ML20078R068

Safety Evaluation Supporting Amend 90 to License DPR-50

ML20078R068
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Site:	Crane
Issue date:	10/28/1983
From:	Office of Nuclear Reactor Regulation
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NUDOCS 8311140334
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 90 TO FACILITY OPERATING LICENSE NO. DPR-50 METROPOLITAN EDISON COMPANY JERSEY arHRAL F0WER AND LIGHT COMPANY PENNSYLVANIA ELEuRIC CUMPANY GPU NUCLEAR C0xPURAi10N THREE MILE ISLAND NUCLEAR STATION, UNIT NO.1 DOCKET NO. 50-289 1.0 Introduction In response to a request (Ref.1) from the TMI-1 Restart Issues Task Force, we have reexamined the Three Mile Island Unit 1 Cycle 5 Reload Application (Refs. 2-3) and the original (March 16, 1979) NRC staff evaluation (Ref.4) of that application to determine their continued validity. The reevaluation

.was conducted in a manner similar to all other operating plants. That is, current, rather than 1979, NRC requirements were applied in the review.

As a result of this reexamination, we identifed two issues that were not addressed in the original reload a,,111 cation, but are now being addressed by all other operating B&W reactors. The first issue concerns the so-called TAFY/ TACO penalty proposed by B&W (Ref. 5) and accepted by the NRC staff (Ref. 6) to account for a previously undetected nonconservatism in the Loss of Coolant Accident (LOCA). initial conditions. The second issue concerns nonconservative cladding swelling and rupture models as discussed in our July 13,1982 letter (Ref. 7) to the ~ licensee, GPU Nuclear Corporation.

i, In resoonse to these two issues, tne GPU Nuclear Corporation submitted on June 20, 1983, Technical Specification Change Request No.127 requesting amendment to Facility Operating License No. DPR-50. This enange request also included several other minor changes to the Technical Specifications and are discussed in the following sections.

2.0 'TAH/TACD Penalty The Three Mile Island Unit 1 (TMI-1) Cycle 5 Reload Application makes use of

. two BW ibel thermal performance codes, tan-3 (Ref. 8) and TACO-1 (Ref. 9).

Althouch both of these codes have been approved for use in safety analysis, we believe (Ref.10) that only the newer TACG-1 code is capable of correctly calculating fission gas release (and therefore rod pressure) at high burnups.

Babcock & Wilcox has responded (Ref. 11) to this concern with an analytical cocparison between the TAFY-3 and TACO-1 codes. In this response, they have stated that the fuel rod internal pressure predicted by TACO-1 is lower than that precicted by tan-3 for fuel rod exposures of up to 42,000 mwd /Mtu.

Although we have not examined this comparison, we notc that the analyses exceed tne =aximu= expected exposure (32,387 m'dn:tu peak asse=bly) for all fuel in the TXI-1 Cycle 5 core.

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For the LOCA analysis (Section 7.2 of the Reload Report), tne average fuel temperature as a function of linear heat rate and the lifetime pin pressure data were calculated with the older TAFY-3 code. The licensee has stated that corresponding parameters used in the generic LOCA analysis are conservative compared with those calculated for Cycle 5 at Three Mile Island Unit 1.

However, information obtained by the NRC staff (Ref.12) indicates that these i

TAFY-3 predictions do not produce higher calculated peak cladding temperatures in the generic LOCA analysis than the newer TACO-1 code. The issue involves excessive fuel densification and lowered fuel rod internal gas pressures at beginning of life. Babcock & Wilcox has proposed a method of resolving this issue which has been adopted by GPU Nuclear (Ref.13). The method relies on reduced peak linear heat rate (PLHR) limits at low core elevations for the first 50 effective full power days (EFPD) of operttion and is based upon a comparison of TAFY-3 and TACO-1 calculated LOCA initial conditions.

Two sets of bounding values for allowable LOCA PLHRs are given as a function of core height. The first set applies to the first 50 EFPD and the second set to the balance of Cycle 5.

We have reviewed the comparison of Cycle b limits to limiting core protection safety limits given in Table 1 of the original reload report (Ref. 2). We find that sufficient margin *is available in the Cycle 5 Technical Specifications on rod index, axial power shaping rod position, and axial power imbalance so as to bound the interim LOCA PLHR limits from 0 to 50 EFPD. We, therefore, conclude that the new limits are satisfactorily incorpo-rated into the Technical Specifications for Cycle 5.

