ML20236W739

License Change Application 161 for Amend to License NPF-1, Revising Tech Specs to Include Effects of Nuclear Fuel Design Changes

ML20236W739
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	11/20/1987
From:	Cockfield D PORTLAND GENERAL ELECTRIC CO.
To:
Shared Package
ML20236W729	List: ML20236W727 ML20236W739 ML20236W761 ML20236W768
References
NUDOCS 8712080199
Download: ML20236W739 (8)

Text

- ___ _ - - - - - - - -

PORTLAND GENERAL ELECTRIC COMPANY EUGENE WATER & ELECTRIC BOARD AND PACIFIC POWER & LIGHT COMPANY Operating License NPF-1 Docket 50-344 License Change Application 161 This License Change Application requests modifications to Operating License NPF-1 for the Trojan Nuclear Plant to include the ef fects of nuclear fuel design changes.

PORTLAND GENERAL ELECTRIC COMPANY By C D. W. Cockfield Vice President Nuclear Subscribed and sworn to before me this 20th day of November 1987.

Y3daG V LuJ Notary [t2blico'fOregon My Commission Expires: /

J.hr..b&sk BETTY J. SIMICH kod20ao299 ADocM 05000072 ja044 NOTARY PUBLIC-OREGON P My commission Expires .IP.MIr.{g PDR {

I I

_ _ _ _ _ ___-------------m-----

- - - - - ---- 1

1 l

l I

LCA 161 Page 1 of 7 j DESCRIPTION OF CHANCE A description of the,. proposed changes to the existing Trojan Technical Specifications (TTS)'is as follows. i i

Page Section Description of Change 2-2' 2.1 Safety Figure 2.1-1 is revised to change the four-loop Limits safety limit on the combination of thermal power,  ;

pressurizer pressure and the highest operating loop T avg. This is due to reanalysis with a revised expression for the Enthalpy Rise Hot Channel Factor F{H. See the change description for TTS Page B 2-2,  ;

below.

2-5, 2.2 Limiting Notes 3 and 4 are changed to revise the margin for 2-9 Safety the maximum trip points (Allowable Values) for i System Overtemperature and Overpower AT.

Settings In the overtemperature AT trip setpoint function:

2-7, - T' and T" are revised for consistency with 2-8 current analysis, 2-7 - The definition of ATo in Note 1 is corrected ,

by deleting the parenthetical phrase. Also, the K1 setpoint is revised from 1.32 to 1.28 for ,

consistency with the revised core limits. l l

2-8 - The ft (AI) function is modified for consistency with the revised axial offset limits, and 2-8 - A typographical error is corrected in Note 2.

B 2-2 Bases 2.1.1 Theenthalpyrisehotchannelfactor,F{He 18 l Reactor Core revised from 1.49 to 1.56.

B 2-6 Bases 2.2.1 The minimum limit of 1.73 for the Departure from Reactor Trip Nucleate Boiling Eatio (DNBR) is replaced by the System Instru- phrase "the safety analysis DNBR limit", such that mentation future analysis changes will not require a change Setpoints to the Technical Specification Bases.

3/4 2-5, 3.2.2 The constant values in the expressions for the 3/4 2-7 Heat Flux Hot limits on Fq(2) are revised to 2.50 and 5.00 for Channel Factor greater than, and less than or equal to 50 percent Fn(Z) power, respectively. The K(Z) function in Figure 3.2-2 is also revised, as established by a Loss-of-Coolant Accident (LOCA) reanalysis.

--_-- -- - 3

t (! i >

.kn-

• y +

.}

u .

o LCA 161 I Page 2 of:.7

~.g 3 Page Section Description'of Change 3/4 2-6a An error.is corrected in Paragraph 4.2.2.2.e.

3/4 2-8 -3.2.3,'4.2.3.2 Therevisedvalueof1.56isinsertedforF{H*

3/4 2-9 'RCS Flow Rate,

~ and FR-

'3/4 2-9a' 4.2.3.2 Therevisedvalueof1.56'isinsertedfor'F{H RCS Flow Rate- in the expression for the limit en F R- .i and F 3' B 3/4 2-1 3/4.2.1 The value for Fq(Z) is revised from 2.32 to

' Axial Flux: to 2.50.

