ML20078G639

Safety Evaluation Supporting Amend 21 to License DPR-79

ML20078G639
Person / Time
Site:	Sequoyah
Issue date:	09/29/1983
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20078G633	List: ML20078G639
References
NUDOCS 8310130043
Download: ML20078G639 (9)

Text

-

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMEhDMENT N0.21 TO FACILITY OPERATING LICENSE DPR-79 TENNESSEE VALLEY AUTHORITY INTRODUCTION TVA requested in their letters of July 1 and July 27, 1983, several changes to the Technical Specifications for Sequoyah Units 1 & 2.

Additional information on the requested changes was provided in letters of August 3, September 7, and September 19, 1983. Also the reference section in the fuel reload descussion provides a comprehensive listing of material that was utilized in the analysis.

One change for Unit 2 is to accommodate cycle 2 fuel reload operations. For this reload, sixty-eight new fuel assemblies will replace spent fuel from the first cycle. The new assemblies are the same as the assemblies in place, except for minor grid modifications to ninimize interactions of grid spacing Juring fuel handling. Also some new burnable absorber rods will be utilized in cycle 2 that have been previously accepted for use in other nuclear plants. Also, the TVA request of July 1, 1983, requested a number of technical specification changes to improve plant operations which are applicable to both Units. These are removing operating restrictions on control rod operations, and adding require-ments on the hydrogen control system.

Technical Specification changes regarding the testing of containment protective fuses from a destructive type of testing to visual inspection was requested.

Every 18 months,10% of the protective fuses are to be tested to ensure their integrity. At Sequoyah there are three types of protective fuses: 6900 and 480 volt fuses crimped inline and 480 volt fuses located in clip type holders.

Removal of the fuses for testing may compromise cable and holder integrity.

FUEL RELOAD By letter dated July 1,1983, the Tennessee Valley Authority (TVA), licensee for Sequoyah Nuclear Plant Unit 2, submitted a request (Ref.1) for a change in the plant Technical Specifications to accommodate the Unit 2, Cycle 2 reload.

The submittal included a reload safety evaluation (RSE) that contained a description of the changes, a justification for the changes, and the proposed Technical Specifications. As stated in the RSE, the reload analysis was accom-plished utilizing the methodoloav described in WCAP-9273, " Westinghouse Reload SafetyEvaluationMethodology"(Ref.2).

8310130043 830929 PDR ADOCK 05000328 P

ppg

. FUEL SYSTEM DESIGN The objectives of the fuel system safety review are to provide assurance that (a) the fuel system is not damaged as a result of normal operation and antici-pated operational occurrences, (b) fuel system damage is never so severe as to prevent control rod insertion when required, (c) the number of fuel rod failures is not under-estimated for postulated accidents, and (d) coolability is always maintained. Our evaluation of the information provided in support of Cycle 2 operation of Sequoyah Unit 2 is described below with regard to these review objectives.

Description The Sequoyah 2 Cycle 2 core will be comprised of 193 fuel assemblies manufac-tured by Westinghouse Electric Corporation (W). Sequoyah 2 is one of the first W plants to utilize a W 17X17 fuel assimbly design. This design is intended to extend fuel capaEility beyond that of the earlier 15X15 design in other W reactors of this approximate size. The primary intent of the design is to reduce stored energy in fuel rods for LOCA conditions.

For Cycle 2 operation, 68 Region 1 fuel assemblies will be replaced with 68 Region 4 (Reload 1) assemblies.

The mechanical design of the Region 4 assemblies is the same as the Region 1 assemblies with the exception of a reconstitutable bottom nozzle design and grid modifications intended to minimize potential grid-to-grid interaction during fuel-handling. Table I provides a comparison of pertinent design parameters.

In the RSE (Ref.1), it is stated that a new Wet Annular Burnable Absorber (WABA) rod design will be utilized for Cycle 2 operation. The WABA design is described in WCAP-10021, Revision 1 (Ref. 3).