3.0 Cladding Swelling and Rupture Models In addition to the issue of initial fuel temperatures and rod internal pressures used in the LOCA analysis, a second issue involving cladding swelling and rupture models has affected the proposed Cycle 5 operating limits for TMI-1. In late,1979, the NRC staff reviewed Emergency Core Cooling System (ECCS) fuel cladding models in light of new data. Adequai:y of the models then in use was ques,t'ioned and new models, developed as Appendix X acceptance criteria, were presented in NUREG-0630 (Ref.14). Each fuel vendor was then askec' to show how, in light of the new models, the plants analyzed with their analytical methods I

continued to meet the applicable LOCA limits. The B&W response (Ref.15) concluded that the impact of the NRC models was small and did not result in analytical results in excess of the LOCA limits.

A more recent B&W calculation (Ref.16), however, found that the cladding swelling and rupture models presented by the staff have a non-trivial effect on LOCA peak cladding temperatures in B&W 177 fuel assembly plants. Because this calculation was applicable to all B&W plants, the licensee was requested (Ref.17) to provide supplemental calculations for TMI-l similar to those provided in Reference 16. The licensee's responses (Refs.18,19 and 20) culminated in a supplemental ECCS calculation (Ref. 21) for TMI-1. This calculation not only considers cladding swelling and rupture effects, but i

also considers the fuel densification effects with a more recent B&W fuel performance code called TACO-2 (Ref. 22). Tne combined analysis results in low core elevation PLHR limits which are more restrictive than those which consider only fuel densification (with TACO-1).

In general, the supplemental calculation utilizes previously approved methods except for the substitution of the NRC cladding models. However, tnere are segments of the analysis (e.g. THETAl-B-Ref. 23) that are currently undergoing NRC review. The licensee has also presented results from a calculation using a_. -..,.

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-3 a new FLECSET heat transfer correlation (Refs. 24 and 25). This correlation appears to offset the NUREG-0630 penalties. However, the benchmarking and other final evaluations of FLECSET have not been completed and provided to the NRC for review. Because the FLECSET correlation has not yet been approved, the licensee has committed (Ref. 22) to aaministratively implement the more restrictive operating limits during early in Cycle 5 operation (i.e., less than 31 EFPD and greater than 80% power). Because of the planned Cycle 5 power ascension schedule, it is expected that these limits will not affect actual operation.

Considering the above, we conclude that the licensee's proposed administrative limits on operation are both appropriate and necessary. Since these operating limits are more restrictive than those previously proposed (Ref.13) for TMI-1, since they are only needed for a brief time period, and since potential but unused compensating benefits may exist, we therefore conclude that the acministrative limits imposed on an interim basis are acceptable for incorpo-rating the NUREG-0630 penalties until our final evaluation of FLECSET is completed.

1 4.0 Other Modifications to the Technical Specifications The licensee has also requested (Ref.13) several ether minor modifications to the plant Technical Specifications. These are discussed below.

The bases for the Technical Specifications incorrectly retained a centerline fuel melt limit based on a fuel assembly design no longer used at TMI-1. The revived fuel melt limit is based on the fuel design (Mark B-4) used in the Cycit 5 core. We find this enange acceptable.

The licensee proposes to change the nuclear power setpoint based on the reactor coolant pump monitors (Item 3 of Table 2.3.1 - RPS Trip Setting Limit) from 91%

to 55% of rated power for 1/1 (one pump in each loop) RC pump operation. The pump

• monitor's power setting is intended to prewnt the core minimum Departure from Nucleate Boiling Ratio (DNBR) from decreasing below the 1.3 limit by tripping the reactor due to the loss of RC pumps. This is a redundant trip function complemental to the flux / flow trip setting where the power level trip setpoint is based on a power to flow ratio adjusted for the power imbalance.

The existing 91% setpoint was established by the original Final Safety Analysis Report. The licensee has indicated that an evaluation accounting for changes in DNBR-related design factors since Cycle 1 has shown the 91% limit to remain conservative for Cycle 5 operation. The decision to reduce the setpoint from 91% to 55% of rated power is based on B&W recommendations for the 177 fuel assembly plants with redundant pump monitors to provide a consistent basis for any future B&W analyses. This change will not significantly impact anticipated operations of TMI-1. Since the proposed change is in a more conservative direction, we conclude that it is acceptable.

The Technical Specifications have also been rewritten so that the quadrant tilt is always determined by the most accurate detector system available. This does not represent a change in tNs existing allowable tilt limits for any of the detector systems.

We find this administrative change acceptable.

5.0 Summary From our reexamination of the Cycle 5 Reload Applicatior., =d from car review of additional information submitted by the licensee, we conclude that this core reload will not adversely affect tne Three Mile Island Nuclear Station's ability to operate safely during Cycle 5 of Unit 1.

