ML20203J149

Proposed Tech Specs Revising Pages 3/4 2-2,3/4 2-7 & 5.0 5-3 Re Power Distribution Limits,Design Features & Max U Weight

ML20203J149
Person / Time
Site:	Byron, 05000000
Issue date:	07/30/1986
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20203J138	List: ML20203J132 ML20203J149
References
1904K, NUDOCS 8608050190
Download: ML20203J149 (11)

Text

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ATTACHMENT 1 Proposed Chances to Appendix A Technical Specifications for Byron Station Units 1 and 2 Facility Operating License NPF-37 Revised Page: 3/4 2-2 3/4 2-7 5.0 5-4

.5 8608050190 DR 860730 ADOCK 05000454 j PDR

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POWER DISTRIBUTION LIMfTS LIMITING CONDITION FOR OPERATION ACTION (Continued)

c. With the indicated AFD outside of the above required target band for more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of cumulative penalty deviation time during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and with THERMAL POWER less than 50% but greater than 15% of RATED THERMAL POWER, the THERMAL POWER shall not be increased equal to or greater than 50% of RATED THERMAL POWER until the indicated AFD is within the above required target band.

SURVEILLANCE RE0VIREMENTS 4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by:

a. Monitoring the indicated AFD for each OPERABLE excore channel:

1) At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and

2) At least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after restoring the AFD Monitor Alarm to OPERABLE status.

b. Monitoring and logging the indicated AFD for each OPEP.ABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and.at least once per 30 minutes thereafter, when the AFD Monitor Alarm is-inoperable. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging.

4.2.1.2 The indicated AFD shall be considered outside of,its target band when two or more OPERABLE excore channels are indicating the AFD to be outside the target band. Penalty deviation outside of the above required target band shall be accumulated on a time basis of:

a. One minute penalty deviation for each 1 minute of POWER 0.PERATION l

outside of the target band at THERMAL POWER levels equal to or above 50%,of RATED THERMAL POWER, and l b. One-half minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER.

4.2.1.3 The target flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective Full Power Days.

l The provisions of Specification 4.0.4 are not applicable. -

l l 4.2.1.4 The target flux difference shall be updated at least once per 31 Effective Full Power Days by either determining the target flux difference pursuant to Specification 4.2.1.3 above or by linear interpolation between the most recently measured value and at the end of the cycle life. The previsions of Specification 4.0.4 re not acplicatie.

h* P'r*rf r fedickd valhe BYRON - UNITS 1 & 2 2/4 :- :

POWER DISTRIBUTION LIMITS SURVEILLANCE RE0VIREMENTS (Continued)

C

2) When the F xy is less than or equal to the F RTP limit for the xy appropriate measured core plane, additional power distribution maps shall be taken and F C compared to F N and F l xy xy xy at least once per 31 EFPD.

e. The F S xy limits for RATED THERMAL POWER (Fxy ) shall be 1.71 for all core planes containing Bank "0" control rods and 1.55 for all unrodded core planes; .

f. The F limits of Specification 4.2.2.2e., above, are not applicable in the following core planes regions as measured in percent of core height from the bottom of the fuel:

1) Lower core region from 0 to 15%, inclusive,

2) Upper core region from 85 to 100%, inclusive,

3) agrid Whin.t2% 4 -ct plane regions _,..:

j W t. : _ u , inclusive, and

.. _ m., .._m., ,.m 2~\

4) Core plane regions within 2% of core height (t 2.88 inches) about the bank demand position of the Bank "0" control rods.

g. With F exceeding F ', the effects of F n Fq (Z) shall be xy xy evaluated to determine ifqF (Z) is within its limits.

I 4.2.2.3 When qF (Z) is measured ft? other than F xy determinations, an overall measured qF (Z) snall be obtained f{ om a power distribution mac and increased by 3% to account for manufacturinc tolerances and further increasec by 5% to account for measurement uncertainty.

D BYRON - UNITS 1 & 2 3/4 2-7

DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The core shall contain 193 fuel assemblies with each fuel assembly containing 264 fuel rods clad with Zircaloy-4. Each fuel rod shall have a nominal active fuel length of 144 inches, nd ;;, tu... . .... i;;; teta'. ..:e,'. ..g3- l g;51?19 ;r::; c emier. The initial core loading shall have a maximum enrichment of less than 3.20 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 4.0 weight percent U-235.

CONTROL ROD ASSEMBLIES .

5.3.2 The core shall contain 53 full-length and no part-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 142 inches of absorber material. All control rods shall be hafnium, clad with stainless steel tubing.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1' The Reactor Coolant System is designed and shall be maintained:

a. In accordance with the Code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,

b. For a pressure of 2485 psig, and

c. For a temperature of 650 F, except for the pressurizer which is 680*F.