Design Evaluation WABA - The WABA rod design consists of annular pellets of aluminum oxide and boron carbide (AL 0 B C) burnable absorber material encapsulated within two 3 3g cor. centric Zircalby tDbings. The reactor coolant flows inside the inner tubing and outside the outer tubing of the annular rod. The topical report describingtheWABAdesign(Ref.3)wasapproved,andtheutilizationofWABA rods in Sequoyah 2 is thus approved subject to certain conditions. These con-ditions concern surveillance and the analysis of core bypass flow.

I

. With regard to surveillance of the WABA rods, Westinghouse proposed a program that will consist of a visual (binocular) examination of approximately 10 percent of the WABA assemblies for evidence of anonelies or loss of struc-tural integrity of the rodlets.

In addition, the surveillance program would include review of routine in-core instrumentation measurements taken during the cycle. Such measurements would be used to help monitor the reactivity worth of the WABAs for comparison with predictions and with the results of the visual examination. Westinghouse offered to perform this surveillance for the first two plants to utilize WABAs.

In response to a staff question (Ref. 4) con-cerning whether Sequoyah Unit 2 is subject to and will follow the provisions of the proposed W surveillance program, TVA stated (Ref. 5) that the program will indeed be carried out at Sequoyah Unit 2, although it was not originally intended to be one of the two WABA surveillance plants.

TABLE 1 FUEL ASSEMBLY DESIGN PARAMETERS SEQUOYAH UNIT 2 - CYCLE 2 REGION 1

2 3

4 Enrichment (W/0 U-235)*

2.124 2.617 3.101 3.50 Gecmetric Density 94.6 94.7 94.6 94.5 (PercentTheoretical)*

Numoer of Assemblies 5

72 48 68 Approximate Burnup at 14,400 16,400 10,500 0

Beginning of Cycle 2 (MWD /MTU)

Approximate Burnup 24,100 27,400 23,100 11,500 Predicted for E0C 2

• All fuel regions except region four are as-built values: Region four values are nominal. An average density of 94.5 percent theoretical was used for Region 4 evaluations.

. Miscellaneous Design Considerations - In addition to surveillance of the WABA rods, additional information was requested (Ref. 4) concerning a few miscel-laneous items that were not fully addressed in the RSE (Ref. 1). The questions and responses (Ref. 5) are addressed below.

Because the RSE presented only beginning-of-cycle (B0C) 2 burnup values for each fuel region, but not predicted (approximate) end-of-cycle (E0C) 2 burnups, our first question requested the E0C target burnups. The previously missing burnup values are shown in Table 1.

They are consistent with second cycle burnups in other W plants and require no further comment.

We also requested further information (Ref. 4) about the reconstitutable bottom nozzles and spacer grid modifications, which are new features in the Region 4 fuel.

In response (Ref. 5) TVA indicated that the reconstitutable bottom nozzle feature is the same as that introduced on other W plants such as Trojan, Farley Units 1 and 2, Salem Unit 1, North Anna Units 1 and 2, and (very recently)

Beaver Valley Unit 1, and that this feature and the spacer grid modifications (which have also been implemented in fuel assemblies in the fore-mentioned plants) were evaluated by the licensee and determined not to involve an unreviewed safety question or Technical Specification change. Therefore, these design changes were performed under the provisions of 10 CFR 50.59. The Region 4 modified design features thus appear to be within the state-of-the-art and to be relatively minor. They are acceptable, therefore.

Another staff question concerned the fuel performance model (Ref. 6) used for the Cycle 2 analysis. That model (PAD 3.3) had been approved subject to cer-tain restrictions.

In response, TVA indicated that the approved W fuel perfor-mance model, with restrictions as modified in the NRC's safety evaluation for Addendum 1 to reference 6, was used in the design of Region 4 fuci.

Inasmuch as the fuel performance analyses were performed with an approved model, it is acceptable without further review.