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4-6.0 Environmental Consideration We have determined that the amendment does not authorize a change in affluent 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 insignifi-cant from the standpoint of environmental impact and, pursuant to 10 CFR {51.5 (d)(4), that an enviromnental impact statement, or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.

7.0 Conclusion 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 Commission'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.

t Dated: October 28, 1933 The following NRC personne1' have contributed to this Safety Evaluation:

J. Voglewede, Y. Hsii, L. Kopp.

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REFERENCES R. Starostecki and D. G. Eisenhut (NRC) memorandum to R. Jacobs et al.

1.

(NRC) on "TMI-1 Restart Issues Task Force" dated April 15,1983 2.

J. G. Herbein (Meted) letter to R. d. Reid (NRC) on " Technical Specification Change Number 86" dated December 28, 1978.

3 J. G. Herbein (Meted) letter to R. W. Reid (NRC) on " Cycle 5 Reload -

Additional Inforsation" dated March 1,1979.

4.

R. W. Reid (NRC) letter to J. G. Herbein (Meted) on " Amendment 50 to

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Facility Operating License No. DPR-50" dated March 16, 1979 5

J. H. Taylor (B&W) letter to L. S. Rubenstein (NRC) dated September 5, 1980.

6.

L. S. Rubenstein (NRC) letter to J. H. Taylor (B&W) cated October 28, 1980.

7.

T. M. Novak (NRC) letter to H. D. Hukill (GPU) dated July 13, 1982.

8.

C. D. Morgan and H. S. Kao, "TAFY - Fuel Pin Temperature and Gas Pressure Analysis," Babcock and Wilcox Company Report BAW-10044, May 1972.

9 R. H. Stuudt et al.,'" TACO: Fuel Pin Performance Analysis," Babcock and Wilcox Company Report BAW-10037P-A, Rev. 2, August 1977.

10. D. F. Ross, Jr. (!!RC) letter to J. H. Taylor (B&W) dsted January 18, 1973.

11. J. H. Taylor (B&W) letter to L. S. Rubenstein (NRC) dated September 5,1980.

12. R. O. Meyer (NRC) me:norandum for L. S. Rubenstein (NRC) on "TAFY/ TACO Fuel Performance Models in B&W Safety Analysis" dated June 10, 1930.

13 9./. Hukill (GPU) letter to Director of Nuclear Reactor Regulation (NRC)

D on " Technical Specification Change Request No.12't" dated June 20, 1993

14. D. A. Powers and R. O. Meyer, " Cladding Swelling Models for LOCA Analysis,"

U. 5. Nuclear Regulatory Cocraission Report NUREG-0630, April 1980.

15. J. H. Taylor (B&W) letter to L. S. Rubenstein (NRC) dated October 28, 1980.

16. J. W. Cook (Consuners Power) letter to H. R. Denton (NRC) dated April 2, 1982 and trans.itting B&W Report No. 12-1132424, Revision 0, "Dounding Analysis Impact Study of NUREG-0630."

17. T. M. Novak (NRC) letter to H. D. Hukill (GPU) dated July 13, 1932.

18. H. D. Hukill (GPU) let.ter to T. M. No.2k (NRC) on " Clad Swell and Rupture Models" dated September 17, 1982.

19. H. D. Hukill (GPU) letter to T. M. Novak (NRC) on " Clad Swell and Rupture Models" dated Nove2er 1,1982.

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20. H. D. Hukill (GPU) 10tt;r to T. M. Novak (NRC) cn " Clad Swell and Rupture Modela cated April 4, 1983

21. H. D. Hukill (GPU) letter to T. M. Novak (NRC) on "NUREG 0630 Cladding Suell and Rupture Models for LOCA Analysis" dated July 14, 1933

22. Y. H. Hsii et al., " TACO 2: Fuel Pin Perfor: nance Analysis," Babcock and Wilcox Company Report BAW-10141P, January 1979.

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23. " Babcock & Wilcox Revisions to THETA 1-B, a Computer Code for Nuclear Reactor Core Thermal Analysis (IN-1445) - Revision 3," Babcock & Wilcox Company Report BA%-1tD94 Rev 3, February 1981.

24. N. Lee, S. Wong, H. C. Yeh, and L. E. Hochreiter, "PWR M.ECHT SEASET Unblocked Bundle, Forced and Gravity Reflood Test Data Evaluation and Analysis Report, NUREG/CR-2256 (EPRI NI-2013 or WCAP-9391),Ibvenber 1981.

25. G. P. Lilly, et al., PWR FLECHT Skewed Profile Low Flooding Rate Test Series Evaluation Report," WCAP-9133, Never.ber 1977 e

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