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,c-November 23, 1970 Docke't No. 50-220
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. 3, Report to the ACRS
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hi,kg m I O N 3 NINE MILE POINT NUCLEAR STATION l4 - Y 3
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t i U. S. Atomic Energy Commission Division of Reactor Licensing n i. i (m. fg .;..... 1
's~ 42 0332 890413 FFHCHAL USE ONLY P OIA :_ -
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1 '. OFFHCHAL USE ONLY i ( ) ABSTRACT j
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The Niagara Mohawk Power Corporation has requested an increase in the
; licensed power level of the Nine Mile Point (NMP) Nuclear Station from )* 1538 MWt to 1850 MWt. Our evaluation is based on: review of " Technical ;
j Supplement to Petition to Increase Power Level - Nine Mile Point Nuclear Station" dated April 1970 and two amendments thereto both dated October 3 1970 submitted by Niagara Mohawk in support of its application to increase g power; review of operations at power levels up to and including 1538 MWt; and review of the startup test program results. We have also reviewed the transient analyses presented in the FSAR and the reanalysis of these
' l transients that was submitted in support of this application.
The proposed increase in pwer to 1850 MWt is based on use of the Hench-Levy heat transfer correlation rather than the Janssen-Levy correlation and minor modifications to the plant. The Hench-Levy heat transfer correlation was reviewed and accepted for the Dresden Unit 2, and we consider its application to NMP to be acceptable. The proposed plant modifications consist of incorporation of reactor scrams from turbine trip and turbine control valve fast closure, the installation of one i additional safety relief valve, increasing the normal reactor operating j pressure from 1000 to 1030 psig and making the corresponding instru-( mentation set point adjustments, f At the operating license stage of review, the capability of the NMP facility engineered safety features and the potential radiological consequences of accidents were evaluated at the stretch rating of 1779 MWt. The calculated radiological doses resulting from the analyzed accidents at the proposed power level of 1850 MWt will not be increased because the primacy coolant activity limits will remain the same and the allowable containment leakage rate will be reduced for the higher power level. Our evaluation of the performance of the emergency core spray cooling i, system at the proposed power level of 1850 MWt has not been completed. We anticipate completing this aspect of our evaluation in time for j presentation at a subcommittee meeting. Provided that the emergency I core cooling system is found acceptable, we have concluded that there
; is reasonable assurance that the health and safety of the public will j not be endangered by operation of the NMP facility with the present core loading at steady-state power levels up to 1850 MWt.
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- c. / ~OFFHCHAL USE ONLY I 11 TABLE OF CONTENTS i
c-Abstract
- Table of ' Contents 11 j
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1.0 INTRODUCTION
AND
SUMMARY
l 2.0 SITE AND ENVIRONMENT 4 [ 2.1 General 2.2 Effluent Releases 3.0 REACTOR CORE CHARACTERISTICS 6 3.1 Core Performance 3.2 Thermal and Hydraulic Analysis for 1850 MWt I
-3.3 Reactivity Control 3.4 Conclusion 1 4.0 PRIMARY COOLANT SYSTEM 11
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i 4.1 Reactor Vessel 1- g. 4.2 Safety and Relief Valves
.l .,' 4.3 Primary Coolant System Piping 4.3.1 Primary System Leak Detection 4.3.2 Biological Shield Integrity 5.0 EMERGENCY CORE COOLING 16 6.0 INSTRUMENTATION AND CONTROL 16 6.1 Control Rod Drives -l > 6.2 Addition of Turbine Trip and Load Rejection Scram E 6.3 Modification of Rod Block Limit 7.0 ANALYSIS OF ACCIDENTS AND EXPECTED TRANSIENTS 21 .h 7.1 Accident Analysis f 7.2 Transient Analysis 1
8.0 STARTUP AND POWER OPERATIONS 23 f. 1 8.1 Startup and Power Testing to 1538 MWt ; 8.2 1850 MWt Power Test Program 9.0 TECHNICAL COMPETENCE 26 I I 10.0 TECHNICAL SPECIFICATIONS 26 I OFFHCHAL USE ONLY
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1.0 INTRODUCTION
AND
SUMMARY
l: On August 22, 1969, the Atomic Energy Commission issued Provisional Operating License (POL) No. DPR-17 to Niagara Mohawk Power Corporation
}, authorizing operation of the Nine Mile Point (NMP) facility at steady-6
) state power levels up to 1538 MWt. By application dated April 20, 1970, the Niagara Mohawk Power Corporation requested an amendment of its license to permit operation at steady-state power levels up to 1850 MWt.
