ML20206J219
| ML20206J219 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 06/20/1986 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20206J203 | List: |
| References | |
| NUDOCS 8606270038 | |
| Download: ML20206J219 (12) | |
Text
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ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT DOCKET NOS. 50-327, -328 (TVA SQN TS 70)
PROPOSED CHANGE TO DELETE MAXIMUM WEIGHT OF URANIUM PER FUSL ROD FROM DESIGN FEATURES SECTION 5.3.I LIST OF EFFECTED PAGES UNITJ 5-4 UNIT 2 5-4 8606270038 860620 PDR ADOCK 05000327 i P PDR
5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 assembly containing 264 fuel rods clad with Zircaloy-4.T uel have
- ighta;fnominal active
'7": ;r::: fuel length of 144 inches.:nd :::t:!-Each fuel rod shall 2r:rir: : ::ximur t: ^. : t enrichment of 3.15 weight pe'rcent U-235.The initial core loading shall have a-maximum ,
of 4.0 weight percent U-235. physical design to the initial core loading um enrichment i;hq9 1 "
_ CONTROL ROD ASSEMBLIES 5.3.2 The reactor rod assemblies. core shall contain 53 full length and no partcontrol length 142 inches of absorber material.The full length control rod assemblies shall c shall be 80 percent silver, The nominal values of absorber material
' control rods shall be clad with stainless steel t bi15 percentAll u ng.
indium and 5 5.4 REACTOR COOLANT SYSTEM
_ DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall bee: maintain d a.
the FSAR, with allowance for normalo degradation n the
. of applicable Surveillsnce Requirements, b.
For a pressure of 2485 psig, and c.
ForF.a 680 temperature of 650*F, except for the pressurizer c s whi h i y0LUME 5.4.2 12,612 +The 100 total water. and steam volume of the reactor scoolant sy t cubic feet at a nominal T of 525 F. em is
,yg 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown ..
on Figure 5 L,A A V,
,,n ,A 1000 s ns v6 SEQUOYAH - UNIT 1 5-4 ' "'"
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DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 193 fuel assemblies with each fueT assembly containing 264 fuel rods clad with Zircaloy -4. Each fuel rod shall have a nominal active fuel length of 144 inches.:nd ::ntni : ::ximur tota! .::ight of 1755 gram: ur:rier The initial core loading shall have a maximum enrichment of 3.15 weight percent U-235. Reload fuel shall be similar l in physical design to the initial core loading and shall have a maximum jlne enrichment of 4.0 weight percent U-235. P; CONTROL ROD ASSEMBLIES 5.3.2 The reactor 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. The nominal values of absorber material shall be 80 percent silver, 15 percent indium and 5 percent cadmium.
All control rods shall be 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 12,612 + 100 cubic feet at a nominal T avg f 525 F.
5.5 METEOROLOGICAL TOWER LOCATION
, 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.
i EAA A gone
..- ,s iv vc SEQUOYAH - UNIT 2 5-4 ?= ^ n d _m. - ; '
ENCLOSURE 2 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT DOCKET NOS. 50-327, -328 (TVA SQN TS 70)
DESCRIPTION AND JUSTIFICATION FOR DELETION OF MAXIMUM WEIGHT OF URANIUM PER FUEL ROD FROM DESIGN FFATURES SECTION S.3.1 i
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4 I DESCRIPTION OF CHANGE This change will remove the maximum fuel rod uranium weight from Design Features Section 5.3.1, " Fuel Assemblies," for both Units 1 and 2 Technical Specifications. See attached marked-up pages (enclosure 1) for additional detail.
REASON FOR CHANGE Design Features Section 5.3.1, Fuel Assemblies, of the Technical Specifications, identifies a maximum total fuel rod weight of 1766 grams of uranium. Recent improvements to the fuel design (including chamfered pellets with a reduced dish and a nominal density increase) have increased fuel weight slightly. The weight increases may have caused the maximum fuel rod weight to exceed the specified maximum value of 1766 grams. This change will delete the specified maximum weight limit to allow the current fuel to be in compliance with Sequoyah Technical Specifications.
JUSTIFICATION FOR CHANGE The proposed change to Design Features Section 5.3.1. of the Sequoyah Technical Specification deletes the maximum fuel rod weight limit of 1766 grams of uranium. The purpose of the change is to permit the use of assemblies with fuel rods over the weight limit and also to reflect the relative insensitivities of this technical specification parameter in the safety analysis. It is judged that this weight difference does not have a
, significant impact on the safety analyses. Other technical specifications cover more important fuel-related parameters; therefore, deletion of the Design Features fuel rod weight limit is not significant to the safe operation of the plant.
. The proposed change of Technical Specification, Design Seatures Section 5.3.1, is given in enclosure 1. This is the only reference to fuel rod uranium weight in the technical specifications. In addition, the FSAR identifies a nominal core total weight of UO-2 (in pounds) for the initial (Cycle 1) core.
