ML20206T765
| ML20206T765 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 07/02/1986 |
| From: | Carey J DUQUESNE LIGHT CO. |
| To: | Tam P Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8607080222 | |
| Download: ML20206T765 (8) | |
Text
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'Af Telephone (412) 393-6000 Nuclear Group InSppiEsport. PA 15077-0004 Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Attn:
Mr. Peter S. Tam, Project Manager Project Directorate No. 2 Division of PWR Licensing - A Washington, DC 20555
- Mail Stop 340 -
Reference:
Beaver Valley Power Station, Unit No. 1 Docket No. 50-334, License No. DPR-66 Information Copy of Proposed License Change to Design Features Gentlemen:
Attached is an information copy of a proposed change to technical specification Design Feature Section 5.3.1 and applicable no significant hazards consideration.
The proposed amendment would revise this section of the technical specifications by removing the fuel rod weight limitation, the statement following which limits maximum enrichment for the initial core loading and provides for the insertion of reconstituted fuel assemblies into the core.
This information copy is provided so that you can begin your review as soon as possible since it is our understanding that NRC aproval of these changes is required prior to startup from the current refueling outage.
This is being interpreted to mean prior to entering MODE 2 which is scheduled to occur on August 11, 1986.
A Technical Specification change request will be forwarded to the NRC immediately following review by our various safety committees.
Very truly yours, J/
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- dent, Nuclear POR P
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Benvar Volley Powar Station, Unit No. 1 D6ckst No. 50-334, Lictnto No. DPR-66 Information Copy of Proposed License Change to Design Features Page 2 cc: Mr. W. M. Troskoski, Resident Inspector U. S. Nuclear Regulatory Commission Beaver Valley Power Station Shippingport, PA 15077 U. S. Nuclear Regulatory Commission c/o Document Management Branch Washington, DC 20555 Director, Safety Evaluation & Control Virginia Electric & Power Company P.O. Box 26666 One James River Plaza Richmond, VA 23261 Mr. Thomas M. Gerusky, Director Bureau of Radiation Protection Pennsylvania Dept. of Environmental Resources P.O. Box 2063 Harrisburg, PA 17120
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DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment building is designed and shall be maintained for a
maximam internal pressure of 45 psig and a
temperature of 280*F.
PENETRATIONS 5.2.3 Penetrations through the reactor ~ containment building are designed and shall bemaintained in accordance with the original design provisions contained in Section 5.2.4 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements.
5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 157 fuel assemblies with each fuel assemby containing 264 fuel rods, except for fuel assemblies which may be reconstituted to replace leaking fuel rods with non-fueled rods (e.g.,
zircaloy or stainless steel),
clad with Zircaloy
-4.
Each fuel rod shall have a nominal active fuel length of 144 inches.
Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 3.3 weight percent U-235.
CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 48 full length and no part length control rod assemblies.
The full length control rod assemblies shall contain a nominal 142 inches of absorber material.
The nominal values of absorber material shall be 80 percent slver, 15 percent indium and 5 percent cadmium.
All control rods shall be clad with stainless steel tubing.
BEAVER VALLEY - UNIT 1 5-4
ATTACHMENT B
i Proposed Technical Specification Change No. 128 No Significant Hazard Consideration J
Description of amendment request:
.The proposed amendment would revise the Design Features of Section 5.3.1 by removing the fuel rod weight limitation, the statement following which limits maximum enrichment for the initial core loading and provide for the insertion of reconstituted fuel assemblies into the core.
Basis for no significant hazard consideration:
,Due to fuel pellet design improvements such as chamfered pellets with reduced dish and a
nominal density increase, the fuel weight has increased i
slightly.
The actual uranium weight has no bearing on the power
- limits, power operation level or decay heat rate.
Although a number 4
of areas involving safety analysis are affected by fuel uranium
- weight, the areas of safety significance have their own limits which are reflected in the FSAR and Technical Specifications.
Technical i
Specifications on power and power distribution control the fission rate
- and, hence, the rate of decay heat production.
The composition of the fuel is closely' monitored to assure acceptable fuel performance for such things as thermal conductivity,
- swelling, j
densification, etc.
The important fuel parameters have been considered and are addressed in the following evaluation.