l The NMP facility was designed initially to operate at a power level of 1779 MWt, but with minor modifications the f acility could be operated at a power level of 1850 MWt. The Niagara Mohawk Power Corporation chose f
; to operate the plant at 1538 MWt for a period of time during which the
- ( performance of the plant would be evaluated. Our POL review was based i
! on the 1538 MWt' power level; however, our evaluation considered the I
! capability of the plant engineered safety features and radiological
{ l consequences of accidents at the stretch rating of 1779 MWt. Based on I the use of the Hench-Levy heat transfer correlation rather than the 4 Janssen-Levy correlation, on which the initial power rating was based, {. the increased power level of 1850 MWt is proposed. The Hench-Levy heat l. transfer correlation was reviewed and accepted for the Dresden Unit 2 f. I and, therefore, its application to NMP is considered acceptable.
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/ : Initial-criticality of NMP was achieved on September 5, 1969, and j full licensed power level (1538 MWt) was reached on January 19, 1970.
The 100-hour full power demonstration run was completed on February 7, 1970. The initial low power and full power testing was completed sat-f .- I isfactorily. The observed thermal, hydraulic and nuclear performance of I the facility was within predictions. The discovery of cracks in a core spray nozzle safe end resulted in a plant outage beginning in March 1970 and ending in July 1970 following replacement of the core spray safe ends. I We have evaluated the NMP facility for operation at power levels up to 1850 MWt.with the present core loading. This evaluation is based on:
- p i review of " Technical Supplement to Petition to Increase Power Level - Nine
( Mile Point Nuclear Station" dated April 1970 and two amendments thereto tl j both dated October 1970 submitted by Niagara Mohawk Power Corporation
'i in support of its application to increase power; review of operations at power levels up to 1538 MWt; and review of the startup test program -f results.
Our review included examination of data from startup testing and ; power operations as reported by the licensee and by the Division of Compliance. The data were found to confirm design predictions and f supported the expected capability to operate at the proposed power o ( OFFHCHAL USE ONLY : 1
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l level of 1850 MWt. Review of recent experimental data regarding emer-
/~ lech gency core spray cooling and its effect on the adequacy of the NMP.
systems is'of special concern. Evaluation of this matter for the proposed h"d). E power level of 1850 MWt is still being pursued actively and has not been i f completed. We anticipate completing our review of this aspect in time i for presentation at a subcommittee meeting. i We have examined the reanalyses provided by the licensee of all anticipated operational transients affected by the power increase that i might be expected to result from any single operator error or equipment h 2 malfunction. The results show that the design and performance objectives
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will be satisfied at the proposed operation of 1850 MWt. In addition, ( the design basis accidents have been reexamined for the higher power level. { f The calculated radiological doses resulting from these accidents at the h l proposed power level of 1850 MWt will not be increased because the primary coolant activity limits will remain unchanged and the allowable b 7 r : ' containment leak rate vill be reduced for the higher power level. Two physical plant modifications will be accomplished prior to
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- f These consist of the 1 increasing the NMP power output to 1850 MWt.
incorporation of a reactor scram on turbine trip, and the installation of one safety relief valve, thereby increasing the total number of valves i i t OFFHCHAL USE ONLY
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OFFHCHAL USE ONLY l . from 15'to 16. In addition to these modifications, changes to the plant :Mk
; will include increasing the normal reactor operating pressure from 1000 i . +A i to 1030 psig and making the corresponding instrumentation set point adjust- hg f' ments. /#
P Several changes to the Technical Specifications are necessary to I reflect the increased power rating. In addition to these changes, we plan to update other technical specifications, notably in the area of 1 effluent release limits. Provided that the NMP emergency core spray cooling system is found to be acceptable for the 1850 MWt power level, we have concluded that { I. there is reasonable assurance that the health and safety of the public 5 ( f will not be endangered .by the operation of Nine Mile Point Nuclear 1 Station at steady-state power levels up to a maximum of 1850 MWt.
