ML20148N471

Safety Eval Supporting Amend 37 to Provisional Oper Lic DPR-22.Concludes Issuance of Amend Will Not Be Inimical to Nation or People

ML20148N471
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Site:	Monticello
Issue date:	11/06/1978
From:	Office of Nuclear Reactor Regulation
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NUDOCS 7811270106
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1. % UNITED STATES

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NUCLEAR REGULATORY COMMISSION l WASHINGTON, D. C. 20556 g

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION 1 1

' SUPPORTING AMENDMENT N0. 37 TO PROVISIONAL OPERATING LICENSE N0. DPR-22 NORTHERN STATFS POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT ]

DOCKET NO. 50-263 l 1

1.0 Introduction Northern States Power Company has proposed changes to the Technical Spec-ifications of the Monticello Nuclear Generating Plant (MNP). The proposed i changes relate to the replacement of 112 fuel assemblies constituting i refueling of the core for seventh cycle operation at power levels up to 1670 Mwt (100%).

In support of the reload application, the licensee has provided the GE BWR Reload 6 licensing submittal for MNP (Reference 1, 2), proposed Technical Specification changes (Reference 1, 2), information on the MNP Loss of Coolant Accident (LOCA) analysis (Reference 3 and 4), and responses to NRC l requests for additional information (Reference 5). The proposed amendment was noticed in the FEDERAL REGISTER on July 10,1978 (43FR29633)'

This reload involves MNPs first loading of the GE Retrofit (8x8 R) fuel. The description of the nuclear and mechanical design of the reload (8x8 R) fuel and the irradiated (8x8) fuel is contained in GEs licensing topical report for BWR reloads (Reference 6). Reference 6 also contains a complete set of references to topical reports which describe GEs analytical methods for nuclear, thermal-hydraulic, tran-sient and accident calculations, and information regarding the applicability of these methods to cores containing (8x8), and (8x8 R) fuel.

Values for each plant-specific data such as steady state operating pressure, core flow, safety and safety / relief valve setpoints, rated thermal power, rated steam flow, and other various design parameters are provided in Reference 6.

Additional plant and cycle dependent information are provided in the reload application, .(Reference 1), which closely follows the outline of Appendix A of Reference 6.

Reference 8, describes the staff's review, approval, and conditions of approval for the plant-specific data addressed in Reference 6. The above mentioned plant-specific data have been used in the transient and accident analysis' provided with the' reload application.

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a Our safety evaluation (Reference 8 ) of the GE generic reload licensing topical report concluded that the nut. lear and mechanical design of the (8x8 R) fuel, and GEs analytical methods for nuclear, thermal-hydraulic, and transient and accident calculations as applied to mixed cores containing (8x8), and (8x8 R) fuel are acceptable. Approval of the nuclear and mechanical design of (8x8) fuel was determined based on information in Reference 7 and expressed in the staff's status report (Reference 9) on that document.

Because of our re iew of a large number of generic considerations re-lated to use of (8x8 R) fuel in mixed loadings with (8x8) fuel, and on the basis of the evaluations which have been presented in Reference 8, only a limited number of additional areas of review have been included in this safety evaluation report. For evaluations of areas j not specifically addressed in this safety evaluation report, the I reader is referred to Reference 8. i l

Thermal hydraulic improvement features discussed in Reference 6, and l other plant improvement features or changes associated with this reload, i or specific plant salient features that may impact on plant operational l performance or safety limit MCPRs are discussed in section 2.2.3. l l

2.0 Evaluatinn j 2 .1 Nuclear Characteristics _

For Cycle 7 operation of MNP, 52 fresh (8x8 R) fuel bundles of type 80RB265 and 60 fresh (8x8 R) bundles of type 80RB282 will be loaded into the core / Reference 1). The remainder of the 484 fuel bundles in the core will be (8x8) fuel exposed during the previous cycles.

The fresh fuel will be loaded in a core pattern as shewn in Figure 1 of Reference 1, which is acceptable.

Based on the data presented in sections 4 and 5 of Reference 1, both the control rod system and the standby liquid control system will have acceptable shutdown capability during Cycle 7.

2.2 hrmal Hydraulics 2 . 2 .1 Fuel Cladding Integrity Safety Limit As stated in Reference 6, the minimum critical power ratio (MCPR) which may be allowed to result from cere-wide or localized transients is 1.07. This limit has been imposed to assure that during transients 99.9% of the fuel rods will avoid transition boiling.