The final staff question (Ref. 4) concerned W's fuel rod revised internal pres-sure design basis. Although the RSE indicated that the internal pressure design basis, as described in WCAP 8964 (Ref. 7), was satisfied for Cycle 2 operation, we pointed out to TVA that an amended criterion had to be used to assure accept-able consequences for transients and accidents.

In response (Ref. 5) TVA stated that the amended rod internal pressure criterion, was specified for Sequoyah Unit 2 Cycle 2 operation. We accept that statement without further review.

Non-Fuel Bearing Component (NFBC) Holddown S3rincs - During a recent refueling of McGuire Unit 1, several broken NFBC loldcown springs were dis-covered (Ref. 8). McGuire 1, like Sequoyah 2, is an upper-head-injection (UHI) W plant. Such plants have holddown assemblies for non-fuel-bearing com-ponents such as thimble plugs and secondary sources. The primary function of the holddcwn assemblies is to provide an axial force on the NFBCs sufficient to oppose flow-induced lift forces during reactor operation.

For the UHI plants,

4 the holddown assemblies function also as part of the injection system, and so the springs in there assemblies are of special design with central turns of larger diameter than the end turns to allow radial flow area for emergency coolant. Because broken springs could have an impact on UHI flow, with resul-tant effects on the LOCA peak cladding temperatures, and because broken springs could also result in fuel damage from loose parts, TVA was asked (Ref. 4) to provide reasonable assurance for Cycle 2 operation that broken springs would not occur or that the potential effects of loose parts, UHI flow restrictions, and increases in LOCA peak cladding temperature would not be significant.

In response (Ref. 5) TVA stated that, although 17 broken springs were identified via examinations performed during refueling, there were no double-ended breaks.

This observation was important because double-ended breaks would have greater potential for UHI blockage or loose parts. With regard to Cycle 2 operation, TVA stated (Ref. 5) that all 94 NFBC holddown springs will be replaced with springs of a new design in which mean stress levels are reduced and material grain size is reduced for better fatigue properties.

Inasmuch as all the original NFBC springs will be replaced with the new design, which has been reviewed and approved for McGuire, we conclude that there is reasonable assurance that NFBC holddown springs will not be a problem during Cycle 2 operation of Sequoyah 2.

NUCLEAR DESIGN The Cycle 2 loading is designed to meet an F (z)XP ECCS analysis limit of n

42.237XK(z). The kinetics characteristics far Cycle 2 are identical with those iTsed for the previously submitted accident analysis except for the most nega-tive Doppler temperature coefficient. The effect of this difference was con-sidered in the analysis. The PALAD0N code which has been reviewed and approved by the staff was used to perform the nuclear design analyses. Control require-ments analyses show that adequate shutdown margin exists for Cycle 2.

The con-trol rod insertion limits remain unchanged from Cycle 1.

The Cycle 2 analysis was performed with the following changes to the power dis-tribution control procedure:

N (1) The partial power multiplier for F was changed from 0.2 to 0.3.

g (2) The constant axial offset control procedure was replaced with the relaxed axial offset control (RA0C) procedure.

(3) The presently used surveillance procedure for F (z) was replaced 9

with procedure.

These changes were requested in order to increase operational flexibility and permit optimization of the loading pattern.

The change in the partial power multiplier f r F"N allows increased F N

at N

reducegpowerwhilemaintagningthesameFs1.55[l+0.3(lb) lim at full power. The allow-

].

The increase in allowable able F is described by F g

g

6-F"$isfy the peaking factor criteria at low power with control rod bank at at low power eliminates the need to change the rod insertion limits to se insertion limit. The effect of this change was considered in the Cycle 2 analysis.

The RA0C procedure is described in a E topical report (Ref. 9). This report has been reviewed by the staff which concluded that the RA0C procedure was an acceptable method for power distribution control in W reactors.

The revised Fn(z) surveillance is described in a W topical report (Ref. 9)

This report hXs been reviewed by the staff which concluded that the revised technique accomplished the same ends as the present one and is acceptable.

