuant to 10CFR$0.90, Illinois Power (IP) hereby applies for amendment of Facility Operating License No. NPF-62, Appendix A -

Technical Specifications, for Clinton Power Station (CPS). This request consists of proposed changes to Design Features Technical Specifications 5.3.1, " Fuel Assemblies" and 5.3.2, " Control Rod Assemblies". Utilizing the wording provided in supplement 1 to NRC Generic Letter 90 02,

" Alternative Requirements for Fuel Assemblies in the Design Features Section of Technical Specifications," this proposed amendment requests that the fuel design features requirements of Specification 5.3.1 be made more generic to allow use of other NRC approved fuel designs.

Also, this proposed amendment requests revision to Specification 5.3.2 I to allow the use of NRC approved control rod designs trhich contain l hafnium metal in addition to boron carbide powder. Additionally, this proposed amendment requests revision to Technical Specification 3.3.1, l " Reactor Protection System Instrumentation," and the Bases for the l

Reactor Protection System Instrumentation Setpoints to transfer the specific value of the simulated thermal power time constant for the Average Power Range Neutron Monitors (APRMs) from the Technical

! Specifications to the Core Operating Limits Report (COLR). The

! simulated thermal power time constant is an inherent characteristic of the fuel and can change from cycle to cycle (as different fuel designs are utilized in the reactor). Therefore, consistent with the guidance of NRC Ceneric Letter 88-14 " Removal of Cycle Specific Parameter Limits from Technical Specifications," it is appropriate to relocate this information to the COLR.

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A description of the proposed " changes and the associated I justification (including the Basis For No Si 6nificant llazards l Consideration) are provided in Attachment 2. Marked up copies of the affected pages from the current Technical Specifications and Bases are provided in Attachment 3. In addition, an affidavit supporting the facts set forth in this letter and its attachments is provided in ,

Attachment 1.

  • IP has reviewed the proposed change against the criteria of 10CFR$1.22 for categorleal exclusion from environmental impact considerations. The proposed change does not involve a significant hazards consideration, or significantly increase the amounts or change the types of ef fluents thc t may be released off-site, nor does it significantly increase individual or cumulative occupational radiation exposures. Based on the foregoing, IP concludes the proposed changes meet the criteria gf .un in 10CFR51.22(c)(9) for a categorical exclusion from the requirement for an Environmental Impact Statement.

CPS plans to install fuel utilizing a different (but NRC-approved) i design (i.e. , CE10 fuel) during the fourth refueling outage (currently '

scheduled to begin September 26, 1993). Therefore, 1P is requesting that these proposed changes be reviewed on a schedule sufficient to support this outage.

Sincerely yours, f

U S. Pe r Senior Vice President C1.J/mfm Attachments

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NRC Clinton Licensing Project Manager l NRC Resident Office Regional Administrator, Region III, USNRC Illinois Department of Nuclear Safety L i l

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Attachment 1 to U 602085 LS 92 004 STATE OF ILLINOIS

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COUNTY OF DEWITT J. Stephen Perry, being first duly sworn, deposes and says: That he is Senior Vice President of Illinois Power Company; that the application for amendment of Facility Operating License NPF 62 has been prepared under his supervision and direction; that he knows the contents thereof; and that to the best of his knowledge and belief said application and the facts contained therein are true and correct.

DATE: This II day ofVebruary, 1993.

Signed: _ r' A

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Subscribed and sworn to before me this p1 h[he$ruary,1993.

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Notary Public

i Attachment 2 to U-602085 Page 1 of 8 LS-92-004

Background

With the advent of new Boiling Vater Reactor (BWR) fuel designs, one of the major changes ic the increase in number of feel rods per bundle, from the current 8x8 array designs to 9x9 or 10x10 arrays. Use of new fuel designs (including the use of modified 8X8 array fuel designs) may renuit in changes to parameters related to the array or fuel design, such as the number of fuel rods, length of the fuel rods, diameter of the fuet rods, time constant for beat transfer through the fuel pellets, number of water rods, initial fuel rod pressure, fuel rod spacer, active fuel length, fuel channel design, bundle tie plate designs, and size of the water rods, etc. Changes made in the advanced fue? designs offer advantages in thermal limit perf.ormance and fuel cycle economics (including burnup, which can reduce the amount of spent fuel generated by CPS).