Seismic Effects on Fuel / Internals and New and Spent Fuel Storage Racks The fuel rod uranium weight as stated in the Technical Specifications is not a
direct input to.the analyses of maximum seismic /LOCA fuel assembly dynamic
- response, seismic response of reactor vessel and internals, or seismic analyses of raw and spent fuel storage racks.
Radiological Source Terms l
Fission product generation is not sensitive to the mass of fuel involved but to the power level.
As long as the power generated by the,cor'e is unaffected, there will be no significant impact on the radiological source terms.
t Fuel Handling Any postulated ' increase in the amount of uranium in the fuel rods i
would not have a significant impact on the fuel handling equipment.
The. spent fuel pit bridge and hoist is designed with a load limit of approximately twice the weight of a
nominal fuel assembly.
The manipulator crane is provided with two load sensors. One load sensor i
provides ' primary protection of the fuel assemblies from structural j
damage if an assembly were to
" hang-up".
A second load sensor provides backup protection against high lift force with a setpoint-above that of the first load sensor.
If the setpoints were unchanged l
despite a slight overall increase in uranium weight, the impact would' be to decrease the assembly weight and the lift force limit reduces the amount of stress the fuel assembly structure would be exposed to if the assembly were to " hang-up".
The manipulator crane margin to capacity limit far exceeds any potential increase in assembly weight due to increases in the fuel rod uranium weight.
t I
l Attachm3nt B Pcgn 2 j
LOCA Safety Analysis j
Uranium mass has no impact on ECCS LOCA analyses.
LOCA analyses are sensitive to parameters such as pellet diameter, pellet-clad gap, stack height shrinking factor and pellet density as they relate to l
pellet temperature and volumetric heat generation.
Fuel mass is not used in ECCS LOCA anlayses.
Non-LOCA Safety Analysis _
Individual fuel rod uranium weight, as reported in the Technical Specifications, is not explicitly modeled in any non-LOCA event.
Total uranium present in the core is input into the transient 3
j
- analyses, but is generated using a methodology independent of the l
value presented in the Technical Specifications.
Thus, any change in the number currently in the Technical Specifications does not impact the non-LOCA transient analyses.
Core Design The mass of uranium is explicitly accounted for in the standard fuel rod design through appropriate modeling of the fuel pellet geometry and initial fuel density.
Variations in uranium mass associated with
+
allowable as-built variations but within the specification' limits for i
the pellet dimensions and initial density are accounted for in the reactor core design analyses.
The Technical Specification uranium mass value has no impact on margin to reactor core design criteria.
i i
Fuel inspection performed during the fifth refueling outage found j
leaking fuel rods in a periferal assembly (G-15).
It was determined that the fuel rod leakage was attributable to baffle jetting.
The
,"~
solution to this
- problem, recommended by_ Westinghouse and used by other utilities, involves fuel assembly reconstitution as a means to allow the insertion of dummy rods into a fuel = assembly which is going i
to be installed in the core in a position where a known problem i
exists (i.e.,
baffle jetting).
In the reconstitution process, the fuel rods in positions subject _to problem conditions would be removed and replaced with chumny rods.
This would be accomplished by inverting the
- assembly, removing the bottom nozzle, inspecting and removing the failed
- rods, replacing the dummy rods containing no 1
- fuel, and replacing the fuel rods with bottom nozzle.
The reconstituted assembly will comply with the original assembly design l
criteria.
I Design Feature Section 5.3.1 describes the reactor core consisting of fuel assemblies containing 264 fuel rods.
In order to allow for the l
insertion of reconstituted assemblies into the
- core, this specification must be revised as follows:
The reactor core shall contain 157 fuel assemblies with each fuel assembly containing 264 fuel
- rods, except for fuel assemblies which may be reconstituted to replace fuel rods 4
j with non-fueled rods (e.g.,
zircaloy or stainless steel),
j clad with zircaloy -4."
i l
a.
Attachm:nt B Paga 3 The impact of the change is negligible, since the reconstituted assemblies will comply with the original assembly design criteria.
Basis for no significant hazards determination:
The proposed changes do not involve a significant hazards consideration because operation of Beaver Valley Power Station, Unit No. 1 in accordance with these changes would not:
i (1) Involve a
significant increase in the probability of occurrence or the consequences of an accident previously evaluated because:
i The variation in fuel rod weight that can occur even without a Technical Specification limit is small based on other fuel design constraints, e.g.,
rod diameter, gap size, UO-2 density and active fuel length; all of which provide some limit on the variation in-rod weight.