Control Rod Operations The licensee requested removal of the interim operating restrictions imposed as a result of deficiencies in the analysis of the control rod drop event.

Westinghouse submitted a topical report supporting the removal of these restric-tions when certain analyses are performed. The staff has approved this report and the analyses have been performed for Sequoyah 2 Cycle 2.

Thus the interim operating restrictions may be removed.

THERMAL HYDRAULIC Cycle 2 reload fuel has no significant variation from Cycle 1 that would affect the thermal margins. However,thelfcenseehasproposedthatthecoefficient of the power dependent term in the F4 equation of the Technical Specification will be increased to 0.3 from its present value of 0.2.

This has the effect of increasing the allowed radial peaking factor at low powers. This change was also accepted and incorporated in Unit 1 Cycle 2 reload.

Sequoyah Unit 2 Cycle 2 reload will ccntain 288 new WABA rods. This number is well within the maximum numbers of WABA rods allowed in reloads as specified in Table 7.2 of reference 3.

Based on the above and since the proposed Technical Specification change does not result in violation of SAFDL, and the number of WABA rods is well within the acceptabic configuration, we conclude that the proposed Lycle 2 operation is acceptable.

ACCIDENTS AND TRANSIENTS The effects of the reload on the design basis and postulated incidents analyzed in the FSAR for four loop operation have been examined.

In most cases, it was found that the effects can be accommodated within the conservatism of the initial assumptions used in the previous applicable safety analysis.

For the incidents which were reanalyzed it was determined that the applicable design basis limits are not exceeded and thus the conclusions presented in the FSAR are still valid.

,.m s-

. The only kinetic parameter not within the limiting range of values used in the previous safety analysis was the Doppler temperature coefficient (DTC). The change was small and since the DTC represents only a small portion of the total negative reactivity feedback, the effect is negligible and no accidents were reanalyzed.

Cycle 2 has a trip reactivity insertion rate which is different from that used for Cycle 1.

Investigation of the transients showed that only the locked rotor and loss of flow analyses may be affected. These transients were reanalyzed and there were no changes to the NRC safety conclusions.

Peaking factor evaluation for the rod out of position and hypothetical steamline break accidents resulted in a minimum DNBR greater than the design limit DNBR.

Thus, these accidents were not investigated further.

The hot-zero power beginning of life rod ejection accident was reanalyzed because the Cycle 2 maximum F exceeded the Cycle 1 values. The results show n

that all acceptance criteria specified in reference 10 were satisfied. Thus, the safety conclusions remain valid.

The change in the allowable F" as function of power resulted in a change to the k constants in the overteherature Delta-T and overpower Delta-T setpoint equations and a change to the overtemperature Delta-T f(AI) function. Since the overtemperature Delta-T trip is used in the bank withdrawal at power acci-dent, this accident was reanalyzed with the new overtemperature Delta-T set-points. The results show that the minimum DNBR remains above the limit, and thus, are acceptable.

TECHNICAL SPECIFICATION We have reviewed the proposed changes to Technical Specifications 2.1, 2.2, 3/4.2.1, 3/4.2.2, 3/4.2.3, and 6.9.1.14 and the bases and find them acceptable.

In accordance with proposed Technical Specification 6.9.1.14 TVA submitted a peaking factor report for Cycle 2 as an appendix to their RSE. The report was revised in a letter dated August 3,1983. We find this report acceptable.

RELOAD CONCLUSIONS We have reviewed the information submitted on the Cycle 2 operation of Sequoyah Unit 2.

We find the proposed Cycle 2, Region 4 refueling to be accept-able from a fuel system mechanical design standpoint.