Currently, Technical Specification 5.3.1, " Fuel Assemblics," provides a fuel design feature description which includes information on the number q of fuel and water rods, cladding material, active fuel length and bundle enrichments. These details will change with new, NRC-approved

, fuel designs or different enrichments of the same design. Also, these details will change if leaking rods are replaced or lead use assemblies arn used. Currently a change to the Technical Specifications would have to be processed each time tuch details are changed. In order to make the fuel des!sn description more generic to allow the use of NRC-approved designs (such as those dercribed in General Electric Standard

} Application for Reactor Fuel (CESTAR-II), NEDE-240ll-P-A and other NRC-approved documents), IP proposes to revise this Technical Specification utilizing the wording provided in Supplement 1 to NP.C Generic Letter 90-02.

Because the number of fuel rods per bundle may be different for the newer fuel designs, the fuel rod diameter (and hence the fuel pellet diameter) may alco be different. A change in fuel pellet diameter may cause a change in the associated thermal power time constant. (A description of how the simulated thermal power time cor.stant in used as part of a neutron monitaring thermal power trip function is provided below.) Since the thermal power time constant is an inherent characteristic of the fuel and may chango each cycle with the installation of new fuel designs, it is appropriate to relocate the simulated thermal power time constant to the Core Operating Limits Report (COLR). This change is consistent with the guidance provided in Generic Letter 88 16, "Ramoval of Cycle-Spscific Parameter Limits from the Technical Specifications."

o IP elso proposes to change Technical Specification 5.3.2, " Control Rod S Assemblies," to allow the installation of NRC-approved control rod designs that incorporate the use of hafnium metal in addition to boron carbide powder. The current design description is genera'i in naturo.

However, it only addresses boron carbide powder (E4C) as an acceptable neutron-ch iorbing control material. This chsnge would identify hafnium metal as another acceptable control natorial. Also, the current wording discuases a stainless steel sheath which is not currently included in all approved control rod designs which any be used at CPS. Therefore, IF proposes to celete this discussion and add a requirement to use only o

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Attachment 2 to U-602085 Page 2 of 8 LS 92 004 control rod designs which have been previously approved by the NRC for use in BWB/6 plants.

Descrig_ tion of Proposed Channes In accordance with 10CFR50.90, the following changes are being proposed:

(1) Table 3.3.1-2, " Reactor Protection System R sponse Times,"

Functional Unit 2.b: Average Power Range Monitor, Flow Biased Simulated Thermal Power-High, Note ** is being revised to replace the specific value of the simulated thermal power time constant, "of 61 0.6 seconds", with the words "specified in the COLR."

(2) Table 4.3.1.1-1, " Reactor Protection System Instrumentation Surveillance Requirements," Note (1) is being revised by deleting the specific value of the simulated thermal power time constant, "the 6f0.6 second," adding the word "that" before the word "the" and adding at the end of the sentence the phrase, "is within the limits specified in the COLR."

(3) Technical Specification 5.3.1, " Fuel Assemblies," is being revised by replacing the entire paragraph with the NRC provided paragraph in Generic Letter 90-02, Supplement 1 to read:

"The reactor shall contain 624 fuel assemblies. Each assembly shall consist of a matrix of zircaloy clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide as fuel material, and water red (s). Limited substitutions of zirconium alloy or stainless-steel filler rods for fuel rods,

-in accordance with NRC-approved applications of fael rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff-approved codea aad methods, and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead-use assemblies that have not completed representative testing may be placed in non limiting core regions."

(4) Technical Specification 5.3.2, " Control Rod Assemblies," is being revised to add hafnium as another acceptable neutron absorber

'~ material, and remove the requirement that the control material be surrounded by a cruciform shaped stainless steel sheath. In addition, this specification will require the use of NRC-approved control rod designs. The specification is being revised to read:

"The reactor coce shall contain 145 crucifonn-shaped control rod ascemblics as appraved by the NRC. The control material shall be boron carbide powder (B4 C) and/or hafnium metal."

  1. (5) Bases for Table 2.2.1-1, " Reactor Protection System Instrumentation Setpoints," Item 2. Average Power Range Monitor is being revised to replace the specific value of the simulated thermal power time constant, "of 610.6 seconds," with the words "specified in the COLR."

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Attachment 2

. to U 602085 Page 3 of 8 LS-92-004 The proposed changes are reflected on the marked-up copies of pages from the CPS Technical Specifications contained in Attachment 3.

Justification for Proposed Changes Fuel Design Features This proposed amendment revises the necessary Technical Specifications to enable additional NRC-approved fuel assembly designs to be used at Clinton Power Station (CPS). Somewhat detailed descriptions of fuel assembly designs are !.cluded in the Design Features section of the Technical Specifications. The description often includes information on number of fuel and water rods, cladding material, active fuel length and bundle enrichments. These details can change with the loading of new, NRC-approved fuel designs or different enrichments of the same design.