The current safety analyses are not based directly on fuel rod
- weight, but rather on design parameters such as power, and fuel dimensions.
These parameters are either (1) not affected at all by fuel rod weight, or (2) are 4
only slightly affected.
However, 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 assumed in the various safety analyses, or to cause acceptance criteria to be exceeded.
The effects are not significant with respect to measured nuclear parameters (power, power distribution, nuclear coefficients),
i.e.,
they remain within their Technical specification limits.
The statement limiting the initial core 4
maximum enrichment to 3.2 weight percent has been deleted since t
it only applied to the initial core loading and is no longer applicable.
- Thus, it is concluded that the Technical specification modification does not involve a
significant increase in the probability or consequences of a previously evaluated accident.
The reconstituted fuel assemblies meet essentially the same design requirements, satisfy the same design criteria as the original fuel assembly, and the use of assemblies will not result in a change to existing safety criteria and design lim'.ts.
(2) Create the possibility of a
new or different kind of accident from any accident previously evaluated because:
All of the fuel contained in the fuel rod is similar to and designed to function similar to previous fuel.
Thus, the existing new and spent fuel i
storage criticality analyses bound the changes observed.
This change is considered administrative in nature and does not create the possibility of a new or different kind of accident.
A single fuel assembly is moved at a time, and the consequences of an accident are bounded by the fuel hand.~~tng accident which is the most severe accident related to fuel manipulation.
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A,ttachm5nt B Paga 4 I
l (3) Involve a
significant reduction in the margin of safety because:
The margin of safety 'is maintained by adherence to other fuel related Technical Specification limits and the FSAR design bases.
I The margin of safety as defined in the technical-specification basis has not been reduced as demonstrated by the review of fuel assembly mechanical changes and since the existing safety criteria design limits will not be changed.
Therefore, removing the fuel rod weight limit and allowing the use of reconstituted assemblies in Design Features Section 5.3.1 q
does not directly affect any safety system or the safety limits,
)
and thus, do not affect the plant margin to safety.
i Conclusion-The proposed changes do not involve a significant increase in the i
probability or consequence of a previously evaluated accident, do not j
create the possibility of.a new or different kind of accident and do
]
not involve a significant reduction in a margin of safety.
1 Although a
number of safety analyses are affected indirectly by fuel weight, the analyses are more sensitive to fuel configuration,
- length, enrichment and physical design which are also specified in the plant Technical Specifications.
The Technical Specifications j
limit power and power distribution, thus controlling the fission rate i
and the rate of decay heat production.
Fuel rod weight does not have j
any direct bearing on the power limits, power operating level, or i
decay heat rate.
The composition of the fuel is closely monitored to assure acceptable fuel performance.
The fuel weight changes that could be made without a
Technical Specification ~1imit are not of l
sufficient magnitude to cause a
significant difference in fuel j
performance as analyzed by Westinghouse.
There are no expected observable changes in normal operation due to the noted fuel rod weight
- changes, and the remaining fuel parameters listed in the Technical Specifications are considered in the Reload Safety Evaluation.
Other Design Basis Events were examined to assess the effects of possible changes in fuel rod weight.
Fuel rod weight will only change as a result of a specific change in the physical design, which is addressed in the Reload Safety Evaluation, or within the manufacturing tolerances, in which ' case the changes in fuel rod i
weight are relatively insignificant.
Changes in nuclear design resulting 'from fuel rod weight changes are controlled as discussed 4
above.
For these
- changes, the effect. on new and spent fuel criticality and fuel handling analyses remain bounded by the existing analyses and Technical Specification Design Feature limits.
l Fuel-handling equipment and procedures are not affected by these l
weight
. changes.
Seismic /LOCA analyses contain sufficient I
Attachm nt B
- agm 5 conservatism to bound these weight changes.
Other accident analyses are not-affected by rod weight as a
direct parameter, and the existing analyses remain bounding.
The analyses show that the use of reconstituted assemblies will 1
comply with the original design criteria.
A reconstituted fuel assembly is not significantly different from the other fuel i
assemblies in the core.
There are no significant changes made to the acceptance criteria for the technical specifications and the analytical methods used.to demonstrate conformance with the technical specifications and regulations have not been significantly changed from those previously found acceptable to the NRC.
Therefore, based on the above, it is proposed to characterize the change as involving no significant hazards consideration.
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