', 2.0 SITE AND ENVIRONMENT
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,! 2.1 General L-4 The calculated consequences of the four design basis accidents (DBA),
4 as presented in the FSAR and evaluated in our earlier report to the ACRS A dated March 24, 1969, were based on a power level of 1779 MWt. These i accidents have been reanalyzed for operation at a power level of 1850 MWt. Because the applicant Proposes an appropriate reduction in the allowable 4
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( -S-con'tainment leakage rate, the potential calculated radiological doses are not changed.. We have, therefore, concluded that the present site is suitable for operation of the NMP facility at power levels up to
- 1850 MWt.
.i 'f At the time of our review of NMP for a provisional operating license,-
l wo informed Niagara Mohawk that it should install strong-motion seismd-graphs at the facility. Subsequently, we were informad by Niagara Mohawk that a strong-motion seismograph would be installed at NMP prior to i i the first ~ major refueling outage. We cont 31 der that this matter is being resolved satisfactorily. k' 2.' Effluent Releases
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( The concentrations of radioactive effluents released from startup through ~ June 1970 have been well below the limits set forth-in the Because of the March to July 1970 outage, [ Technical . Specifications .
! an equivalent of only about one full-power month of operation wcs ! accumulated through June 1970. During this period, a total of approxi- '[t- mately 5 curies in liquid waste and approximately 250 curies in gaseous waste were discharged. The 31guid waste activity released represents
{ t about 20% of the concentration limits specified in 10 CFR 20 for c-unidentified gross beta activity, but represents less than 0.1% of the j- 10 CFR 20 concentration limits for individual isotopes. The total l 4 l 1 OFFHCHAL USE ONLY l
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~(. gaseous activity discharged represents less than 0,01% of the corresponding 10 CFR 20 concentration limits. 'I 3.0 REACTOR CORE CHARACTERISTICS i
j' 31 Core Performance
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The core thermal and hydraulic performances were evaluated at various i
& power levels during the startup testing program. Tests were performed at power levels of 387, 786, 1125, 1283, 1495 and 1538 MWt with two y y
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' control rods decoup1'ed and valved out of service. The data obtained at the 1538 MWt power level gave a calculated heat flux of 281,000 Btu /hr-ft2
.x for the hottest rod based on the Local Power Range Monitor (LPRM) located
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nearest to the point of maximum heat flux. A Minimum Critical Heat Flux { Ratio (MCHFR) of 2.58 at 120% power and a total peaking factor of 2.60 f' t' were calculated from these data. The corresponding design values given l
' in the FSAR are 299,000 Btu /hr-ft2, MCHFR 2 1.5 at 120% power, and total f peaking f actor of 3.08; thus, the results demonstrate that the reactor
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.; core operated within the thermal and hydraulic limits on which the Technical Specifications are based.
3.2 Thermal and Hydraulic Analysis for 1850 MWt kl Operation of the NMP reactor at 1850 MWt with rated recirculation M { flow results in thermal and hydraulic core parameters equivalent se Y, to tho (4 0. C. M%~ tr %y*W~) , f OFFHCHAL USE ONLY
l OFFHCHAL USE ONLY { of Dresden Unit 2. The increased reactor power rating is achieved pri-marily by using the more recent Hench-Levy heat transfer correlation (APED-5286, September 1966) in lieu of the Janssen-Levy correlation (APED-3892, April 1962). This change in heat transf er correlation l results in a change of the thermal-hydraulic limits principally for recirculation flow rates in the range from 50% through 100% of rated; however, the margin in power between the limiting safety system settings and the safety limit within this ' flow range remain nearly the same for 1850 Wt as they were for 1538 Wt. For flow rates between 20% and h 50% of rated, these margins are reduced sligPtly; however, the margins i y over this range are at least as large as those for flow rates between I; 50 and 100% of rated. We reviewed the use of the Hench-Levy heat i
! transfer correlation in connection with our operating license review of Dresden Unit 2 and found it to be acceptable. Table 1 compares 5
j the thermal and hydraulic data for NMP at 1538 and 1850 We with those 7 of Dresden Unit 2 at 2527 Wt. 1 I i OFFHCHAL USE ONLY
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TABLE I
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COMPARISON OF NMP WITH DRESDEN-2 THERMAL'AND HYDRAULIC DESIGN PARAMETERS,
.> Nine Mile Point.