For this recson, IP proposes to revise the fuel design requirements to be more generic but still require that these designs be developed and analyzed using NRC approved codes and methods. This pro osed r design description is consistent with NRC Cenetic Letter 90-02 Supplement 1 and amendments approved for other operating BWRs. Fuel designs used at CPS will be those approved by the NRC in General Electric Standard Application for Reactor Fuel (CESTAR-II) NEDE-240ll-P-A. In addition, design evaluations, as required by 10CFR50.59, will ensure that the licensing basis for the plant continues to be maintained while utilizing advanced fuel designs, g In addition to allowing the use of additional NRC-approved fuel designs, the proposed wording for Fuel Assembly Design Features, as provided in NRC Ceneric Letter 90-02 Supplement 1, allows limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods in accordance wich NRC-approved applications of fuel rod configurations.

llowever, the propoued wording will still require the fuel assemblies to be analyzed with applicable NRC Staff-approved codes and methods and be

, shown by tests or analyses to comply with all fuel smfety design bases, In addition, prior to reconstituting any fuel assemblies, the applicability of the test data used to derive the correlations and limits for the departure from nucleato boiling ratio (DNBR) or for the criticel power ratio (CPR) for proposed configurations will be evaluated. The effect on the mechanical design such as the effect of differential thermal expansion on the proper seating of the fuel rod or on the relaxation of the spacer spring which could lead to fretting wear will he evaluated. In addition, changes in the fuel design that affect the spacer strength or the mass, stiffness, and fundamental frequency of the fuel assembly to ensure that the seismic and loss-of-coolant accident (LOCA) design loading conditions will not cause any structural deformation that could prevent fuel coolable geometry or control rod insertion will be considered-.

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Attachment 2 i . . to U 602085-l Page 4 of 8 LS-92-004 Control Rod Desinn Features a

i- Section 5,3.2 of the CPS Technical Specifications currently requires that the neutron absorber or control material for the control rods be composed of boron carbide powder (B4C). Although the design description

, is general in. nature, the lack of a reference to other neutron absorbing materials precludes the use of control rods having designs utilizing a-

. different composition of control materials, such as hafnium metal. The i ability to utilize different control rod designs containing hafnium

! would be beneficial in that longer control rod lifetimes-could be

achieved, thus resulting in a smaller number of control rods having to be replaced over the life of the plant. This would reduce the amount of radioaccive waste handling required, as.well as reduce the overall-i volume of radioactive waste being generated, stored on site, and ultimately shipped off-site for disposal.

i The existing Technical Specification describes in general terms the design of control. rods which were widely used-at-the time of CPS's

! initial licensing. Subsequent to initial licensing, new control rod designs have been' developed, reviewed.and approved by the NRC,.and'are in use at many operating reactors. Control rods which use hafnium are in use at many BVRs. In order to make use of the new or-upgraded control rod. designs being effered by nuclear. vendors, IP proposes that Section 5.3.2 of the CPS Technical Specifications be modified to allow j

j the use of hafnium as an absorber material. _-This would allow the use of B4 C powder as well as hafnium metal since both have been approved for,

use in control rod designs.

The use of control rods containing hafnlum (for NRC approved control rod 2

designs) does not significantly change the neutronic or mechanical-characteristics of the -control rod. The reasons for choosing. hafnium as a partial 1 substitute for boron carbide is that the ' reactivity life of hafnium is longer than boron carbide as it transmutes to other high absorption cross section isotcpes-and that irradiation assisted! stress j- corrosion cracking of boron carbide rod cladding-is- reduced, leading to longer-blade life.

The proposed Technical Specification change does not. affect any.of-the -,

! other Technical Specifications ansociated with.the control rods. For i example, the required control. rod scram insertion ~ times contained within ,

Specification 3.1.3.2 will be unchanged and must still be met for'any

$ control. rod to be considered operable. ~ All control blades will.be.

generically approved by the NRC for use in BWR-6s.