VOLUME 5.4.2 The total water and steam volume of the Reactor Coolant System is 12,257 cubic feet at a ncminal T,yg of 588.4*F.

5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.

BYRON - UNITS 1 & 2 5-4

3 ATTACHMENT 2 Summary of Changes Three changes have been identified in the Byron Station Technical Specifications due to differences between Unit 1 and Unit -'

2, or are administrative in nature as indicated below:

1) Page 3/4 2-2, Surveillance Requirements 4.2.1.4 Replace "Ot" with "the percent predicted value."

2) Page 3/4 2-7, Surveillance Requirements 4.2.2.2f3 Insert the words "Within 2% of " at the beginning of the sentence and delete "at 17.5 2%, 31.8 2%,

46.0 1 2%, 60.3 1 2% and 74.6 i 2%, so the sentence reads, "Within 2% of grid plane regions inclusive, and"

3) Page 5.0 5-3.0 5.3.1 Delete the words "and contain a maximum total weight of 1619 grams uranium."

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7 ATTACHMENT 3 Evaluation of Sionificant Hazards Consideration A. Axial Flux Difference (page 3/4 2-2)

Description of Amendment Request The requested change is to replace "Ot" from the Technical Specifications with the words " percent predicted value" at the end of cycle life for determining target flux difference.

The Byron Technical Specifications allows the target flux difference to be updated every 31 days by linear interpolation between the most recently measured value and 0% at the end of cycle. The target flux difference, at end of cycle is a predicted

• value that may vary between Byron Units 1 and 2 and may also change with each fuel reload.

The proposed amendment would allow Commonwealth Edison to use the predicted percent value, which would be obtained from the Nuclear Design Report or other similar documentation provided by the fuel vendor, thus resulting in a more accurate interpolation. The proposed amendment will also prevent having to make a Technical Specification change each time the predicted value is revised.

f This proposed Technical Specification change is also applicable to Braidwood Units 1 and 2 should be included in the draft Braidwood Technical Specifications.

l Basis for No Significant Hazards Determination Commonwealth Edison has evaluated this proposed amendment and determined that it involves no significant hazards considerations. In accordance with the criteria of 10CFR 50.92(c),

a proposed amendment to an operating license involves no significant hazards considerations if operation of the facility in accordance with the proposed amendment would not:

! (1) Involve a significant increase in the probability or consequences of an accident previously evaluated l

because:

(a) A number obtained from the Nuclear Design Report, is being substituted for zero into an NRC approved methodology for determining the target flux difference. This value better reflects the actual end of cycle flux, resulting in a more accurate axial flux difference when the predicted percent value is used. Therefore, using~the predicted end of s[qnkfkcaNtkyfn!EeaseEEe'pfo$$b!Skt or i

t consequences of a previously evaluatek accident.

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(b) The accident of most interest in this evaluation is a loss-of-ccolant accident (LOCA). There is no causal relationship between the axial flux difference and accidents. Therefore, the probability is unaffected.

(2) Create the possibility of a new or different kind of accident from any accident previously evaluated because:

(a) The change being made is to a number that is used in an NRC approved method of calculating the axial flux difference. The proposed amendment does not involve any new equipment, or the manner in which the system is being operated. Hence, the possibility of a new or different kind of accident being created than --

was previously evaluated is unaltered.

(3) Involve a significant reduction in the margin of safety because:

(a) Using the predicted end of cycle flux number instead of zero will allow us to calculate a more accurate target flux difference, providing greater confidence that the margin of safety has not been reduced.

Therefore, since the proposed license amendment satisfies the criteria specified in 10CFR50.92. Commonwealth Edison has t determined that a no significant hazards consideration exists for this item and requests its approval in accordance with the provisions of 10CFR50.91(a)(4).

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l l

F B. Grid Plane Locations (page 3/4 2-7)

Description of Amendment Request The requested change is to delete the reference to exact grid plane locations on fuel bundles from the Technical Specifications and substitute in its place a reference to within 12%

of each of the grid plane regions. This change is sought accomodate the fact that the grid strap locations vary between the reduced bow design and the non-reduced rod bow design.

The initial core load at Byron Unit 2 and Braidwood Units 1 and 2 will utilize two regions, one with a reduced rod bow and the other with the standard rod bow design. By deleting the exact grid plane locations from the Technical Specifications, we will be able j to accomodate these reload specific differences without having to seek a Technical specification change on a cycle-by-cycle basis.

l There are no hardware changes associated with the proposed license amendment. The change is considered to be generic in nature.

