ML20125A452

Safety Evaluation of Plant Conformance to Requirements of App K to 10CFR50 & Acceptability of Proposed GETAB-based TS

ML20125A452
Person / Time
Site:	Monticello
Issue date:	09/30/1975
From:	Office of Nuclear Reactor Regulation
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Shared Package
ML20125A447	List: ML20125A444 ML20125A452
References
NUDOCS 9212080353
Download: ML20125A452 (10)

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SAFETY EVALUATION 0F-THE

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• MONTICELLO CONFORMAtlCE'TO- -  ;

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j THE REQUIREMENTS OF APPENDIX K TO 10 CFR 50 AND. ACCEPTABILITY OF-t . . .. .

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PROPOSED GETAB-BASED-TECHNICAL SPECIFICATIONS ~'

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REACTOR SYSTEMS BRANCH

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; l Conformynce to all Requirements of Appendix K to 10 CFR 50_

1.0 i

On December 27,1974, the Atomic Energy Comission issued an Order for Modification of License implementing the requirements of. ,

i I 10 CFR 50.46, " Acceptance Criteria and Emergency' Core Cooling Systems for Light Water Nuclear Power Reactors. One of the ,

' requirements of the Order was that "...the licensee shall submit  :

i a reevaluation of ECCS cooling performance calculated in accordance j with an acceptable evaluation model which confoms with the pro-visions of 10 CFR Part 50, 50.46." The Order also required that '

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! l the evaluation shall be accompanied by such proposed changes in Technical Specifications as may be necessary to implement the j ' evaluation results. '

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On June 9,1975 the licensee submitted an evaluation of the ECCS performance for Monticello.lll An amendment requesting chan es

) to the Technical S ecifications for Monticello to implement he l i

1 results of the eva untion was submitted on August 4,1975.t2 The i

licensee incorporated further information relating to the details of the ECCS evaluation by reference to the Quad Cities Unit No.-2 submittal (3) on ECCS evaluation as an appropriate lead plant l '

i analysis to show compliance to the 10 CFR 50.46 criteria and i- Appendix K to 10 CFR Part 50. -The Order for Modification of License issued December 27, 1974, stated that evaluation of ECCS cooling

- perforpjance. pette b4 sed on the vendor's evaluation model as modified ,

j in accordance'with the changes described in the staff Safety Evaluatio'n c Report of the Monticello Nuclear Generating Plant dated December 27, 1974.'

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l The background of the staff revied of the General Electric (GE) ECCS j

i models ai ' heir application's to Monticello is described in the r

i Staff Sa1 .j Evaluation Report (SER) for that facility dated "

December 27,1974 (the December 27,1974 SER)issuedinconnection with the Order. The bases for acceptance of the principal- portions l

! of the evaluation model are set forth in the staff's Status Report i 1

of October 1974 and the Supplement to the Status Report of Hovember 1974 which are referenced in the December 27, 1974 SER. The December

! I 27,1974 SER also describes the various changes require 5 in the

) earlier GE evalu$ tion model.. Together the December 27,1974 SER and t

the Status Report and its Supplement, describe an acceptable ECCS evaluation model and the basis for the Staff's acceptance of the model.

The Monticello evaluation which is covered by this $ER properly

conforms to the accepted model. .

With. respect to refloo'd and refill computations, the Monticello l ,

analysis was based on the modified version of the SAFE computer code, with explicit consideration of the! staff recommended limitations, t

' as described in the December!27,1974 SER. The Monticello evaluation i

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' did not attempt to include any furthei credit for other potential changes which the December 27, 1974 SER indicated were under consideration by GE at that time.

' During the course of our review, we concluded that additional individual break sizes should be analyzed to substantiate the t

break spectrum curves submitted in connection with the evaluation 6

provided in August 1974. .

We also requested that other break locations be studied to sub-l stantiate that the liniiting break location was the recirculation line, i

The additional analyses (performed on the lead plant Quad Cities

• UnitNo.2(3)andincorncratedbyreference)supportedtheearlier

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submittel which concluded that the worst break was complete severence of the recirculation line. These additional calculations provided further details with regard to the limiting location and size of break as well as worst single failure for the Monticello design.

The limiting break continues to be the complete severence of the recirculation line assuming a failure of the LPCI injection valve.

