ML20237G648

Safety Evaluation Supporting Amend 6 to License DPR-21

ML20237G648
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Site:	Millstone
Issue date:	08/06/1987
From:	Office of Nuclear Reactor Regulation
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NUDOCS 8708240173
Download: ML20237G648 (9)

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• e mfog\ UNITED STATES

[ g NUCLEAR REGULATORY COMMISSION

.5 ij WASHINGTON, D. C. 20555 8

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION J i

SUPPORTING AMENDMENT NO. 6 l TO FACILITY OPERATING LICENSING NO. DPR-21 NORTHEAST NUCLEAR ENERGY COMPANY MILLSTONE NUCLEAR POWER STATION, UNIT NO.1 DOCKET NO. 50-245

1.0 INTRODUCTION

By letter from E. J. Mroczka, Northeast Nuclear Energy Company (NNECO), to the NRC, dated May 21, 1987 (Ref. 1), Technical Specification changes were proposed for the operation of Millstone Nuclear Power Station, Unit No.1, Cycle 12 with a reload using General Electric (GE) manufactured fuel assemblies and GE analyses and methodologies. Enclosed in the May 21, 1987, letter were the requested Technical Specification changes and reports (including References 2

! and 3) discussing the reload and analyses done to support and justify Cycle 12 operation. Supplemental information was submitted by NNECO in letters dated i June 30 and July 11, 1987 (Ref. 9 and 10, respectively).

The reload for Cycle 12 is generally a normal reload with no unusual core features and characteristics. Technical Specification changes are few and l primarily related to Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits for all of the fuel, Linear Heat Generation (LHGR) limits for the new fuel and Minimum Critical Power Ratio (MCPR) limits for all of the fuel using Cycle.12 core and transient parameters and extended operating regions and conditions. ;The new fuel is the extended burnup type which has been used in l

several recent boiling water reactor (BWR) reloads (see, for example, Reference 4).

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i The Cycle 12 reload submittal includes a number of operating flexibility -]

options: single loop operation, load line limit analysis, extended load line limit analysis and feedwater temperature reduction (FWTR) at the end-of-cycle.

The effects of these operating flexibility options have been included in the Cycle 12 reload safety analysis. However, single loop operation has not been approved for Millstone 1 and will be the subject of a separate licensing action.

J l 2.0 EVALUATION 2.1 Reload Description j l

The Millstone Unit I reload will retain 384 8P8x8R GE fuel assemblies from the previous cycle and add 196 new GE8x8EB fuel assemblies. The reload safety analysis is based on a previous cycle core nominal average exposure of 19.9 GWd/MTV and Cycle 12 end-of-cycle exposure of 21.6 GWd/MTU. The loading will be a conventional scatter pattern with low reactivity fuel on the core periphery.  ;

2.2 Fuel Design i

The new fuel for Cycle 12 is the GE extended burnup fuel GE8x8EB. The fuel designation is BD338A. This fuel type has been approved in the Safety Evalua-tion Report for Amendment 10 to GESTAR-II (see Refs. 5 and 6). The specific description of this fuel has been submitted in Amendment 18 to GESTAR-II, but since this amendment has not yet been accepted, the fuel description has also been prtsented for Millstone Unit 1 in Reference 3. The staff concludes that this fuel description is acceptable.

In operation, the GE8x8EB fuel will be assigned a number of axial lattice regions. Appropriate MAPLHGR limits, which have been determined by approved thermal-mechanical and loss-of-coolant accident (LOCA) analyses, will be applied to each of these regions. There was extensive interaction among the m

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3 staff, GE, and a number of utilities in deciding on an acceptable format for presentation of this information, suitable for plant use and staff requirements for Technical Specifications. Reference 7, as amended by Reference 9, pro-vides an example of the Technical Specification for multiple lattice fuel bundles. The agreed upon Technical Specification presents the least and most  !

! limiting l'attice MAPLHGR as a function of burnup. However, the plant process computer contains, and acts on, full details of the MAPLHGR information. When hand calculations of MAPLHGR are required (i.e., when the process computer is inoperative), the most limiting MAPLHGR values as a function of burnup are used as limits for all the lattices of that bundle type. The staff concludes that the MAPLHGR values for Cycle 12 are acceptable. A proprietary report (Ref. 3),  ;

reviewed by the staff, provides complete details of the lattice definitions and MAPLHGR limits. i The proposed LHGR limit for the GE8x8EB fuel is 14.4 kW/ft rather than the 13.4 j kW/ft for other GE fuel. This LHGR has been reviewed and accepted for this )

fuel in the GE extended burnup fuel review (Ref. 5). (See the referrals in i Reference 8 to References 18 and 19. These references are responses to ques-tions and presentations relating to the GE8x8EB fuel which provide information on the 14.4 kW/ft LHGR.). The staff concludes that the proposed LHGR limit is l acceptable for Cycle 12.