I'* Current: Proposed Design p- Rating Dresden-2, Parameter Rating g Power level, MWt 1538 1850 2527 No. of fuel bundles 532 532 724 4 Average power per bundle, MWt 2.9 3.5 3.5 MCHFR 2.1.5 @ 120% ? 1.9 @ 300% 2.1.9 @ J 00% , power power power ! Hench-Levy Hench-Levy.
- Heat transfer correlation Janssen-Levy s
Average power density, kW/ liter 34.1 41 41 f ( 0' Maximum linear heat generation 13.1 17.5 17.5 rate, kW/ foot N. - 109 131 13?. Average. hest flux, -- j 103 Btu /hr-ft2. 335* . R. 299 - 400 405
- Maximum heat flux, C l- 103 Bru/hr-ft2 }
i 3600 4250 4530 Maximum center fuel temperature,~*F Peaking factors 1.30
. Local 1.30 1.30 1.57 1.57 1.57
'? Axial 1.50
' Radial 1.51 1.50 Total product 3g8 3.0g 3.0g 98 I
I Core recirculation flow, 106 lbs/hr 67.5 67.5 f Steam flow rate, 10 6 lbs/hr 5.99 7.29 9.95 26 31 30 Core average void fraction, % ( Reactor pressure, psig 1000 1030 1000 OFFHCHAL USE ONLY a
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(' - From Table I , it is apparent that the proposed design rating ecre
. ;- parameter values for NMP at 1850 MWt are equivalent-to those of the s
gi Dresden Unit 2 'and, therefore, represent no extension of BWR themal and
.i* hydraulic operating limits.
I-We have reviewed the analyses of the various transients that can I ' be expected to occur during the operating lifetime of the. plant at the proposed stretch power level. This review consisted of comparing the results of transients analyzed for the 1538 MWt oparation with the resulta
-of the reanalysis of all transients affected by the. increase in power
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and corresponding increase in reactor operating pressure from 1000 psig
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to 1030 psig. Transients of concern are those resulting from cotitrol (- _ 0 I p rod withdrawals, changes in recirculation flow rate, additf,on of cold L 3 water and changes in system pressure frem actiona such as isolation.
- For all of the transients, the same criteria for iuel damage are
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l utilized. Fuel damage limits are detirad by MCHFR of 3eas than 1.0 and maximum linear heat generation rate of greater than 17.5 kW/f t. p.
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L turbine trip was added and one safety valve was added to the system. l.
- For all of the transients 2nalyzed, the MCHFR rqaains well above unity. .[
1 3.3~ Reactivity Control {' The equipment and systems for reactivity control of the NMP reactor, 1 i
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such as contro'1 rods, control rod drives and hydraulic system, and 4 standby liquid control system, are not significantly affected by the
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4 h proposed increase of power level. The shutdown margin was demonstrated
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to meet. the requirements cf the Technical. Specif1;:ations during startup
.4 testing. The demonstration of sh 2tdewn margin verified that the reactor
.g was suberitical with the central rod completely withdrawn and an adjacent rod withdrawn to a position equivalent to 0.?.S percent delca k, as required i 'by the Icchnice.1 Specifications. In addition, control rod withdrawal F
1 sequences were run to verify that the maximum control rod notch worth
-j-is 1 css than 0.1 percet delta k per notch.
t 4 3.4 ConcItsion We have concluded that the HBP reactor . core can be op2 rated safely 4 at power levels up to 1850 KJt 'on the basis o.E the cat.isfact.ory core 44 j perfor aance resf ts from startup t.osting and on the basia of th? pre-viously acceptable e.pplication of the Hench-Levy heat traasfer corre-A istion to Dresden Unit 2 ooiling water reacter. N,., . l, -. n 1-
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, j. Our review of the reactor primary coolant system included evaluation h of the effect of the power increase on the reactor vessel, on the number of
.t 4 . safety valves and on the operating performance of.the primary coolant 3
I ~ system experienced to date._ In evaluating the operating performance of-f the primary ' coolant system, we considered the recent problem with furnace-sensitized stainless steel components, the resultant inservice inspection program, the concerns with the structural integrity of the biological shield, and the current status ' of primary system leak detection capability. , 4.1 Reactor Vessel u.
