ML17324B182

Proposed Tech Specs Re Reactor Core Fuel Assembly & Control Rod Assembly Design Features

ML17324B182
Person / Time
Site:	Cook
Issue date:	12/22/1986
From:	INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:
Shared Package
ML17324B181	List: ML17324B180 ML17324B182
References
NUDOCS 8612300264
Download: ML17324B182 (22)

Text

DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment building is designed and shall be maintained in accordance with the original design provisions contained in Section 5.2.2 of the FSAR.

PENETRATIONS 5.2.3 Penetrations through the reactor containment building are designed and shall be maintained in accordance with the original design provisions contained in Section 5.4 of the FSAR with allowance for'ormal degradation pursuant to the applicable Surveillance Requirements.

5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 193 fuel assemblies with each fuel assembly containing 204 fuel rods clad with Zircaloy -4. Each fuel rod shall have a nominal active fuel length of 144 inches. The initial core loading shall have a maximum enrichment of 3.35 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 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:

D. C. COOK - UNIT 1 5-4 Amendment No.

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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 fuel assembly containing 264 fuel rods clad with Zircaloy -4. Each fuel rod shall have a nominal active fuel length of 144 inches. The initial core loading shall have a maximum enrichment of 3.3 weight percent U-235.

Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 3.84 weight percent U-235.

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 4.1.6 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 0 F, except for the pressurizer which is 680 F.

D. C. COOK - UNIT 2 5-4 Amendment No.

Attachment 3 to AEP:NRC:1016 Westinghouse Electric Corporation Letter Dated September 4, 1986

I SEP 1 1 )986 Nuclear Technology Oivision Westfnghouse Water Reactor Electrfc Corporation . Divisions Box 355 Pittsburgh Pennsylvania 15230 0355 AEP-86-676 September 4, 1986 NS-OPLS-OPL-II-86-184 Mr. M. P. Ai xich, Vice President "

and Director Nuclear Operations American Electr'c Power Service Corporation One Riverside Plaza Columbus, Ohio 4~?16 AMERICAN ELECTRIC POWER SERVICE CORPORATION D. C. COOK UNIT 1

,SUPPORT NG DOCUMENTATION FOR TECHNICAL SPECIFICATION DESIGN FEATURE FUEL ASSr~LY DELETION AND JUSTIFICATION FOR CO> !I!iLM OPERAT ON SAFETY EVArLUATION

Dear Mr. Alexich:

'Ihe purpose of this letter is to confirm our telephone conversation informing you of an issue relating to the Technical Specifications on your Unit No. and as 1 manufactur ed fuel character istics.

Westinghouse units employing Standardized Technical Specification fomat defined in Section 5.3.1 a maximum uranium weight per rod value. It has come to WestinEhouse's attention that as-manufactured fuel rod uranium weight have exceeded this value for fuel which has been shipped to your plant site and is scheduled for future operation in an upcaaing cycle or is in your current operating unit/cycle.

The subject technical specification value a's stated in Section 5.3. was intended 1 to be descriptive and representative of the fuel loading and has not been used as a direct input to any safety analysis. It is judged that the weight difference rod uranium weight may exceed the specified maximum uranium weight 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 safe operation of the plant.

Westinghouse has provided a generic safety evaluation (see Attachment A) which may be applied to your unit in support of continued operation pending modification of tech specs for those plants which are known to have exceeded their specified fuel rod maximum uranium weight. Attachments B and C provide a Proposed Amendment and Basis for Ho Significant Hazards Determination in order Uo delete the tech spec fuel rod maximum uranium weight as specified in Section 5.3.1. The proposed change of Technical Specification Design Features Section 5.3. 1 is provided in Attachment This is the only reference to fuel rod uranium weight in the technical specifications.

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All Westinghouse fuel supplied has been manufactured within all appropriate specifications and r elated manufacturing controls. 'Ihe Westi'nghouse Nuclear Fuel Division has been manufacturing fuel pellets with higher as built densities in order to reduce pellet hydrogen content and for some applications a chamferred pellet design with slightly higher pellet mass in order to enhance fuel per formance. It is only during the last three years that rod-wise uranium weights instead of assembly weights have been supplied for Special Nuclear Material (SNM) accountability purposes due to more detailed accountability procedures at the manufacturing plant and not for the intent of confirming this tech spec. lt is, however, by this accountability procedure that this issue has come to light. 'Ihe data flow for these SNM considerations will be unaffected by this amendment.