2.3 Nuclear Design i

The nuclear design for Cycle 12 has been performed by GE with the approved methodology described in GESTAR-II (Ref. 6). The results of these analyses are givenintheGEreloadreport(Ref.2)instandardGESTAR-IIformat, The results are within the range of those usually encountered for BWR reloads. In particular,.the shutdown margin is 2.4 and 1.3% delta Keff at beginning-of-cycle and at the exposure of minimum shutdown margin, respectively, thus meeting the current Technical Specification required amount of 0.38% delta K,ff. The standby liquid control system also meets shutdown requirements with a shutdown Since these and other Cycle 12 nuclear design mar 5in of 5.5% delta Keff.

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parameters have been obtained with previously approved methods and fall within expected ranges, the staff concludes that the nuclear design is acceptable.

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2.4 Thermal-Hydraulic Design The thermdl-hydraulic design for Cycle 12 has been performed by GE with the approved methodology described in GESTAR-II (Ref. 6) and the results are given in the GE reload report (Ref. 2). The parameters used for the analyses are those approved in Reference 6 for the Millstone 1 class BWR-3. The GEMINI system of methods (approved in Ref. 8) was used for relevant transient analyses.

The operating limit MCPR (0LMCPR) values are determined by the limiting tran-sients, which are usually the local rod withdrawal error (RWE) and the core-wide transients feedwater controller failure, loss of feedwater heating and load rejection without by-pass (LRWOBP). The analyses of these events for Cycle 12 using the standard, approved (Ref. 6) ODYN Options A and B approaches for pressurization transients provide new Cycle 12 Technical Specification values of OLMCPRs, as a function of average scram time, for operation in both l

standard and extended operating regions. For all standard operating condi-tions, the LRWOBP event is controlling at both Options A and B limits. With a selected rod block setting of 108, the RWE is not limiting. However, the ]

licensee has opted to retain the current Technical Specification rod block l setting of 107. This is acceptable since it is conservative with respect to l RWE MCPR margin. These OLMCPRs are reflected in Technical Specification changes. Approved methods (Ref. 6) were used to analyze these events (and

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others which could be limiting) and the analyses and results are acceptable and fall within expected ranges.

The licensee states that they have examined the Cycle 12 reactor core and that ,

it is typical of previously evaluated reactors which have acceptable thermal-hydraulic stability margins (Ref. 10). The staff concludes that this ]

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assessment is acceptable since it conforms to the staff position of Generic Letter 86-02 (Ref. 11) for BWR-3s. ]

2.5 Transient and Accident Analyses

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The transient and accident analysis methodologies used for Cycle 12 are de-scribed in the NRC approved GESTAR-II (Ref. 6). The GEMINI system of methods (Ref. 8) option was used for the transient analyses. The limiting MCPR events for Cycle 12 are indicated in Section 2.4 above. The core wide transient  !

analysis methodologies and results are acceptable and fall within expected ranges.

The RWE.was analyzed on a plant and cycle specific basis (as opposed to the statistical approach) and a rod block set point of 108 was selected to provide a delta critical power ratio of 0.24 for both types of fuel bundles. However, the Technical Specifications will retain the present rod block setting of 107, which is conservative (see Section 2.4). The mislocated assembly event is not j analyzed for reload cores on the basis of NRC approved studies (see Reference l S.2-59 of Ref. 6) indicating the small probability of an event exceeding MCPR limits. The disorientation event was analyzed with standard methods for the Cycle 12 D lattice (non-symmetric water gaps) fuel, giving a-nonlimiting value l

of MCPR (Ref. 10). The staff concludes that local transient event analyses are acceptable.

l l The limiting pressurization event, the main steam isolation valve closure with

flux scram, analyzed with standard GESTAR-II methods gave results for peak j steam dome and vessel pressures well under required limits. These are accept-able methodologies and results.  !

'l I LOCAanalyses,usingapprovedmethodologies(SAFE /REFLOOD/ CHASTE)andparame-ters, were performed using MAPLHGR values for the new reload fuel bundles (GE8x8EB). The results are within the limits of 10 CFR 50.46 and are acceptable.

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l Since banked position withdrawal sequence rod patterns are used for Millstone 1, a cycle specific control rod drop accident analysis is not required. The l ,

l basis for this position and NRC approval is presented in Amendment 9 to Reference 6.