- The reactor vessel is made of SA-307,_ Grade B, carbon steel and I the calculated lifetin,e neutron fluence on the vessel wall is reported L
as 5 x 1017 (energies greater than 1 Mev). The FSAR presents a curve
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showing the. corresponding end-of-life Nil Ductility Transition (NDT)
- f. temperature of 40*F. In addition, the licensee also considered a " worst" i
I' case by tripling the neutron fic.ence and using a " worst curve" for 4 predicting the temperature shift; the upper bound NDI temperature thus obtained was 150*F. The proposed power increase from 1538 to 1850 MWt a would change the lower NDT temperature estimate from 40' to 42*F and
'l the upper bound estimate from 150 to 160*F. We have concluded that f:
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[ OFFHCHAL USE ONLY b , ; this effect.is negligible considering the uncertainties inherent in these j i estimates. h: -4.2 Safety and Relief Va3ves The transients associated with operation of the safety and relief j-valves have been reanalyzed by the licensee for a power level of 1850 tnt. It' The ' turbine trip transient analysis, which assumcs a f ailure of the bypass valves to open and resultant actuation of the safety valves, demonstrated the adequacy of the capacity and. set point requirements for compliance -I with ASME Boiler and. Pressure Vessel Code Section 1. The design basis t
; {. for these valves is to provide adequate pressure relief for the transient associated with a turbine trip without assuming reactor scram, operation
( of solenoid-operated relief valves, or operation of bypass valves to 1 s j' prevent overpressurization of the system. The transient analysis of 4 safety valve actuation shows that the addition of a sixteenth safety
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valve is necessary to meet tl sriginal design criteria at the proposed 1
-l 1850 MWt power level. The applicant will install an cdditional safety valve set to relieve at 1218 psig with the other fif teen valves remaining at their original set points which are at or above 1218 psig. Since the basis for sizing the safety valves for NMP has not changed, we have concluded that the safety and relief valves provided for 1850 MWt oper-l j ation are acceptable for their intended purpose.
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OFFHCHAL USE ONLY ( 4.3 Primary Coolant System Piping On March 2, 1970, the NNP reactor was shut down for repairs to the
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main generator exciter reduction gear. On March 6, 1970, during routine i drywell inspection, primary coolant 3eakage was observed from the west 1 core spray nozzle safe end. At this time, the equivalent of approximately
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one full-power (1538 MWt) month of power operation had been accumulated. The ensuing investigation of the leak, which proved to be the result of cracks in the safe end, included evaluations of materials, stress analyses and metallurgical examinations. The following Niagara Mohawk g reports documenting the results of the investigation were provided to 3 J the Committee: f ( { (a) "Peactor Primary System Investigation at Nine Mile Point
! Nuclear Station" dated May 1, 1970.
' " Reactor Primary System Investigation at Nine Mile Point (b) i Nuclear Station - Report No. 2" dated May 11, 1970.
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. (c) " Program for Restoration to Service Based on Reports of i
'. Primary System Investigation - Nine Mile Point Nuclear
- Station" dated May 11, 1970.