The safety limit MCPR for MNP is being raised from 1.06 to 1.07 because the distribution of fuel rod power within the (8x8 R) fuel bundles is flatter than that of the (8x8) fuel . The reason for the flatter power

ll distribution is the presence of two rather than one water rods in ij l'

(8x8 R) - fuel . The issue has been addressed in Reference 8 and the l .07 limit has been found acceptable for BWRs with uncertainties in L flux monitoring and operational parameters no greater than those li listed in Table 5-1 of Reference 6, for which the CPR distribution f is within the bounds of Figures 5.2 and 5.2a of Reference 6. It has .,

been proposed in Reference 1 that these conditions are met for MNP l-Cycle 7, (see section 3.;0).  ;

2.2.2. Operating Limit MCPR [

Various transients or perturbations to the CPR distribution could reduce  ;

the CPR below the intended operating limit MCPR during Cycle 7 operation L of MNP. The most limiting of these operational transients and poten-tial fuel loading errors have been analyzed by the licensee to determine which event could potentially induce the largest reduction in the critical power ratio (aCPR) (References 1 and 2).  ;

e The transients evaluated were the generator load rejection without bypass, g feedwater controller failure at maximum demand, the turbine trip with i; failure of the bypass valves, loss of a 100 F. feedwater heating, and the control rod withdrawal error. Initial conditions and transient input parameters as specified in Sections 6 and 7 of Reference 1 were i:

assumed.

The calculated systems responses and aCPRs for the above listed l operational transients and conditions have been analyzed by the licensee.

Table S-1 lists the aCPRs for the various fuel types at the specified cycle exposure. Also included in Table S-1 are the results of the fuel loading error, Section 2.6.3, and the maximum vessel pressure discussed in Section. 2.4.

TABLE S-1 l ACPR MCPR tl Transient Limiting Exposure Time 8x8/8x8R Operating Limit [

l Load Rejection E0C .24/.24 N/A l without Bypass  !

Turbine Trip + EOC .26/.26 1.33/1.33 wi thout Bypass  :

Loss of 100*F BOC-EOC .16/.17 N/A [

Feedwater Heater [

Feedwater B0C-E0C .22/.22 N/A Controller Fail ure l

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-ACPR MCPR Condition Limiting Exposure Time 8x8 / 8x8R Operating Limit Rod-Withdraw BOC-E0C .10/.15 N/A Error (105%)

' Fuel Loading Error- BOC-EOC .38/.33 (see section 2.6.3)

Overpressurization + '

(MSIV Closure) Peak vessel pressure assuming 7 of the 8 SRVs is 1248 psi Notes + See section 2.2.3.2 Addition of the most severe aCPR to the safety limit (1.07) gives the appropriate operating limit MCPR for each fuel type. This will assure that the safety limit MCPR is not violated due to transients or a fuel loading error.

We have determined that the operating limit MCPRs listed in Table S-1 are acceptable for Cycle 7 operation at the MNP plant, (see section 2.6.3).

2.2.3. Thermal-Hydraulic Improvement Features 2.2.3.1 Load Shedding RPT The Monticello Plant electrical protection design scheme incorporates trip circuitry for the recirculation pump M-G set drive motors. This existing feature avoids the transfer of large, non-essential electrical loads from the auxilliary transformer to the reserve transformer in the event of a turbine or generator trip. The pump trip uses devices having a long history of reliable operation with certain elements of redundancy; however, it does not meet the same rigorous standards as does the reactor protection system. When fuel thermal limits were analyzed, it was found that limiting transients were not affected by the recirculation pump trip. The reason is that the recirculation pump trip occurs too late to have any effect in reducing thermal limits. Turbine and generator trips were modeled to include the pump trip to best represent the existing equipment.

Because the load shedding recirculation pump trip (LSRPT) has no influence on the (aCPR) thermal limits, we find the results of the transient modeling which include the LSRPT acceptable.

2.2.3.2 Simmer Margin The licensee has proposed Technical Specification changes (Reference 2) to increase the simmer margins on their eight safety / relief valves (SRVs) by 28 psi. Additional transient, overpressure, and LOCA analyses were performed tol justify the setpoint increase in the SRVs to 1108 psi. In i

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addition to the increased setpoint, the licensee assumed credit for '

only seven of the eight valves during the analyses. ,

1 Results of the transient and overpressurization analysis showed increases o f 0.0) in aCPR and 8 psi incre'se a in the vessel pressure respectively.

These results are included in Table S-1.  !

Analysis of the design basis loss-of-coolant accident indicated that the increased setpoint of the safety / relief valves has no effect on the resul ts . However, for small breaks, the reactor will remain pressurized l until the initiation of the automatic depressurization system (assuming l the single failure of the HPCI). Tne increase in the setpoint of the safety / relief valves result in a slight increase in inventory loss of l the break during this period. The small break analysis predicts a PCT of approximately 40 F higher than that for the case of the old SRV set-point, for the most limiting small recirculation line break. This small increase in PCT is due primarily to the fact that the higher setpoint on the SRVs result in higher vessel pressure which increases inventory loss and delays ECC systems initiation slightly.