We have reviewed the evaluation of ECCS performance submitted by Northern States Power Co. for Monticello and conclude that the evaluation has been performed wholly in conformance with the ree quirements of 10 CFR 50.46(a). Therefore, operation of the reactor would meet the requirements of 10 CFR 50.46 provided that operation is limited to the maximum average planar linear heat generation rates (MAPLHGR) of figures 3.11.1-A, s.11.1-B, 3.11,1-C,'3.11.1-D, and 3 11.1-E of the Northern States Power Co letter dated August 4, 1975(21, and to a minimum critical power ratio 01CPR) greater than 1.18.

However, certain cFanges must be made to the proposed technical specifications to conform with the evaluation of ECCS performance.

j The largest recirculation break, area assumed in the evaluation was 3.9 square feet. This break size is based on operation with a closed valve in the equalizer line between the two recirculation loops.

j Therefore a license condition must be added whicit prohibits reactor operation unless the valve in the equalizer line is closed.

The ECCS performance analysis assumed that reactor operation will be limited to a MCPR of 1.18. However, a more limiting technical specification limits operation of the reactor to a MCPR of 1.33 for

! < 7 x 7 fuel and 1.41 for 8 x 8 fuel based on consideration of a l .,

turbine trip transient with failure of bypass valves, my 9

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The Technical Specifications should report as an abnormal occurrence, operation in excess of the limiting MAPLHGR values, even if corrective action was taken upon discovery. We believe that such events should be reported in conformity with the Technical Specifications.

An evaluation was not provided for ECCS performance during reactor Therefore operation with one recirculation 1000 out of service.

continuous operation under such condition will not be authorized until the necessary analyses have been performed, evaluated and determined acceptable. .

The steamline break accident analys1 )as presented by the lice (byreferencetoQuadCities genericreviewofitE00-20360.{gjt25 4

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2.0 Techn'ical Specification Chances to Implement Conformance to~~

Appendix K to 10 CFR 50 The proposed Technical Specification Limiting Conditions of Operation present two limitations on power distribution related

! to the LOCA analysis. These are the limiting assembly maximum l average planar power density. HAPLHGR, and the minimum power ratio I limit related to boiling crisis, MCPR. The MCPR value used in the 4 LOCA analysis was 1.18 and this value is 'ess than the value determined from the transient analysis which will be incorporated I in the proposed Technical Specifications. The baser for esta-blishing the limiting value of MAPLHGR are indicated above.

The licensee did not include the equalizer line area in the LOCA 4

i analysis, therefore, the Technical Specifications must require that the equalizer line valves remain closed at all times *during reactor operation. The LOCA analysis did not address one loop operation, therefore, the Technical Specifications should not allow continuous operation with one loop out of service.

The LOCA analysis assumed all ADS valves operated for small line breaks with HPCI failure. The Technical Specifications should be modified so as not to allow continuous operation with any ADS valve out of service. As with other ECCS equipment one valve may be out of service for'7 days.

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b 3.0 Conclusions Regarding Conformance to all Requirements of Appendix K to 1RTR 50 i

On the basis of our review of the information provided by the

< 6 licensee for Monticello, we conclude that the safety analyses i are acceptable with respect to confonnance to all requirements

\; of paragraph 50.46 of 10 CFR Part 50 once the referenced MAPLHGR 1

and MCPR Technical Specification changes are incorporated.

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4.0 Evaluation of GETAB Based Technical Specifications 5

The GE generic 8 x 8 fuel reload topica1 is referenced for the description of the thermal-hydraulic methods used to establish the thermal margins. However, based on our review of this topical we have found the GETAB application description to be incomplete.

j Therefore, we have evalugted the Monticello thermal margins based on the NEDO-10958 report and plant specific input information provided by the licensee.

' The fusi cladding integrity safety limit MCPR for both the 8 x 8 and 7 x 7 fuel is 1.06. It is based on the GETAB s.tatistical analysis which assures that 99.9% of the fuel rods in the core are expected to avoid boiling transition. The uncertainties in i the core and system operating parameters and the GEXL correlation i

(Table 4-1 of HED0-20694)7 combined with the relative bundio power i distribution in the core form the basis for the GETAB statistical determination of the safety limit MCPR(8)The and are bases for theseThe acceptable. un-i certainties are reported in NEDO-20340 bundle power distribution used in the GETAB analysis eqnservatively -

• assumes more high power bundles than would be expected dpring-operation of the reactor.

