ML20128E999

Safety Evaluation Supporting Amend 84 to License DPR-22, Respectively

ML20128E999
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Site:	Monticello
Issue date:	01/27/1993
From:	Office of Nuclear Reactor Regulation
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NUDOCS 9302110202
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LAFETY EVA1.UA110N BY THE OFFICE OF NUCLEAR REACTOR REGULATION A

RELATED TO AMENDMENT NO.

84 T0 FACILITY QPERATING LI WESE NO. DPR-22 HORTHERN STATES POWER COMPANY MONTICELLO NUCLEt,R GENERATING PLANT

@CKET NO. 50-263_

5 1.0.11(11QDUCT10N By letter dated September 16, 1992, as supplemented November 3, 1992, the Northern States Power Company (the licensee) requested an amendment to the Technical Specifications (TS) appended to facility Operating License No. DPR-22 for the Monticello Nuclear Generating Plant.

The proposed amendment approves 15 changes to permit implementation of an expanded operating domain resulting from maximum extended load line limit analysis (MELLLA) and increased core flow (ICF).

The November 3,1992, submittal provided clarifying information that did not i

change the initial proposed no significant hazards determination.

2.0 EVALVATION 2.1 D G R E110N OF PROPOSED TS CHANGES Averaae Pnwer Ranae Monitqr (APRM) Flow-Biated Scram Set Point - TS 2.3. A. and Table LL1:

The proposed amendment would change the flux scram lines on the power / flow map and establish a new two loop APRM flux scram line of 0.66W +

70%.

The existing two-loop APRM flux scram function is 0.58W + 624.

The single-loop line would remain at 0.58(W-5.4) + 62%.

(Note:

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percentage of drive flow required for 100% core flow.) This change would allow higher power operation when core flow is below 87%.

In addition, a high flow clamp of 120% rated power would be added to the APRM scram specifications.

The Updated Safety Analysis Report-(USAR), Section 7.3.5.2.2 indicates that the 120% clamp instrumentation is currently-installed.

Ayarjao Power Ranae Monitor flow-Biased Rod Block Set Point - Table 3.2.31 The proposed amendment would similarly change the rod block monitor (RBM) set point requirements, establishing a two-loop set point function of 0.66W + 58%,

and a high flow clamp of 108% rated power.

The present two-loop operation rod block set point function is 0.58W + 50%. The proposed changes maintain the same maximum set points (120% for APRM scram.and 108% for rod block) as currently approved. The changes also maintain the same margin (12%) between 9302110202 930127 ADOCK0500g3--

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t the APRM scram and rod block set points as provided in the current Technical Specifications. As for the scram function, the single-loop operation requirement would not be changed.

Ihe Bases (LJ. A) and Core Operal.ing Limits Report fC01H)_Reportina Ets kements (15 6.7.a):

Ine Bases sections of the 15 would be revised to reflect the changes to affected Limiting Safety System Settings and Limiting Conditions for Operation.

References to supporting analyses would be included in the revised Bases. Also, reporting requirements for the COLR would be changed to require that the COLR for each reload include a power / flow map.

2.2 EYALu&II.03 The proposed amendment affects safety system set points for two protective functions, (1) the APRM flow-biased scram, and (2) the RBM trip.

Both functions utilize the APRM neutron monitoring system.

The function of the slope and bias circuits of the APRM flow-biased scram is to account for-the decreasina mar vith reduced recirculation flow. gin to fuel damage at a given power-level The function of the RBM is to prevent rod withdrawal under conditions that could initiate a rod withdrawal error event leading to local fuel damage.

Operation with the proposed changes implemented on these protective functions has been analyzed for limiting accidents, transients, and thermal-hydraulic stability.