Difference.

(AFD) a '

B 3/4 2-4, Bases 3/4.2.2 The revised value of 1.56 is inserted in'the-B 3/4 2-5 and 3/4.2.3: expressionfor.maximumF{H,andrevisedvalues Heat Flux Hot for DNBR margins are incorporated based on incorpo-Chanm.1 Factor, ration of a new rod bow penalty. Also, the 2.5 per-etc. cent DNBR penalty due to core bypass flow is encounted for in the new safety limit DNBR value.

B 3/4 2-6' Bases 3/4.2.5 The minimum limit of 1.73 for DNBR is replaced by DNB Parameters the phrase " greater than or equal to the safety analysis DNBR limit" such that future analysis changes will not require a change to the Technical Specification Bases.

B 3/4 4-1 Bases 3/4.4.1 The minimum limit of 1.73 for DNBR is replaced by Reactor Cool- the phrase "the safety analysis DNBR limit" such ant Loops and that future analysis changes will not require a Coolant change to the Technical Specification Bases.

Circulation B 3/4 5-2 3/4.5.5 A third basis is added to indicate that RWST minimum Refueling volume and' boron concentration must meet suberiti-Water Storage cality requirements for cold post-large break LOCA Tank (RWST) conditions.

5-4 5.3.1 This section is revised to provide for loading of j Fuel substitute rods for fuel rods as required for fuel l cf Assemblies reconstitution.

L i h  !

L

l l

1 LCA 161 Page 3 of 7 r

REASON FOR CHANGE These proposed changes to the Trojan Technical Specifications, Appendix A to Operating License NPF-1, are being made to provide for. improved operating economy. Portland General Electric Company desires to utilize an upgraded fuel design for future Trojan core loadings, beginning with-Cycle 11. Features of the upgraded fuel consist of fuel assemblies with y"

reconstitutable top nozzles, axial fuel blankets and the capability of achieving extended fuel burnups as well as satisfying higher nuclear peaking factors [F and F9 (Z)}. These features were previously reviewed (and approved) by the Nuclear Regulatory Commission (NRC) as Westinghouse Topical Report WCAP-10444-P-A, " Reference Core Report VANTAGE 5 Fuel Assembly",-September 1985. Analysis supporting the higher nuclear peaking factors is described in the enclosed Plant Safety Evaluation Report for the Trojan Nuclear Plant Fuel Upgrade.

The reconstitutable top nozzles incorporate a mechanical disconnect feature that facilitates removal of the top nozzle as required for fuel assembly repairs, if needed. The upgraded fuel assemblies also incorpo-rate fuel rods with long tapered end plugs to provide for easier replace-

. ment of individual fuel rods.

The axial fuel blanket consists of a nominal 6 inches of natural uranium oxide pellets at each end of the fuel pollet stack to reduce neutron .l leakage and improve uranium utilization. This core reactivity decrease is compensated by slightly increasing the enrichment of the remaining fuel.

The upgraded fuel design is capable of achieving a lead fuel rod average burnup of approximately 60,000 mwd /MTU. The basis for this extended burnup design was previously reviewed (and approved) by the NRC as Westinghouse Topical Report WCAP-10125-P-A, " Extended Burnup Evaluation of Westinghouse Fuel", December 1985.

Use of the upgraded fuel desig? in Trojan reload cores is optimized by the use of increased nuclear peaking f actors, Fh and FQ (z), to reduce burnable neutron absorber requirements, improve fuel economy, and increase nuclear design flexibility. Changes to the nuclear design bases provided in the Trojan Nuclear Plant Updated Final Safety Analysis Report (FSAR) are not required. Anticipated changes to the Trojan FSAR are presented in Section 5.0 (Accident Analysis) of the enclosed Plant Safety Evalua-tion Report for the Trojan Nuclear Plant Fuel Upgrade, f