In comparing the tabulatM <' i.s n' uncertainties for Monticello and those reported in NEM/ M w we have found only one tion for the TIP readings difference. The Manticepn H a se 1 Jes uncertainty is 8.7% wheree Uw 24 fE/uC-10958 report shows 6.3%.

The increase in uncertainty cm 9 Mice 110 is a consequence of the increase in uncertainty iri ttc c*asurement of power in a reload core. A TIP reading uncertainty of.ti.3% would be applicable if this were the initial- core. In both cases the TIP reading uncertainties are based on a symetrical planar power distribution.

The consideration of bypass flow has also been taken into account in the determination of the MCPR limit. Finger springs have been attached to the lower end fittings of the reload fuel in order to maintain the core bypass flow within the range of the bounding analysis. In the bounding analysis,12% bypass flow is assumed.

The uncertainty of this bypass flow is factored in the total core flow uncertainty that is used in the GETAB analysis.

The operating limit MCPR is based on the most limiting transient, a turbine trip without bypass from 90% power and 100% flow conditions.

The calculated decrease in MCPR during this transient is 0.27 for 7 x 7 fuel and 0.35 for 8 x 8 fuel. The resulting operating Itmit MCPR is 1.33 for 7 x 7. fuel and 1.41 for 8 x 8 fuel.

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l The required operating limit MCPR is a function of the magnitude  ;

and location of the axial and rod-to-rod power peaking. In. ,

determining the required MCPR, axial and local peaking representative '

of beginning of-cycle were assumed. That is. R-factors of 1.075 for 7 x 7 fuel and 1.102 for 8 x 8 fuel and an axial peaking factor  :

of 1.57 at a point 1/4 of the heated length below the top of the l

fuel were assumed. This is the worst consistent set of local and i axial peaking factors. During the cycle the local peaking and therefore the R-factor is reduced while the peak in the axial shape moves toward the bottom of the core. Although the operating limit MCPR would be increased by approximately 1% by the reduced end-of-cycle R-factor, this is offset by the reduction in MCPR resulting i from the relocation of the axial peak to below the midplane, 1

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, p 5.0 Conclusions Regardino Acceptability of GETAB-Based Technical Specifications The APRM scram and rod block setting changes suggested in Mr. Mayer's July 10, 1975 letter to D. L. Ziemann are not part of the GETAB-GEXL changes. A definitive s'. ability analysis has not been presented a

for the APRM scram and rod block setting changes so these changes

• cannot be accepted. However, the GETAB/GEXL changes are well

! documented and are highly desirable in view of the much improved data base for the GEXL over that for'the Hench-Levy MCHF correla-l-

tion. The proposed technical specification changes for_ in-

- corporating the GETAB/GEXL analysis are acceptable.

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(b References 1). Monticello Nuclear Power Station LOCA Analyses Conformance with j

10 CFR 50 Appendix K (det Pump Plant), June,1975.

a 2). License Amendment Request Dated August 4,1975, Monticello Nuclear Generating Plant.

3). Quad Cities Unit 2 Special Report No.15, Supplement C, April 8,1975, April 21, 1975.

t 4). Status Report on the Licensing Topical Report " General Electric Boiling Water Reactor Generic Reload Application for 8 x 8 Fuel," NEDO-20360, Revision 1 and Supplement 1 by Division of Technical Review, Office of Nuclear Reactor Regulation, U. S. Nuclear Regulatory Comission.

April,1975.

, 5). " General Electric BWR Generic Reload Application for 8 x 8 fuel,"

NED0-20360 Revision 1. November,1974.

6). " General Electric BWR Thermal Basis (GETAB): Data Correlation and Design Application," NED0-10958, 73NED9, Class I, November,1973.

7). " General Electric BWR Reload No. 3 Licensing Submittal for Dresden Unit 3," NED0-20694 December,1974.

8). " Process Computer Performance Evaluation Accuracy," and Amendment 1, NED0-20340 and NED0-20340 1. dated June,1974 and December,1974.

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