Simulated Thermal Power-Time Constant The Average lower Range Monitor (APRM) Flow-Biased Simulated Thermal Power-High Function modifies the neutron flux' signal to conservatively.

approximate the thermal power being transferred' from the f ael to the -

reactor coolant. The APRM Flow Biased Simulated -Thermal Power High trip setpoint automatica'ly varies as a function of recirculation drive flou but is clamped at an upper limit which~is lower than the APRM Neutron (

Flux-High Setpoint. ,The APRM Flow Biased Simulated Thermal Power-High' function provides protection against transients where thermal power

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Page 5 of 8 LS-92 004 i

increases slowly (such as the Loss of Feedwater Heating event) and protects the fuel cladding integrity by ensuring the Minimum Critical Power Ratio (MCPR) Safety Limit is not_ exceeded. During these events,

, the thermal power increase does not significantly lag the neutron flux

response and, because of a lower trip setpoint, the APRM Flow Biased 3 Simulated Thermal Power High function will initiate'a scram before-the Neutron Flux-High function. For rapid neutron flux increase events, the

. thermal power lage the neutron flux and the APRM Neutron Flux-HIgh

function will provide a scram signal before the APRM Flow Biased Simulated Thermal Power-Hi E h function setpoint is exceeded. As described within Section 15.1.1.2.2 of the CPS Updated Safety Analysis

, Report (USAR) and approved by the NRC in the Clinton Safety Evaluation Report (SER), the flow biased thermal power monitor (TPM) is the primary.

. protection system trip in mitigating the consequences of a loss of l feedwater heating (or any decrease in core coolant temperature) event.

i SER Section 15.1 notes that the "TPM conservatively estimates thermal

power by passing the APRM5si nal through a time constant." This j simulated thermal power time constant is dependent, in part, on the fuel pellet diameter. Per design, the fuel pellet diameter is reduced when the number of fuel rods is increased as-is.the case with the 9x9 or 4 10x10 fuel designs., As the fuel pellet dianeter decreases, the thermal

! power time constant also. decreases. This is because it takes less time i

for the heat to reach the outer surface of the fuel pellet as the_ radius

decreases. The simulated, conservative time constant for the 8x8 designs is 6 10.6 seconds (as compared to the actual time constant of

) seven to 10 seconds). For the 9x9'and 10x10 designs the actual time a constant is smaller, and hence the corresponding simulated thermal power

} time constant must also be smaller.

The NRC indicated in Generic Letter 88-16 that NRC verification of the i values of the cycle-specific parameter limits which were established j using NRC-approved methodologies is not practical or necessary.

4 Removing the cycle-specific parameter limits from:the Technical Specifications relieves the licensee of the requirement to submit a j- Technical Spocification change roquest to support each refueling. It therefore also relieves the hRC of the. requirement to process a license amendment for each plant refueling. _This represents a significant i savings in resources for both IP and the NRC with no reduction in the i level of safety for CPS. Consistent with other cycle specific parameters that were previously removed from the CPS Technical

Specifications and-relocated to the COLR, the proposed Technical Specification change will remove the cycle-specific thermal power time-j canstant from the Technical Specifications and allowed it to be maintained in'the COLR as this cycle specif.ic parameter limit will I continue to be established using NRC-approved methodologies. The
smallest valuelof the simulated thermal power time constant for all the fuel _ types , nstalled in the core for the cycle will be listed' in the U COLR.

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, besis-For_Bo Sinnificant Hazards Consideration i

In accordance yith 10CFR50.92, a proposed change to the Operating l License (Technical Specifications) involves no significant hazards-l 1

Attachment 2 to U 602085 Page 6 of 8 LS-92-004 consideration if operation of the facility in accordance with the proposed change would not: (1) involve a significant increase in the probability or consequences of any accident previously evaluated, or (2) creste the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety. The proposed changes are evaluated against each of these criteria below.

(1) The proposed changes do not involve a significant increase in che probability or consequences of any accident previously evaluated.

There will be no change in the operation of the facility as a result of the amendment.

The fuel design requirements are being proposed to be more generic but still require that these designs be developed and analyzed using NRC-approved codes and methods. This approach is consistent with NRC Ceneric Letter 90-02 Supplement 1. Further, design evaluations, as required by 10CFR50.59, will ensure that the licensing basis for the plant continues to be maintained while utilizing advanced fuel designs. In addition to allowing the use of NRC-approved advanced fuel designs, the proposed wording for Fuel Assembly Design Features, as provided in NRC Ceneric Letter 90 02 Supplement 1, allows limited substitutions of zirconium alloy or stainless steel filler rods for fuel roda in accordance with NRC-approved applications of fuel rod conffgurations.

However, the proposed wording will still require tha fuel assemblies to be analyzed with applicable NRC staff-approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. As a result, appropriate evaluations will, as discussed in the Generic Letter supplement, be performed.

These evaluations will ensure that there is no significant increase in the probability or consequences of any accident previously evaluated.