We have reviewed the nuclear design and find the proposed reload to be accept-surveillgnce able. The use of the RA0C procedure and the use of the proposed F n The change of the partial power multipTier for F procedures are acceptable.

from 0.2 to 0.3 is accettable. The interim operating procedure for the rod dYop protection may be discontinued for Cycle 2.

l g We have reviewed the proposed Technical Specification changes 2.1.1, 2.2.1, 3/4.2.1, 3/4.2.2, 3/4.2.3, and 6.9.1.14 and find them acceptable.

Hydrogen Control Requirements In their letter of July 1,1983, TVA requested cnanges to the Technical Specifi-cations of Unit 2 to make tuem consistent with the approved Unit 1 specifications.

The permanent hydrogen igniter system has been previously demonstrated to be acceptable for mitigating the consequences of hydrogen protection during degraded core accidents. The Unit 2 hydrogen mitigation system has been changed during the outage period to duplicate the Unit I system (Amendment No. 24).

Containment Protective Fuses Fuses are utilized in nuclear power plants as overcurrent protective devices for high power electrical curcuits penetrating the reactor containment. Every 18 months, the technical specifications call for testing 10% of the fuses to ensure their integrity. Unit 1 technical specifications were (Amendment No. 20) modified to permit visual inspection rather than destructive testing of a cer-tain number of fuses. The Unit 2 change is identical to that for Unit 1.

There-

. fore, the revisions will make the requirements for both Units consistent in this area.

REFERENCES 1.

Letter from L. M. Mills (TVA) to E. Adensam (NRC) with " Reload Safety Evaluation -- Sequoyah Unit 2, Cycle 2," July 1, 1983.

2.

F. M. Bordelon, et al, " Westinghouse Reload Safety Evaluation Methodology," Westinghouse Report WCAP-9273, March 1978.

3.

Letter from E. P. Rahe, Jr. (W) to C. O. Thomas (NRC),

" Westinghouse WABA Evaluation Report," W Report WCAP-10021, Revision 1, (Proprietary), October 18, T982.

4.

Letter from E. G. Adensam (NRC) to Mr. H. G. Parris (TVA),

"Sequoyah Unit 2 Cycle 2 Fuel Reload Questions," September 1, 1983.

5.

(a)

M. Tokar and C. Stahle (NRC) Telecummunication with J. E. Wills, et al, (TVA), September 2,1983.

(b) Letter from L. M. Mills (TVA) to E. Adensam (NRC) with response to NRC Questions on Sequoyah 2 Cycle 2 Reload, September 7, 1983.

_g_

" Improved Analytical Models Used in Westinghouse Fuel Rod Desi 6.

Computations," Westinghouse Report WCAP-8785 (Non-Proprietary)gn and WCAP-8720 (Proprietary), October 1976.

7.

" Safety Analysis for Revised Fuel Rod Internal Pressure Design Basis," Westinghouse Report WCAP-8964 (Non-Proprietary), March 31, 1977.

8.

Letter from H. B. Tucker (Duke Power Company) to J. P. O'Reilly (NRC) with Reportable Occurrence Report No. 369/83-11, March 24, 1983.

9.

E. P. Rahe, Jr. (W) to C. Berlinger (NRC) " Relaxation of Constant Axial Offset Control" NS-EPR-2649 Parts A and B, August 31, 1982.

10.

D. H. Risher, "An Evaluation of the Rod Ejection Accident in Westinghouse PWR's Using Spatial Kinetics Methods," WCAP-7588, Revision 1-A, January 1975.

ENVIRONMENTAL CONSIDERATION We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 551.5(d)(4),

that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.

CONCLUSION The Commission made a proposed determination that the amendment involves no significant hazards consideration which was published in the Federal Register (48 FR 36130) on August 15, 1983, and consulted with the State of Tennessee.

No public comments were received and the State of Tennessee did not have any comments.

We have concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Comission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

Dated: September 29, 1983 Principal Contributors: Carl Stahle, Licensing Branch No. 4, DL Margaret Chatterton, Core Performance Branch, DSI Michael Tokar, Core Performance Branch, DSI John Emami, Power Systems Branch, DSI L