Although a number of safety analyses are effected 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 limit power and power distribution, thus controlling the fission rate and the rate of decay heat production. Fuel rod weight does not have any direct bearing on the power limits, power operating level, or decay heat rate. The composition of the fuel is closely monitored to ensure-acceptable fuel performance. The fuel weight changes that could be made without a technical specification limit are not of sufficient magnitude to cause a significant difference in fuel 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.
JUSTIFICATION FOR CHANGE Other design basis events were examined to assess the effects of possible changes in fuel rod weight. Fuel rod weight will only change as e. 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 weight are relatively insignificant. Changes in nuclear design resulting from fuel rod weight changes are controlled as l discussed 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. Fuel-handling equipment and procedures are not effected by these weight changes.
Seismic /LOCA analyses contain sufficient conservatism to bound these weight changes. Other accident analyses are not effected by rod weight as a direct parameter, and the existing analyses remain bounding.
In summary, the deletion of the maximum fuel rod weight limit in the technical specifications is proposed because the limit is not significant to the safe operation of the plant.
Westinghouse's " Nuclear Safety Evaluation Checklist" and " Safety Evaluation Justifying Continued Operation with Uranium Rod Weight Discrepancy" are attached as Attachments 1 and 2, respectively.
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ATTACllMENT 1 SE1 56-16 9 Customer Reference No(s).
Westinghouse Reference No(s). ~
(Change Control or RFQ As Applicable)
WESTINGHOUSE NUaEAR SAFETY EVALUATION CHECK LIST
- 1) W1 EAR PLANT (S) CCNr:IC
- 2) CHECK LIST APPLICABLE *ID: Fun Ron URANtuu wrynur (Subject of Change) _
3)
The :.-itten safety evaluation of the revised procedure, design change or modification required andrequired is attached. by 10CFR50.59 has been prepared to the extent incomplete fcr any reason, explain on Page ?.If a safety evaluation is not require Parts A and B of this Safety Evaluation Check List are to be completed only en the basis of the safety evaluation perfxced.
CHECK LIST - PART A (3.1) Yes (3.2) Yes _
No J., A change to the plant as described in the FSAR7 (3 3) Yes No 1 A change to procedures as described in the F3AR7 (3.4 ) Yes J.,No J A test or experiment not described in the FSAR7 No A change to the plant technical specifications (ALendix A to the Operating License)?
4)
CHECK LIST - PaBe PART 2.) B (Justification fr Part 3 answers must be inc (4.1)
Yes _ No L Will the probability of an accident previwsly (4.2) evaluated in the FSAR be increased?
Yes ,_ No 1 Will the consequences of an accident previously (4 3) evaluated in the FSAR be increased?
Yes _ No 1 May the possibility of an accident which is 1 different than any already evaluated in the FSAR (4.4) be created?
Yes No 1 Wul the probability of a malfunction of equipment important to safety previmsly evaluated in the (4.5) Yes FSAR be increased?
No 1 Will the consequences of a malfunction of equipnent important to safety previously evaluated (4.6) Yes in the FSAR be increased?
No 1 May the possibility of a malftnction of equi; ment important to safety different than any already (4.7) Yes evaluated in the FSAR be created?
. No L Will the margin of safety as defined in the bases to any technical specification be reduced?
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If the anwers to any of the above questions are mknown, indicated under
- 5) REPARKS and explain bels. l If the anwar to any of the above questions in 4) cannot be answered in the negative, based on written safety evaluatien, the change cannot be approved without an application for license amendment sutraitted to the NRC pursuant to 10CFR50.90 '
- 5) REPARKS:
None The follwin evaluation, gggummarizes the justificatien upon the written safety for answers given in Part B of the Safety Evaluation Check List:
See Attached Safety Evaluation II) Reference to document (s) containing written safety evaluation:
FDR FSAR UPDATE Section: Page(s): Table (s): Figure (s):
Reason for/ Description of Change:
Prepared by (Nuclear Safety): M
- _ Date: 88 Coordinated with Engineer (s): db /- Date: ths coordinating Group Manager (s):
YM_ _ Date: 2# (o Nuclear Safety Group Manager: u _
Date: 28
/
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ATTACHMENT 2 SAFETY EVALUATION JUSTI"YING CONTIWED OPERATION WITH URANIUM ROD WEIGHT DISCREPANCY The Design Features section of the Technical Specifications identifies a maximun total weight of uraniura in each fuel red.
Due to fuel pellet design improva::ents such as chamfered pellets with reduced increase, the fuel weight has increased slightly. The actual uraniun weight dish and a nominal density has no bearing on the power limits, power operating level or decay heat rate.