The use of hafnium as a neutron-absorbing material has been specifically approved by the NRC for use in BUR control rod assemblies. Use of NRC-approved control rod designs and materials will not significantly alter the neutron absorption (reactivity x

worth), mechanical properties (e.g. , corrosion resistance) or other functional characteristics (e.g. , weight and dimensions) of the control rods in an adversu way. Control rods containing hafnium are designed to be neutronically and physically compatible with the existing B C4 rod design. The proposed change does not alter the required number of control rods nor does it affect any of the Technical Specifications relating to operability or testing of the control rods (e.g., the shutdown margin and scram timing requirements are unaffected). Therefore, the proposed change will not involve a significant increase in the probability or consequences of any accident previously evaluated.

With respect to the proposed amendment to replace the specific value of the cycle-specific simulated thermal power time constant in the Technical Specifications with a reference to the Core Operating Limits Report (COLR), the simulated thermal power time l constant specified in the COLR will continue to be determined

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Attachment 2-

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Page 7 of 8 LS92-004 utilizing NRC-approved analytical methods and will continue to be used for calibration of the APRM Flow Biased Simulated Thermal Power-High trip function in accordance with the Technical

. Specifications. The transfer of the specific value of the fuel and cycle-specific simulated thermal power time constant from the

, CPS Technical Specifications has no impact on the implementation of che associated Technical Specifications. Based on the above,

, the proposed change has no impact on the probability or consequences of any transient or accident occurrence.

(2) The proposed changes do not create the possibility of a new or

', different kind of accident from any accident previously evaluated.

As stated above, no safety functions or plant operation will be

altered as a result of this amendment.

! As described in item (1)-above, the proposed changes to the fuel i design requirements will still require that i.he designs be developed and_ analyzed using NRC-approved codes and methods. In addition, fuel- reconstitution will be perfor!aed within the guidelines of Generic Letter 90-02 Supplement 1 and as a result, no new failure modes will be introduced. Therefore, the proposed change does not create the possibility of a new or different kind

of accident from any accident previously evaluated.

The use of NRC-approved control rod designs using hafnium as an absorber material does not produce any new mode of plant operation or alter the control rods in such a way as to_ adversely affect their function or opera'uility since the new control rods are designed to be compatible with the existing control rods, Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change to replace the specific value of the simulated thermal power time constant with a reference to the COLR is in accordance with the guidance provided in Generic Letter 88-16 for requesting removal of the values of cycle-specific parameters from Technical Specifications. The establishment of the simulated

thermal power time constant will continue to be in accordance with i an NRC-approved methodology._ As a result, no new-failure modes are introduced. Therefore, this change cannot create the possibility of a new or different kind of accident from any accident _previously evaluated.

J-(3) TLe proposed change does not involve a significant reduction in a margin of safety.

As described in item (* above, the proposed changes to the fuel design requirements will still require that the designs be developed and analyzed using NRC-approved codes and methods. In addition, fuel reconstitution sill be performed within the guidelines of Generic letter 90-02 Supplement 1. As a result, the proposed change will not involve a significant reduction-in the margin of safety.

Attachment 2 to U-602085 Page 8 of 8 LS 92 004 The proposed change regarding the luclusion of hafnlum as an acceptable control rod absorber in the Design Features section of the Technical Specifications does not significantly affect the neutronic or mechanical characteristics of the control rods since the hafnium containing control rods are designed to be compatible with the existing design and reload licensing criteria. The proposed change does not change the required number of existing control rods nor does it affect the existing Technical Specifications related to control rods (i.e., required shutdown margin, scram time, etc.). In addition, the proposed change will require that the control rods used at CPS be of those desi 6ns which have been reviewed and approved by the NRC. Further, the margins of safety will continue to be verified in a:cordance with 10CFR50.59 as part of the reload development and review process.

Therefore, the proposed change does not involve a significant reduction in the margin of safety.

With respect to the proposed change to replace the specific value of the simulated thermal power time constant with a referento to the COLR, the value is fuel design dependent and will continue to be determined in accordance with NRC-approved uethods. The proposed amendment does not alter _the requirement that the plant be operated in accordance with the value established for the simulated thermal power time constant. The removal of this value from the CPS Technical Specifications is coincident with its incorporation into the Core Operating Limits Report, which is submitted to the Commission. The CPS administrative procedures control revisions of this value. Therefore, this proposed change is administrative in nature and does not impact the operation of the facility in a manner that involves a significant reduction in the margin of safety.

Based upon the foregoing, IP concludes that this request does not involve a significant hazards consideration.

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