Westinghouse is reviewing all available manufacturing data to assure that all related cases of rod uranium weight are covered.

Westinghouse Nuclear Safety and Nuclear Fuel personnel will be available to answer any questions of this issue and to support your plant operation in any NRC r elated discussion or submittal required.

If you have any questions or comments, please contact the undersigned.

Very truly yours, p- c.

H. C. Walls, Project Manager Projects Department F. Scapellato/dmr cc: M. P. Alexich J. G. Feinstein V. Vander Burg J. Harkowsky S. H. Steinhar t D. R. Hafer J. R. Jensen R. W. Jurgensen W. G. Smith B. Svensson M. J. Parvin, W

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ATTACHMEÃr A SAr=TY EVALUATION JUSTIFYING CONTINUED OPERATION WITH URANIUM ROD WEIGHT DISCREPANCY

SECL 86-169 Customer Reference No(s).

Mestinghouse Reference No(s).

(Change Control or RFQ As Applicable)

WESTINGHOUSE NUCLEAR SAFETY EVALUATION CHECK LIST

1) MJCLEAR R.ANT(S)

2) CHECK LIST APPLICABLE '10:

(Subject of Change)

3) The written safety evaluation of the revised procedure, design change or modification required by 10CFR50.59 has been prepared to the extent required and is attachea. If a safety evaluation is not required or is incomplete fcr any reason, explain on Page 2.

Parts A and B of this Safety Evaluation Check List are to be completed only on the basis of the safety evaluation perfcrmed.

CHECK LIST - PART A (3.1)

(3.2)

Yes No ~

~ A change to the plant as described in the FSAR2-to procedures as described in the FSAR'?

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Yes No A change (3,3) Yes No A test or experiment rat described in the FSAR'?

(3.4) Yes No A change to the plant technical specifications (Appendix A to the Operating License)2

4) CHECK LIST - PART B (Justification fa Part B angers must be irx:luded on Page 2.)

(4.1) Yes No ~ Mill probability the in of an accident previously (4.2) Yes No ~ Mill in evaluated the FSAR be irx:reased2 the consequences of an accident previously (4.3) Yes No ~ different~ssibility of evaluated May the the FSAR be irx:reased2 an accident which than any already evaluated in is the FSAR (4.4) Yes No ~ Mill probability be created2 the important to of malfunction of a

safety previously evaluated equipnent in the (4.5) Yes No ~ Will FSAR be irx:reased2 the consequences equipnent important to of malfmction of a

safety previously evaluated in Yes

'4.6)

No ~ the FSAR be irx:reased2 May the important

~ssibility of malfmction of to safety a

different equipnent than any already in No ~ Mill evaluated to the margin any the FSAR be created'?

qf safety technical specification as defirad in be reduced2 the bases Page 1 of 4

If the answers to any of the above questions are unknown, indicated under

5) RENRKS and explain bel~.

If the answer to any of the above questions in 4) cannot be answered in the negative, based on written safety evaluation, the change canret be approved without an application fcr license amendment sutmitted to the NRC. pursuant to 10CFR50.90

5) REMARKS:

None The follcwinf1)ummarizes the justification upon the written safety evaluation, for answers given in Part 8 of the Safety Evaluation Check List:

See Attached Safety Evaluation (1) Reference to document(s) containing written safety evaluation:

FOR FSAR UFDATE Section: Page(s): Table(s): Figure (a ):

Reason for/Description of Change:

Prepared by (Nuclear Safety): . +zW s<

Coordinated with Engiraer(s): Date: ~>>

Coordinating Group Manager(s): Date: <><

Nuclear Safety Group Manager: Date: ~>>>> /~

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SAFETY EVALUATION JUSTIFYING CONTIMJED OPERATION WITH URANIUM ROD WEIGHT DISCREPANCY The Design Features section of the Technical Specifications identifies a maximun total weight of uraniun in each fuel rod. Due to fuel pellet design improvements such as chamfered pellets with reduced dish and a nomirsl density increase, the fuel ~eight has increased slightly. The actual uranian weight has no bearing on the po~er limits, power operating level or decay heat rate.