The results of the Monticello analyses are reported in (a) NEDC-31849P

" Maximum Extended Load Line Limit Analy>is for Monticello Nuclear Generating Plant Cycle 15," (b).NEDC-31849P-1, " Maximum Extended Load Line Limit Analysis for Monticello Nuclear Generating Plant Cycle 15, Supplement 1," and (c) NEDC-31778P " Safety Review for Monticello Nuclear Generating Plant Increased Core Flow Operation Throughout Cycle 14."

These documents provide the results of new analyses and evaluations supporting HELLLA and ICF operation and discussions providing the basis for not reanalyzing certain events.

Events addressed in these reports include:

A.

Transients:

(1) Feedwater Controller Failure with Maximum Demand (MELLLA and ICF)

(2) Turbine Trip w uh Bypass failure (MELLLA and ICF)

(3) Loss of Feedwater Heating (MELLLA)

(4) Closure of One Turbine Stop Valve at 100% Power and 75% Flow (MELLLA)

(4) Transfer to Backup Pressure Regulator (MELLLA)

(5) Slow Flow Runout (MELLLA)

(6 Main Steam Isolation Valve Closure With Flux Scram (MELLLA and ICF)-

(7 Rod Withdrawal Error Event (MELLLA and ICF)

(8 Rod Drop Accident (ICF) 8.

Design Basis Accident (DBA)

(1) DBA-LOCA Containment Pressure Response and containment dynamic loads (MELLLA)

(2) DBA-LOCA Peak Clad Temperature (MELLLA and-lCF)

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C.

Analysis of effect of acoustic loads and flow-induced loads on reactor internals (MELLtA and ICF)

D.

Analysis of effects on flow-induced vibration (MELLLA and ICF)

E.

AnticipatedTransientsWithoutScram[ATWS)(MELLLA)

TransieAta: Analysis rasults for the turbine trip with bypass failure and feedwater controller failure to maximum demand for 100A power and 75% core flow initial conditions resulted in peak vessel pressures, peak neutron flux and heat flux, and delta-Critical Power Ratios (CPRs) less than the corresponding values for the transient with 100% power and 100% core flow initial conditions.

The peak vessel pressure for the HSIV closure with flux scram event initiated from 100% power and 75% core flow was slightly less than the corresponding values for the transient with 100% power and 100% core flow initial condition.

For the closure of one turbine stop valve event and for the transfer to backup pressure regulator event, the peak neutron flux values for the 100% power and 75% core flow initial condition were well below the 120% rated power APRM clamp.

The slow flow runout event was analyzed due to the fact that the power increase for this event occurs along a steeper rod line. Reanalysis was performed for various. flow conditions down to 30% rccirculation pump speed, i

with each flow condition on the maximum possible rod line.

The analysts demonstrated that results are acceptable using current flow dependent minimum critical power ratio (MCPR,) and flow dependent maximum average planar linear heat generation rate factor (MAPFAC,) limit curves.

Pod Withdrawal Error (RWE1:

The RWE analysis for the Extended Load Line Limit Analysis (ELLLA) region (Amendment 29) bounds the RWE for-the MELLLA and ICF regions.

Desian_ B_giis Accidents - Peak Clad Temperature (PCT):

SAFER /GESTR and 10 CFR Part 50, Appendix K calculations for a 102% power and 80% core flow initial condition indicate a PCT of 1623*F. This is less than the 1769'F PCT calculated based on the 102% power and 105% initial flow condition.

This indicates that for a 75% flow initial condition, the PCT would be expected to be close to that calculated for the 80% flow initial condition, considerably under the 2200*F limit.

For ICF operation, the initial core heatup will be lessened, but the core uncovery duration is increased resulting in.a slightly increased PCT, on the order of 10'F, which is insignificant.

-Desian Basis Accidents - Cont 11nment-Pressure Response:. The'DBA-LOCA short--

term containment response for initial conditions in the HELLLA region was found to be bounded by the current USAR analysis limiting value-of 42 psig which is below the containment design value of 56 psig. -Long-term containment-response is unaffected by initial flow.