I The results of the investigation were discussed with the ACRS at its l l June 1970 meeting prior to NMP's resumption of operation. f ( OFFHCHAL USE ONLY
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l Authorization to resume operation of NMP was granted on June 19, 1970, and included requirements consistent with the ACRS recommendations. 1 { Discussion of these recommendations follow in Sections 4.3.1 and 4.3.2. The NMP facility was returned to operation on July 6, 1970. I 4.3.1 Primary System Leak Detection Systems are presently installed to detect reactor coolant system leakage within the primary containment by measurement of the sump liquid accumulation rate and the dew point of the con-tainment atmosphere. As recommended by the ACRS, a third leak g detection system will be installed. This system will recir-t j culate a portion of the primary containment atmosphere through
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( Samples an external loop by a positive displacement type blower. f 1. will be drawn continuously from this loop through an air radiation monitor having a belt-type filter and an alarm. The licensee has informed us that delivery of the equipment for this system
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is expected by early 1971 and that installation will be accomplished at the first convenient outage following delivery. f. I The current operational experience with the two functioning leak detection systems indicates that the sump accumulation rate method will detect leak rates as low as about 0.5 gpm f! and the dew point system may detect rates of 0.5 to 1.0 gpm. I ( OFFHCHAL USE ONLY ;
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- Operating experience with the dew point system is limited and NMP operating personnel do not' consider that its reliability .s
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.has been established. We will. continue to follow the details of this program and maintain our requirement for demonstrating f'
a sensitive leak detection capability. I 4.3.2 Biological Shield Integrity The biolosical' shield consists of an approximately 24-foot diameter cylinder attached te the reactor vessel support pedestal and extending upward about 45 feet. The-Committee raised a question regarding the capability of this shield to withstand
.I the pressure that could be developed as a result of f ailure of r
( a nozzle safe end. This matter has been analyzed by the licensee 1- and our evaluation is based on the information submitted in the t y Second Addendum dated October 1970 and on a meeting with the
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I- applicant on November 4,1970, and subsequent telephone con-3
-t j versations. We consider the biological shield wall to have been adequately designed for the postulated Loss-of-Coolant l Accident (LOCA) which produces 40 psi in the reactor vessel cavity. The applicant has determined that the shield wall f
has the capability to withstand 96 psi from the standpoint c OFFHCHAL USE ONLY
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I 16 - 1. of strength and that missiles that could violate containment f < integrity would not be generated. We agree that this design I provides a sufficiently conservative margin when compared I. . . ..
.with the 40 psi design load assumed by the applicant. Local i-p.
stresses at the top and bottom'of.tht ' wall', at its support.
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points, have been analyzed by the applicant and found to be
.within design allowable stresses (133% of normal code allow-s i able stresses). - We find this' design approach and the stress
' limits under the postulated loading conditions acceptable,
- j. 5.0 EMERGENCY CORE COOLING Recent evaluation of experimental data concerning the efficiency of
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( ~. FLECHT tests, has raised questions
- core. spray cooling systems, i.e.,
2 -regarding the validity of the prior analysis of ECCS capability at Nine Mile Point. This matter is under active review. We will report the
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.'- - results of our evaluation of this matter at the subcommittee meeting. [
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6.0 INSTRUMENTATION AND CONTROL
} Control Rod Drives
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The startup testing of control rod drives included timing the com-f .plete withdrawal'and insertion performance of all 129 control rod drives. Control' rod scram insertion times were also ascertained to be within the Technical Specification limits. As the testing program progressed, the
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- control rod scram insertion times increased although not to the limits in the Technical Specifications. This progressive increase in control <l1 rod scram insertion time resulted in a reactor shutdown on November 20, 1969, to remove and replace the 400-mesh inner control-rod drive screens
[- with strainers having rectangular openings 10 by 100' mil. 6 During the shutdown for control rod drive screen replacement at NMP, index tube deformations were observed at Oyster Creek. The cor-rective action recommended by General Electric was to reduce the accu-mulator. pressure. The nitrogen precharge pressure on the NMP accumulators was reduced from 800 psig to 575 psig (with concomitant reduction-in k operating ' accumulator pressure from '1400 psig to 1100 psig) . On a-g' December 9, 1969, the NMP reactor was returned to operation and control i i rod scram insertion measurements showed that the time for 90% insertion of-all rods averaged 2.75 seconds which is well within the Technical l' . Specification limit of 5 seconds. These rod' insertion times did not 5 1 change significantly during the operating period December to February f 1970. The licensee stated that the performance of eight selected
-f control rod scram times would be measured periodically to ascertain
.I if any deterioration of the system has occurred. The licensen sub-
- sequently informed us that these tests have been performed twice to date a
j and that the times have averaged 2.8 seconds for 90% insertion. We s have concluded that the control rod surveillance program is adequate
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to monitiir control rod drive performance and that no new safety con-I siderations, in this regard, are introduced by the proposed increase I; i in power. i i~ 6.2 Addition of Turbine Trip and Load Rejection Scram To reduce the severity.of the transient resulting from a turbine trip in the event of failure of the bypass valves to open, two additional scrams-will be installed. These scrams are similar to those installed in other BWR plants (Dresden-2 and 3, Millstone) which are designed for a comparable or higher power density than NMP at stretch power. 'These j : scrams are from two sets of instruments: (1) a turbine trip scram k = taking a signal from turbine stop valve closure, and (2) a load rejection g- _( [ scram taking a signal from the turbine control valve acceleration relay.