(

i LCA 161 page 4 of 7 SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION The proposed revisions to the Trojan Technical Specifications do not involve a significant hazards consideration because operation of the Trojan Nuclear plant in accordance with these changes:

.(1) Would not involve a significant increase in.the probability or consequences of an accident previously evaluated. The changes to the nuclear fuel design, ie, the removable fuel assembly top nozzle and the axial fuel blanket, are not involved in accident initiation and have been analyzed and determined to have no adverse affect on the consequences of an accident. The changes to power peaking limits also have no effect on accident initiation. They could, '

however, have some effect on the consequences of an accident. The possible accident co'nsequence changes have been evaluated by reana-lyzing the Loss-of-Coolant Accident (LOCA) and the Departure from Nucleate Boiling Ratio (DMBR) margin. The results of these analyses demonstrate that the existing limits remain applicable, and there are no significant changes in the accident consequences due to the changes in the power peaking factor limits. I Changes to reactor trip setpoints for Overtemperature and Overpower AT were made to retain the DNBR protection provided by these trips and thus do not increase the probability or consequeneca of an acci-dent. Additional changes concern the fuel design description to allow substitution of non-fuel rods or vacancies in fuel assemblies to facilitate fuel assembly reconstitution. The details of such substitutions will be evaluated as part of the reload safety evaluation When specific changes are proposed.

(2) Would not create the possibility of a new or different kind of acci-dent from any previously evaluated. The changes do not affect com-ponents, systems or structures that have the capability of causing accidents. The power peaking factor changes affect normal operating limits that are routinely evaluated for reload cores. New or dif-ferent accident situations are not created by small changes in these limits.

It should be noted that some of the changes incorporated in the small break LOCA analysis, for example a revised methodology, higher assumed power level and reduced charging pump flow rate, have caused the limiting small break to shift from a pipe diameter of 4 inches to 3 inches. This is not a new or different kind of accident from those previously analyzed.

As previously noted, the fuel design changes, ie, the reconsti-tutable top nozzles and the axial fuel blanket, are not significant changes from the current fuel design. For the axial fuel blanket, L __ __ _ _ _.

i l

i LCA 161 Page 5 of 7 {

1 the possibility of interchanging enriched fuel and natural uranium ]

blanket fuel pellets is minimal because the pellets are different q lengths, and all fuel rods are gamma scanned as part of the produc- j tion process. In any event, the'results of such a misloading are bounded by the analysis described in Trojan Final Safety Analysis Report (FSAR) Section 15.4.7, " Inadvertent Loading and Operation of j a Fuel Assembly in an Improper Position".

(3) Would not involve a significant reduction in a margin of safety.

The DNBR accident analysis safety limits have been changed to 3 cccount for.the higher power peaking factors. These safety limits {

continue to meet the required design limits (with margin). The l changes to the Overpower and Overtemperature AT setpoints to l accommodate the new core limits utilize the previously used I methodology and retain the margin originally included. The LOCA I analysis demonstrates that the Emergency Core Cooling System (ECCS)  !

acceptance criteria of Title 10, code of Federal Regulations, Part f 50, Section 46'(10 CFR 50.46) continue to be met. j In the March 6, 1986 Federal Register (51 FR 7750-51), the Nuclear Regulatory Commission (NRC) published a list of examples of license amendments that are not likely to involve a significant hazards con-sideration. Example (iii) of that list relates to a change resulting j from a nuclear reactor core reloading, if no fuel assemblies signifi-  !

cantly different from those found previously acceptable to the NRC for a previous core at the facility are involved. As noted in Section 3.0 (Nuc1 ear Design) of the enclosed Plant Safety Evaluation for the Trojan Nuclear Plant Fuel Upgrade, the change from the current standard fuel ,

core to a core containing the upgraded fuel will not require changes to I the current nuclear design bases presented in the Trojan FSAR. An evaluation of the fuel upgrade has demonstrated that the impact of the l fuel upgrade does not cause a significant change to the physics charac- l teristics of the Trojan core beyond the normal range of variations seen j from cycle to cycle. I Example (vi) of the April 6, 1986 list relates to a change which may l either result in some increase to the probability or consequences of a previously analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptance ,

criteria, with respect to the system or component, specified in the l Standard Review Plan. As noted above, the nuclear design bases in the Trojan FSAR remain applicable. Use of the upgraded fuel does not require changes in the nuclear design methods or philosophy, although increased emphasis is placed on the use of three-dimensional modeling due to the ,

axially heterogeneous characteristics of the core with axial fuel l blankets.