Although a number of areas truolving safety analysis are affacted by Nel uranium weight, the areas of safety significance have their own 11 tits which are reflected in the FSAR and Technical Specifications. Technical hence, the rate of decay heat procuction.Specificaticns on power and p The composition of the fuel is c1csely tenitored to assure acceptable fuel perfcrmance for such things as thermal concuctivity, swelling, densification, etc. The important fuel parameters have been considered and are addressed in the follcWing evaluation as pertaining to Westinghouse supplied com;onents and services. {
Sei ric Errem en Fuel /Teterr.nis mM New and Srset We1 Staraeo Raelen The fuel rod uranina weight as stated in the Technical Specifications is not a direct ir.put to the analyses of eaximun seismic /LOCA fuel assembly dynamic of new and spent fuel storage racks. response, seismic response of reactor ve Radielneien1 Secree Teces Fission the powerproduct level. generation is not sensitive to the mass of fuel involved but to there will be no significant impact on the radiological source terms met ham 1ine Any postulated increase in the amount of uraniun in the fuel rods would not !
i have a significant impact on the fuel handling equipnent. .
bridge of and hoist is Neldesigned with a lead limit of approximately twice theThe spent fuel pit weight a nomiral assembly. The manipuistor crane is provided with two load sensors.
One load sensor provides primary protection of the fuel assemblies from structural danage if an assembly were to " hang-up".A second 1 cad sensor provides backup protection against high lif t force with a setpoint above that of the first load sensor.
slight overall increase in uraniun weight, the impact wwld be to decrease potential fcr fuel damage since reducing the difference between the fuel assembly weight and the lift fcree limit reduces the amount of stress the fuel I assembly structure would be exposed to if the assembly were to " hang-up". The manip.tlator crane margin to capacity limit far exceeds any potential increase in assembly weight due to increases in the fuel red uraniun weight. l Page 1 of 2
-- - --,,------g- ,7m, y4-+,- w, -n---m-, - -4
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LOCA Safe +v Analysis i Uraniun mass has no impact on ECCS LOCA analyses. LOCA analyses are sensitive to parameters such as pellet diameter, pellet-clad gap, stack height shrir. king factor and pellet density as they relate to pellet temperature and vclumetric heat generation. Fuel mass is not used in ECCS LOCA analyses.
re-LOCA Suc tye Arm 1v mj_s Individual fuel rod uraniun weight, as reported in the Technical Specifications, is not explicitly modeled in any non-LOCA event. Total uraniun
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present in the core is input into the transient analyses, tut is generated using a methodology independent of the value presented in the Technical Specifications.
Thus, any change in the number currently in the Technical Specifications does not impact the non-LCCA transient analyses. ,
c m B . fen The mass of uraniun is explicitly accounted fcr in the standard fuel rod design
.through appropriate modeling of the ft.el pellet geometry and initial fbel l density. Variations in uraniun mass nssociated with allcwable as-built 1 variations tut within the specification limits fcr the pellet dimensions and initial density are accounted fcr in the reactor core design analyses. The ;
Technical Specification uraniun mass value has no impact on margin to reactor core design criteria.
l The conclusion of these evaluations is that there is no snreviewed safety '
question associated with operation of the tnit(s) with a fuel rod weight in l excess of that defined in Section 5.31 of the Technical Specifications.
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ENCLOSURE 3 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT DOCKET NOS. 50-327, -328 (TVA SQN TS 70)
DETERMINATION OF NO LIGNIFICANT HAZARDS CONSIDERATIONS FOR DELETION OF MAXIMUM WEIGHT OF URANIUM PER FUEL ROD FROM DESIGN FEATURES SECTION 5.3.1 a
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BASIS FOR NO SIGNIFICANT HAZARDS DETERMINATION i
The proposed amendment discussed above shall be deemed to involve a significant hazards consideration if there is a positive finding in any of the following areas.
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- 1. Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
4 Response: No 1
The deletion of the fuel rod uranium weight limit does not significantly increase the probability or consequences of previously evaluated accidents.
! 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, U0-2 density and active fuel length; all of wh'ich 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 effected at all by fuel rod weight, or (2) are only slightly effected. However, a review of design parameters which may be effected 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 coetficients), i.e., they remain within
! their technical specification limits. Thus, it is concluded that the j technical specification modification does not involve a significant increase i
1 in the probability or consequences of a previously evaluated accident.
- 2. Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
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, Response: No i
j The creation of a new, different kind of accident from any previously evaluated accident is not considered a possibility. All of the fuel contained 4
in the fuel rod is similar to and designed to function similar to previous fuel.
l Thus, the existing new and spent fuel storage criticality analyses j
bound the changes observed. This change is considered as administrative in nature and does not create the possibility of a new or different kind of
, accident.
3.
i Will operation of the facility in accordance with the proposed change involve a reduction in margin of safety.
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Response: No The margin of safety is maintained by adherence to other fuel related technical specification limits and the FSAR design bases. The deletion of j fuel rod weight limits in the technical specification Design Features Section S.3.1 does not directly effect any safety system or the safety limits, thus, l
I not effecting the plant margin to safety. i l
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