Based on the above results and the discussions provided in Sections 2.2, 2.4 and 2.6, we find the increased setpoint on the SRVs acceptable. l 2.4 Overpressure Analysis The overpressure analysis for the MSIV closure with high flux scram, which is the limiting overpressure event, has been performed in accordance with the requirements of Reference 8. As specified in Reference 8, the sensitivity of peak vessel pressure to failure of one safety valve has also been evaluated. We agree that there is sufficient margin between the peak cal culated vessel pressure and the overpressure design limit (1375 psi) to allow for the failure of at least one valve. Therefore the limiting overpressure event as analyzed by the licensee is acceptable.

2.5 Thermal Hydraulic Stability The results of the thermal hydraulic stability analysis (Reference 1) l show that the channel hydrodynamic and reactor core decay ratios at the Natural Circulation - 100% Rod Line intersection (which is the least stable physically attainable point of operation) are below the (1.0) stability limit.

Also, the licensee has proposed restrictions on operating in the natural l circulation mode. The restriction prohibits steady state operation I without forced circulation for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Since the reactor decay ratio is 0.574, and the start of an idle recirculation pump from natural circulation is prohibited unless the temperature difference be-tween the loop started and the core coolant is 150*F, we find the proposed time limit in natural circulation acceptable.

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2.6 Accident Analysis 2 . 6 .1 ECCS Appendix K Analysis l l

Input data and results for the ECCS analysis have been given in References l 1, 3, and 4. The information presented fulfills the requirements for such analyses outlined in Reference 8. l We have reviewed the analyses and information submitted for the reload and conclude that the MNP plant will be in conformance with all require-ments of 10 CFR 50.46 and Appendix K to 10 CFR 50.46 when: (1) it is operated with the "MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE" values given in Table 3.11.1 of Reference 1, (2) it is operated at a Minimum Critical Power Ratio (MCPR) equal to or greater than 1.20 (more restrictive) MCPR .

limits are currently required for reasons not connected with the Loss-of- l Coolant-Accident, as described in Section 2.2 and Section 2.6.3.

l 2.6.2 Control Rod Drop Accident The worst case control rod drop accident (CRDA) can occur under startup conditions when the characteristic parameters for the accident meet the.

requirements for bounding analyses described in Reference 6. This is adequate to show that the design basis of 280 cal /gm peak fuel enthalpy for a startup CRDA is met (Reference 8).

For MNP, the characteristic accident parameters for the worst startup CROA satisfy the requirements for bounding analyses as described in Reference 6. Therefore the postulated CRDA would be 1280 cal /gm which is acceptable.

2.6.3 Fuel Loading Error The potential fuel loading errors (FLE) involving misoriented bundles and mislocated bundles have been evaluated. The analysis of the fuel l loading error is discussed in Reference 6 and approved in Reference 8. l The limiting fuel loading error aCPRs were calculated by a conservative

, older GE analysis. For similar plants, FLE analyses have indicated  ;

conservatisms of approximately 50% in aCPR when the old approved analysis l is compared to the newly approved GE analysis (Reference 10). Even j though we recognize the large conservatism in the older analysis, the apparent conservatism in the calculated values of the older analysis methods cannot be quantified for this specific plant and cycle. In the interim, unless a new analysis is provided for Cycle 7, the accepted aCPR value for the fuel loading error must be based on the resu,lts of the older conservative analysis.

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.This ACPR value, when added to the safety limit MCPR of 1.07, would i result in an operating limit MCPR of 1.45 for the 8x8 and 1.40 for the l 8x8 R fuel . These results would indicate that if a FLE occurs this cycle, some of the fuel rods in the bundle could experience boiling transition j and fail .

The licensee proposed that based on the absence' of a FLE during their previous five refuelings, and their demonstrated excellence in core verification prccedures end inspection techniques, the imposition of a MCPR penalty based on a hypothetical FLE was not warranted.

Even though we agree that MNP has an excellent record and excellent core verification and inspection procedures, we cannot, based on industry wide experience using similar techniques, accept what is essentially the zero probability position proposed by the licensee.

To resolve this issue NSP subsequently proposed a means for the detection of abnormal fuel degradation should the postulated FLE occur. This means of detection will be accomplished by measurements of off-gas radioactivity levels at the steam jet air ejector. To assure that further fuel degrada-tion as a result of a fuel loading error will not continue, NSP has proposed an offgas limit based on the SJAE activity monitoring system.

This offgas activity will serve to warn the operator that there could be a FLE avent in the core. The proposed offgas limit, equivalent to 0.236 ci/sec (30 minute decay) for 15 minutes, would limit the activity release to less than or equal to the activity release expected from a single misloaded bundle. ,

If the offgas limits are exceeded, the operator shall make power adjust-ments necessary to increase the operating limit MCPRs to >l .45/1.40 (multiplied by the appropriate Ky) for the 8x8/8x8 R fuel to assure that the worst misloaded bundle would remain above the safety limit MCPR.