The thermal-hydraulic design for the upgraded fuel has been reanalyzed

} with FAH equal to 1.62 to support a Technical Specification increase j

\

E_ j

f i LCA 161 page 6 of 7

, inF{H from 1.49 to 1.56. The effect of the increase in FAH 0" Departure 1from Nucleate Boiling Ratio (DNBR) is offset'by a reduction in the safety analysis DNBR margin. Thus, the transient value of DNBR (for the affected DNB events) will be lower. For this reason, revised core limits were generated based on the increased limit on.FAH, Based on the new core protection limits, the Overtemperature'and Overpower AT setpoint equation constants were recalculated, consistent with the methods outlined in WCAp-8746; " Design Bases for the Thermal Overpower and Overtemperature Delta-T Trip Functions".

A Loss-of-Coolant Accident (LOCA) analysis was also performed to demon-strate that use of the upgraded fuel design, with FAH increased to 1,62 and the maximum LOCA Fq increased to 2.5, is acceptable with respect to the ECCS Acceptance Criteria of 10 CFR 50.46. In addition to the increased peaking factors, the analysis also assumed a core thermal power level of 102 percent of 3558 MWt, and 11.5 percent uniform steam generator tube plugging., The analysis results demonstrate that for both small and large breaks (up to and including the double-ended severance of-a reactor coolant pipe), the ECCS meets the 10 CFR 50.46 acceptance criteria.

Lastly, Example (ix) of the April 6, 1986 list relates to a repair or replacement of a major component or system important to safety, if the-following conditions are met:

(1) The repair or replacement process involves practices Which have been successfully implemented at least once on similar components or systems elsewhere'in the nuclear industry or in other industries, and does not involve a significant increase in the probability or consequences of an accident previously evaluated or create the possibility of a now or different kind of accident from any accident preciously evaluated; and (2) The repaired or replacement component or system does not result in a  ;

significant change in its safety function or a significant reduction in any safety limit (or limiting condition of operation) associated with the component or system.

Consider the nuclear ~ fuel upgrade as replacement of a major component important to safety. As noted in Section 1.0 (Introduction and Summary) of the enclosed plant Safety Evaluation Report for the Trojan Nuclear riant Fuel Upgrade, four fuel assemblies Which incorporated the recon-L stitutable top nozzles, axial blankets, and extended burnup capability were loaded into the V.C. Summer Unit 1 core and were used for power production from December 1984 to March 1987. After two cycles of

(

operation, with an average burnup of approximately 30,000 mwd /MTU, all 1; four assemblies had maintained good mechanical integrity with no evidence of mechanical damage or weer on the VANTACE 5 components and were reloaded into the core for a third operating cycle.

t ; .

, , s

.-l i

t .A LCA 161 Page 7 of 7 _q 4

{

. SAFETY / ENVIRONMENTAL EVALUATION

SUMMARY

'{

. . t. ,

I

. Safety and environmental evaluations were performed as required by-10'CFR 50 and the TTS. This review determined that.'the' proposed changes

' do not create an unreviewed-safety question since Plant operations remain consistent with the' Updated FSAR, adequate surveillance is' maintained, and there is no'significant adverse impact on the environment. An'evalu -

ation of the radiological impact'on postulated accident consequences due to the extended fuel burnup is presented in Appendix B of the enclosed Plant Safety Evaluation' Report for the Trojan Nuclear Plant Fuel Upgrade.

i

-l y

1 l

l l

i I

i l

ARA /kal 'l 3

6048k.1087 I

i i

i I

l

_ - - _ _ _ _ - _ _ _ _ _ _ _