This change is also applicable to Braidwood Units 1 and 2 and should be included in the draft Braidwood Technical Specifications.

Basis for No Significant Hazards Determination Commonwealth Edison has evaluated this proposed amendment and has determined that it involves no significant hazards consideration. In accordance with the criteria of 10CFR50.92(c),

the proposed amendment does not:

(1) Involve a significant increase in the probability or consequences of an accident previously evaluated because:

(a) Deleting the specific Technical Specification has no impact on the physical design of the plant. The actual location of the grid straps are appropriately modeled for accident analysis purposes using NRC approved methods. Hence, the probability and consequences of all previously evaluated accidents is not altered.

(2) Create the possbility of a new or different kind of accident from any accident previously evaluated, because:

(a) The function of the grids straps is not adversely affected by changing its location to 2% of each of the grid plane regions.

Therefore, a revision to the grid strap location will not create the possiblity of a

, new or different kind of accident previously evaluated.

(b) This is considered to be an administrative change to take into account differences in grid strap locations for reduced rod bow and non-reduced rod bow designs.

3. Involve a significant reduction in a margin of safety, because:

(a) The information being deleted from the Technical Specifications is cycle specific in nature and is available in the nuclear design report. This change has no direct affect on operations assuring that the margin of safety has not been compromised.

Based on the preceeding assessment, Commonwealth Edison has determined that this proposed amendment involves no significant hazards consideration and request its approval in accordance with the provisions of 10CFR 50.91(a)(4).

1904K

C. Maximum Uranium Weight per Fuel Rod (5.0 5-4)

Description of Amendment Request The requested change is to delete the words "and contain a maximum total weight of 1619 grams uranium" from the Technical Specifications. This change is sought to account for weight differences in the Byron Unit 1 fuel versus the fuel supplied at Byron Unit 2 and Braidwood Units 1 and 2.

• Due to variations in the fuel manufacturing process and as new fuel designs are employed, slight differences in the maximum uranium weight per fuel rod have arisen. These differences may continue to arise in the futue when new reloads are used. This proposed license amendment addresses this possibility and prevents j the need to seek cycle by cycle Technical Specification changes.

Westinghouse typically specifies a maximum weight per fuel rod value. New fuel pellet designs have included higher as built densities in order to reduce pellet hydrogen content, as well as chamferred pellet designs with slightly higher pellet mass in order to enhance fuel performance. The practice of the fuel vendor has been to supply utilities with rod-wise weights instead of assembly

weights for accountability purposes at the manufacturing plant.

This value was meant to be descriptive and representative of the fuel loading and has not been used as a direct input to any safety analysis. No safety analyses are affected by the deletion of this value.

This change is also applicable to Braidwood Units 1 anil 2 and should be included in the draft Braidwood Technical Specifications.

! Basis for No Significant Hazards Determination i

j Commonwealth Edison has evaluated this proposed amendment and determined that it involves no significant hazards consideration. In accordance with the criteria of 10 CFR50.92(c) the proposed amendment does not:

(1) Involve a significant increase in the probability or i

consequences of an accident previously evaluated i

because:

(a) A deletion of the maximum weight of uranium per fuel rod does not increase the probability of a previously evaluated accident, or the i

consequences thereof.

1

The core design, and the safety analyses related to it, are based on parameters other than the weight of individual fuel rods (such as power and fuel dimensions). These parameters are either not directly affected by fuel rod weight, or are only slightly affected. A review of design parameters which may be affected indicates that a change in fuel weight does not cause other design parameters to exceed the values assumc3 in various safety analyses or to cause acceptance criteria to be exceeded. This is due to the fact that the safety analyses codes use fuel rod enrichment levels not weight values. Therefore, the effects of the change are within Technical Specification limits. Since deleting this value has no significant effect on safety analyses, it is concluded that the probability or consequences of an accident previously evaluated being significantly increased is unaltered.

(2) Create the possibility of a new or different kind of accident previously evaluated, because; (a) The change being made is to delete the maximum total weight of uranium per fuel rod from the Technical Specifications. The proposed amendment does not involve any new equipment, or the manner in which the system is being operated. Hence, the possibility of a new or different kind of accident being created than was previously evaluated is unaltered.

(b) The change is considered to be administrative in nature.

(3) Involve a significant reduction in the margin of safety, because:

(a) The margin of safety is maintained through adherence to other fuel related Technical Specification limits and the FSAR design bases. The deletion of fuel rod weight limits in the Technical Specifications does not directly affect any safety systems, thus the plant safety margin is unaffected.

Based on the preceeding assessment, Commonwealth Edison has determined that this proposed amendment involves no significant hazards consideration and request its approval in accordance with the provisions of 10CFR 50.91(a)(4).

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