Continued operation of the plant would then be determined by the most limiting condition relative to the FLE MCPR values or the limits on offgas listed in the Technical Specifications.  ;

In addition to the detection capabilities and Technical Specification requirements, the core verification procedures will also be augmented by an independent verification by the NRC Office of Inspection and Enfortement.

t In summary, we find acceptable for this operating cycle a procedure requiring operating limit MCPR's as shown in Table S-1, to be increased to %_1.45 for the (8x8) fuel and yl.40 for the (8x8) fuel in the event of indicated fuel

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, s 2.7. Cycle 7 Requirements Because this is the first cycle that the Monticello (BWR-3) Plant will be using the (8x8 R) fuel with drilled lower tieplates, the licensee has proposed additional surveillance requirements. The requirements will l consist of "In-Service" monitoring to detect instrument tube vibration, l and E0C inspections of the channel boxes. The surveillance requirements l proposed by the licensee consist of the following:

1. During TIP calibration procedures unfiltered TIP traces will be l examined and evaluated for characteristics indicative of instrument tube vibration.

2. End-of-Cycle inspections of tne channel boxes will be performed. A minimum of 10. channel boxes will be inspected to determine the extent and nature of any indicated channel box wear.

3. Results of the inspections and evaluations which are not consistent with similar inspections performed on BWR-4 plants will be reported to the NRC.  ;

l We find the above proposed Cycle 7 requirements acceptable.

3.0 physics Startup Testing The safety analysis for the upcoming cycle is based upon a specifically designed core configuration. We also have assumed that, after reloading, l

the actual core configuration will conform to the designed configuration.

A startup test program can provide the assurance that the core conforms to the design. We require that a startup test program be performed and -

the minimum recommended tests are: 1

1. A visual inspection of the core using a photographic or videotape record.

2. A check of core power symmetry by checking for mismatches between symmetric detectors.

3. Withdrawal and insertion of each control rod to check for criticality and mobility.

4. Comparison of predicted and measured critical insequence rod pattern i for nonvoided conditions.

In the future, we anticipate requiring a description of each test  ;

sufficient to show how it provides asrurance that the core conforms to the design. The description is anticipated to include both.the acceptance criteria and the actions to be taken in case the acceptance criteria are not obtained.

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_9 In addition to the requirements above, we request that a brief written report of the startup tests be submitted to the NRC within 45 days of the completion of the tests.

l 4.0 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 j determination, we have further concluded that the amendment involves an i action which is insignificant from the standpoint of environmental impact l and, pursuant to 10 CFR 951.5(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.

5.0 Conclusions Based on our evaluation of the reload application and available informa-tion, we conclude that it is acceptable for the licensee to proceed with Cycle 7 operation of Monticello Nuclear Generating plant in the manner proposed.

We have reviewed the proposed changes to the Technical Specifications and find them acceptable. '

We have concluded, based on the considerations discussed above, that:

(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the change does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activi-ties 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.

Dated: November 6, 1978

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Re ferences

1. Letter, L. Mayer, Northern States Power Company, to NRC dated August 10, 1978 transmitting " Supplemental Reload Licensing Submittal" for Monticello Nuclear Generating Plant, Reload 6, NE00-24133, July 1978.

2. Letter L. Mayer, Northern States Power Company, to NRC dated Auoust 16, 1978.

3. Loss-of-Coolant Accident Analysis Report for Dresden 2, 3 and Quad Cities Unit 1, 2 Nuclear Power Stations (Lead Plant) NED0-24046, August 1977.

• 4 Loss-of-Coolant Accident Analysis Report for Monticello' Nuclear Generating Plant, NEDO-24050 September 1977. l

5. Letter, L. Mayer, Northern States Power Company, to NRC dated l September 9, 1978. l

6. General Electric Boiling Water Reactor Generic Reload Fuel Application, NEDE-240ll-P, May 1977.

7. General Electric Boiling Water Reactor Generic Reload Application I for 8x8 Fuel NED0-20360, Rev.1, Supplement 4, April 1,1976. j

8. Safety Evaluation of the GE Generic Reload Fuel Application (NEDE-240ll-P), April 1978,

9. Status Report on the Licensing Topical Report "Gcneral Electric Boiling Water Generic Reload Application for 8x8 Fuel," NED0-20360, Revision 1 and Supplement 1 by Division of Technical Review, Office of Nuclear Reactor Regulation, United States Nuclear Regulatory Commissinn, April 1975.

10. Letter, Ronald Engel, GE, to Darrell Eisenhut, NRC, Fuel Assembly Loading Error, June 1,1977. ,

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