Although a rabber of areas involving safety analysis are affected by fuel uraniun weight, the areas of safety significance have their mn limits which are reflected in the FSAR and Technical Specifications. Technical Specifications on gower and power distribution control the fission rate and, hence, the rate of decay heat production. The com~sition of the fuel is closely monitored to assure acceptable fuel perfcrmance fa such things as thermal conductivity, melling, densification, etc. The im~rtant fuel parameters have been considered and are addressed in the fall<wing evaluation as pertaining to Westinghouse supplied components and services.

The fuel rod uraniun weight as stated in the Technical Specifications is not a direct input to the analyses of maximun seismic/LOCA fuel assembly dynamic response, seismic response of reactor vessel and internals, or seismic analyses of new and spent fuel storage racks.

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 core is unaffected, there wQ1 be no significant impact on the radiological source terms.

Any postulated irerease in the ancvnt of uraniun in the fuel rods would not have a significant impact on the fuel handling equipnent. The spent fuel pit bridge and hoist is designed with a iced limit of approximately twice the weight of a remiral fuel assembly. The manipulator crane is provided with two iced sensors. Ore lead sensor provides primary protection of the fuel assemblies from structural damage if an assembly were to "hang-up>>. h second 1ced sensor provides backup protection against high above that of the first lead sensor. If lift force with a setpoint the setpoints were mchanged despite a slight overall irarease in uraniun weight, the impact would be to decrease the potential fcr fuel damage sirae reducing the difference between the fuel assembly weight and the lift fcrce limit, reduces the amount of stress the fuel assembly structure would be exposed to if the assembly were to <<hang-up". The manipQator crane margin to capacity limit far exceeds any potential irerease in assembly ~eight due to irareases in the fuel rod uraniun weight.

Page 3 of 4

Uranjgn mass has no impact on ECCS LOCA analyses. LOCA analyses are sensitive to paraneters such as pellet diameter, pellet-clad gap, stack height shrinking factor and pellet density as they relate to pellet temperature and volunetr ic heat generation. Fuel aass is not used in ECCS LOCA analyses.

Individual fUel rod uraniun weight, as re~rted in the Technical Specifications, is not explicitly modeled in any non-LOCA'vent. Total uraniun present in the core is input into the transient analyses, but is generated using a methodology independent of the value presented in the Technical Specifications. Thus, any change in the rember currently in the Technical Specifications does not impact the non-LOCA transient analyses.

The mass of uraniun is'explicitly accounted fcr in the standard fuel rod design

.through appropriate modeling of the fuel pellet geometry and initial fUel density. Variations in uranim mass associated with alliable as-built variations but within the specification limits for the pellet dimensions and initial density are accounted for in the reactor core design analyses. The Technical Specification uraniun mass value has no impact on margin to reactor core design criteria.

The conclusion of these evaluations is that there is no unreviewed safety question associated with operation of the wit(s) with a fuel rod weight in excess of that defirad in Section 5.3.1 of the Technical Specifications.

Page 4 of 4

ATTACHMENT B PROPOSED AMENDMENT

PROPOSED AMENDMENT REASON FOR CHANl'E Design Features Section 5.3.1, Fuel Assemblies, of the Technical Specifications, identifies a maximan total fuel rod weight of 2236 g f I . h f ld lg,flam chamfered pellets with a reduced dish and a naninal density increase), have increased fuel weight slightly. '?he weight increases have caused the maximum fuel rod weight to exceed the specified maximun value of ?2. 6 gram limit. &is change will delete the specified maximum weig t t to allow the current fuel to be in canpliance with the D. C. Cook Unit 1 Specifications (see the attached marked-up spec cation .

SAFETY/ENVIRONMENTAL EVALUATION Senary of Change of h~i The proposed 2236 change I&id fl grams to I l<<h Design Features Section of uranian.

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5.3,1 I M of'he D. C. Cook Unit 1 igh IBZ The purpose of the change is to permit igh 1~ dN I fl the relative insensitivities of this technical specification parameter in the safety analysis. It is fudged 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 planta Evaluation Ne proposed change of Technical Specification Design Features Section 5.3.1 is given in Attachment D. This is the only reference to fuel rod uraniun 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 affected indir ectly by fuel weight, the analyses are more sensitive to fuel configuration, length, cnr ichnent 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 canposition of the fuel is closely monitored to assure acceptable fuel perfonnance. 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 nodal operation due to the noted fuel rod weight changes, and the r enaining fuel parameters listed in the Technical Specifications are considered in the Reload Safety Evaluation.