Containment dynamic loads, including pool swell, condensation oscillation, chugging, and vent clearing are also not impacted by ooeration in the MELLLA region.

2 4

Desian Basis Accidents - Effect on Reactor Vessel Internals:

Pressure differential forces on fuel assemblies and vessel internals for DBAs (the main

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steam line (MSL) break upstream of flow restrictors is the worst case) initiated from the MELLLA region are bounded by the values for higher flow operation. Reanalysis was performed for ICF operation and stresses were found to be within allowable values.

For the MELLLA region and ICF, flow-induced vibrations and acoustic loads are evaluated with particular attention on the shroud, separator assembly, jet sumps, and incore housing and control rod guide tubes.

It was concluded t1at balanced recirculation loop operations in the MELLLA region and ICF operations up to 105% rated core flow are acceptable.

(Note:

The licensee is currently evaluating test results from a series of jet pump tests which were conducted as a result of cracking in a jet pump upper riser brace.

Data indicates that riser brace harmonic frequencies may exist at certain pump speeds.

Pending completion of full scale tests and analysis of findings, the licensee is restricting operation of the recirculation pumps in the area of concern.

The lower riser braces at Monticello are capable of c

providing the required jet pump support and the upper bc :es are considered-unnecessary.

The potential effects of loose parts resu: cing from upper riser brace f ailure have been analyzed by the vendor and licensee, and found to result in no safety concern.)

. Thermal-Hydraulic Stability:

Thermal-hydraulic stability reanalysis was not necessary in support of MELLLA/ICF.

Stability performance of GE fuel, which is used exclusively at Monticello, is generically demonstrated in NEDE-22277-P-1 " Compliance of GE BWR fuel Designs to Stability Licensing Criteria." This NEDE-22277-P-1 has been incorporated as part of the GESTAR 11 methodology, This methodology utilizes calculations to show that individual channels are as-stable or more' stable than the core, thus indicating the core-wide limit cycle oscillations will not occur.

AM:

The limiting ATWS event (MSIV Closure with one safety relief valve out of service, recirculating pumps trip and alternate rod insertion) was reanalyzed for the 100% power and 75% flow initial condition in the MEl,LLA region.

Analysis results of peak vessel pressure, peak containment. pressure, and peak suppression pool bulk temperature were within applicable design guidelines.

Summary:

The new and earlier analyses encompass and bound the range of effects on accidents and transients resulting from-the expanded MELLLA/ICF operating regime.

The areas examined are the same areas examined in previous MELLLA-ICF reviews such as that for Fermi-2 reported in the staff's Safety Evaluation of May 15, 1991, supporting Amendment 69 to Operating License NPF-43.

Based on the results'of the analyses and evaluations riescribed in the --

report, maximum extended lohe line operation will not result in significantly reduced fuel thermal margins and will n9t compromise the structural integrity of the containment or of the reactor vessel or its internals.

Based on the findings, the proposed amendment is acceptable.

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The analyses conducted in support of HELLLA and ICF operation are cycle specific and will be reanalyzed for each reload as necessary.

However, the associated TS changes encompass anticipated future reloads.

The addition of the cycle dependent Power to flow Hap to COLR as described in TS 6.7.a is a clarification and therefore, the staff finds it acceptable.

3.0 AIA11_C.0MLIAT10B In accordance with the Commission's regulations, the Minnesota State official was notified of the proposed issuance of the amendment. The state official had no comments, 4,0 ENVIRONMENTAL CONSIDEPAIl0B The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.

The staff 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 offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding (57 FR 48823). Accordingly, the amendment meets the aligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9),

Pursuant to 10 CFR 51.22(b), no environmental impa::t statement or environmental assessment need be prepared in connection with the issuance of the amendment.

5.0 CONCLUSION

The staff has 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 pro 30 sed manner. (2) such activities will be conducted in compliance wit 1 the Commission's regulations, and (3) the issaance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors:- William Long, A. H. Hsia Date: January 27, 1993

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