2.6 Feedwater Temperature Reduction at End-of-Cycle 12 Appendix B of the Cycle 12 reload submittal (Ref. 2) provides the results of analyses performed for the feedwater temperature reduction (FWTR) at the end-of-cycle 12. The FWTR reduction is accomplished by valving out-of-service )

the last stage feedwater heaters to provide a feedwater temperature reduction f of 7S*F. This FWTR extends the end-of-cycle 12 from 21.6 GWd/MTU to 22.4 l I

GWd/MTV. The pressurization events were reanalyzed for the FWTR. The LRWOBP transient remained the limiting event and established the OLMCPRs for both the Options A and B ODYN categories for the exposure range from end-of-cycle 12 to ,

the extended end-of-cycle 12. The MCPR analyses for FWTR use standard methods and with results in expected ranges and are acceptable.

2.7 Technical Specifications The Technical Specification (TS) changes for Cycle 12 are to provide for: )

(a) The 14.4 kW/ft LHGR limit for the new (GE8x8EB) fuel. The changes are to Definition M, TS 2.1.2.A.1.b, TS 2.1.2.B.1.k. and TS 3.11.B and are acceptable.

1 (b) MAPLHGR limits for the fuel. The changes, which were revised, in part, in Reference 9, are to TS 3.11.A.1 and Figures 3.11.la and 3.11.lb. Figure

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3.11.1c has been deleted. These changes are acceptable.

(c) The new MCPR limits for Cycle 12, including extended cycle operation. The changes are to TS 3.11.C and Basis 3.11.C and are acceptable.  ;

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Each of the above changes has been previously discussed and approved in this i

review.

3.0 ENVIRONMENTAL CONSIDERATION

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This amendment involves a change in the installation or use of facility compo- l nents located within the restricted area as defined in 10 CFR Part 20. The staff j has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released off- ,

i site and that there is no significant increase in individual or cumulative occupa-tional radiation exposure. The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment  ;

need be prepared in connection with the issuance of this amendment.

4.0 CONCLUSION

The staff has reviewed the reports submitted for the Cycle 12 operation of l

Millstone 1 with extended operating regions. Based on this review, we conclude that appropriate material was submitted and that the fuel design, nuclear design, thermal-hydraulic design, and transient and accident analyses are acceptable. The Technical Specification changes submitted for this reload

! suitably reflect the necessary modifications for operation in this cycle.

The staff has concluded, based on the considerations discussed above, that:

l (1) there is reasonable assurance that the health and safety of the public will J 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 the amendment will not be inimical to the common defense and security or to the health and safety of the public, i

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5.0 REFERENCES

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1. Letter and enclosure from E. J. Mroczka (NNECO) to NRC, dated May 21, 1987. Application requesting changes to the Millstone 1 Technical Speci-fication for Cycle 12 operation.

2. GE Report 23A4845 Revision 0, dated April 1987, " Supplemental Reload Licensing Submittal for Millstone Point Nuclear Power Station, Unit 1, Reload 11."

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3. " Loss-of-Coolant Accident Analysis Report for Millstone Unit 1 Nuclear I

Power Station - Supplement 1," NEDE-24085-1-P, Supplement 1, April '1987.

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4. Letter (and enclosure) from R. Clark (NRC) to E. Bauer (PEC), June 1987 (Cycle 8 core reload for Peach Bottom Unit 2). i l

5. Letter (and attachment) from C. Thomas (NRC) to J. Charnley (GE) dated May l 28, 1985, " Acceptance for Referencing of Licensing Topical Report /

NEDE-24011-P-A-6, Amendment 10."

6. GESTAR-II, NEDE-24011, Revision 8 " General Electric Standard Application for Reactor Fuel."

! 7. Letter from J. Charnley (GE) to M. W. Hodges (NRC) dated March 4,1987,

" Recommended MAPLHGR Technical Specifications' for Multiple Lattice Fuel Designs."

8. Letter (and attachment) from G. Lainas (NRC) to J. Charnley (GE) dated l

March 22, 1986, " Acceptance for Referencing of Licensing Topical Report, j

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NEDE-24011-P-A, 'GE Generic Licensing Reload Report,' Supplement to Amendment 11."

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Letter from E. J. Mroczka (NNECO) to NRC, dated June 30, 1987. This reference provides revised wording for Technical Specification 3.11.A.I. -

Letter from E. J. Mroczka (NNECO) to NRC, dated July 11, 1987. This reference provides revised page 12 of Reference 2.

" Technical Resolution of Generic Issue B-19: Thermal-Hydraulic Stabili-ty " Generic Letter No. 86-02, dated January 23, 1986.

4 lc ipal Contributor: D. Fieno --

td: August 6,1986 f

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