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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 weight are relatively insignificant. Changes in nuclear design resulting fr an fuel rod we1ght changes ar e controlled as discussed above, For these changes, the effect on new and spent fuel criticality and fuel handling analyses r emain bounded by the existing analyses and Technical Specification Design Feature 11mits. Fuel-handling equipnent and procedures are not affected by these weight changes. Seismic/LOCA analyses contain sufficient 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.

Conclusion In sawary, the deletion of the maximun fuel rod weight limit in the Techn1cal Specifications is proposed because the limit is not significant to the safe operation of the plant.

ATTACHMENT C BASIS FOR NO SIGNIFICANT HAZARDS DETERMINATION

PASIS FOR NO SIGNIFICANT HAZARDS DETERMINATION The proposed aaendment discussed above shall be deemed to involve a significant hazards areas consideration if there is a positive finding in any of the following o Mill operation of the facility in accordance with thi involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The delet1on of the fuel rod uranium weight lind.t 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 dimeter, gap size, UO-2 density and active fuel length; all of which provide sane limit on the variation in rod might. 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 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 assigned in the various safety analyses, or to cause acceptance criteria to be exceeded..

The effects are not s1gnificant with r espect to measured nuclear parameters (power, power distribution, nuclear coefficients), i.e., they remain within-their Technical Specification limits. Thus, it 1s concluded that the Technical Specification modification does not involve a significant increase in the probability or consequences of a previously evaluated accident.

2. Mill operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident fran any accident previously evaluated?

Response: No The creation of a new, different kind of accident fran any previously evaluated accident is not considered a possibility. 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 storage criticality analyses bound the changes observed. This change is considered as administrative in nature and does not create the possiblity of a new or different kind of accident.

3. Mill operation of the facility in accordance with the proposed change involve a reduction in a margin of safety.

Response: No We margin of. safety is maintained by adherence to other fuel related Technical Specification limits and the FSAR des1gn bases. ~ deletion of fuel rod weight limits in the Technical Specifications Design Features Section 5 3.1 does not directly affect any safety system or the safety limits, thus, not affecting the plant margin to safety.

ATTACHMENT 0 MARKS-UP DESiGN FEATURE FUEL ASSEMBLY SPECIFICATION

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to AEP:NRC:1016 Exxon Nuclear- Company Letter Dated October 15, 1986

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.OCT 2 0 1986 E)/ON NUCLEAR COMPANY, INC.

600 108TH AVENUE NE. PO BOX 90777. BELLEVUE, WA 98009%777 (906) 453<<LX10 October l5 l 986 ENC-AEP/0529 Indiana 8 Michigan Electric Company c/o Richard B. Bennett Engineer, Nuclear Materials 8 Fuel Management American Electric Power Service Corp, One Riverside Plaza, 20th Floor Columbus, OH 432 I 5

Dear Rich:

Re: Letter, R.B. Bennett (AEP) to H.C. Shaw (ENC), "Removal of Rod Weight Reference in Technical Specification 5.3. I," dated September l7, 1986 (AEP-ENC/0259)

In response to your request contained in the above re ference, the proposed D.C. Cook Unit 2 Technical Specification change has been reviewed. Also reviewed was the documentation attached to the above reference detailing Westinghouse's generic evaluation for this Technical Specification change and its application to D.C. Cook Unit I.

Exxon Nuclear concurs with the generic Westinghouse evaluation of the proposed removal of the maximum rod weight from Technical Specification 5.3.I. Review of the Exxon Nuclear analyses performed in support of D.C. Cook Unit 2 indicates that this maximum Technical Specification rod weight was not used in any of the calculations. Thus, its removal from the Technical Specification will not affect the results of any of the safety analyses performed by Exxon Nuclear.

If you have any questions regarding the above review, please feel free to contact our Mr. Jerry Holm, telephone 509-375-8I42.

Very truly yours,

. C. haw Contract Administrator pkc c<< M. P. Alexich J. M. Cleveland D. H. Malin V. VanderBurg J. S. Holm (ENC)

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