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'i The startup testing program included trubine trip tests at 384,
[ 768, 783'and 1538 MWt. The tests at 783 and 1538 MWt resulted in i reactor scrams as predicted from high pressure and high flux, respec-I tively. The maximum heat flux associated with the turbine trip at t 2 i- 1538 MWt was 247,000 Btu /hr-ft , compared with the predicted
't 299,000 Btu /hr-ft ; and the MCHFR was 1.96 @ 120% power compared with f the predicted design MCHFR limit of 1.5.
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OFFHQAL USE ONLY b The results of the turbine trip te.st described above give confidence in the licensee's . analysis of the same transient at the higher power level of 1850 MWt and in the determination of the safety valve sizing and 4, [.- .capac' i ty as described in Section 4.2 of this report. To assure that g. L. the MCHFR limit will not be exceeded during a turbine trip transient at 1850 MWt, scrams wi11 be initiated whenever the turbine stop valve reaches a position of $ 10 percent closed and when acceleration relays which sense the hydraulic oil pressure in the turbine control system
' indicate f ast closure' of the turbine control valve. The licensee's s
analysis.of the turbine trip transient shows that reactor scram would
.' be initiated within 10 milliseconds and the resultant flux and pressure
( peaks in the reactor would be less severe than those ascociated with f- operation at 1538 MWt without these scrnms.
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i. We have determined that the turbine stop valve closure scram circuits I I satisfy the provisions of IEEE-279. Further, this scram circuit and the I
- turbine control valve f ast closure scram circuit do not compromise the f
.i existing reactor protection systems. ,.
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! The load rejection scram provided by the turbine control valve fast ]
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- [ closure a..celeration relays can be defeated through a single failure;
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5, however, this scram function merely anticipates the action of the l
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n; [ .o redundant turbine trip circuits, and therefore we have concluded that its design is satisf actory for the p%ypose intended.
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An automatic bypass circuit is to be added as part of the modifi-2
,{- cation. This circuit will defeat the turbine trip and load rejection
's scrams below 45 percent of full power. Power is censed as a function 1
"i - of turbine first-stage pressure by four pressure switches. We have determined that the pressure switches are redundant and that they are properly grouped to achieve adequate physical separation. I We have concluded that the proposed addition of scrams from turbine l trip.and load rejection is acceptable and that the instrumentation proposed g .-
$; to accomplish these actions is adequate.
( 6.3 Modification of Rod Block Limit i~ The licensee stated in the "Second Addendum to Technical Supple-i ment - to Petition to Increase Power Level" that the Average Power Range f Monitor (APRM) rod block neutron flux set point is changed from 110%
'l to 106% of rated neutron flux for 1850 MWt operation. In discussions on this subject, however, the licensee states that he is in fact using I
[ a set point of 106% for operation at 1538 MWt and that documentation l: of the use of this setting would be submitted to the 7C . This change { was made as a result of an analysis of the flux monitor characteristics that i i OFFHCHAL USE ONLY
m n ---* - ._ ,,.., , n u f.- . OFFHCHAL.USE ONLY
-b 1
shows that the same protection is afforded at the 106% setting as that
. previously assessed for the 110% setting in that the MCHFR remains 21.0
- 4
'3-for the rod withdrawal error accident. We have ' concluded that this 3 _ change is acceptable. t 7.0 ANALYSIS OF ACCIDENTS AND EXPECTED TRANSIENTS
. 7 .1' Accident Analysis The four major postulated accidents (design basis accidents: loss of primary system coolant, steam line break outside the drywell, drop of a fuel assembly during refueling, control rod drop) have been reanalyzed for the proposed power level of 1850 MWt. For the main steam line break
..g-C accident, the offsite radiation doses depend upon the concentrations of radioactivity in the primary coolant system; thus, the increase in power T
. level does not change the calculated radiological doses since the limit on primary coolant activity is not changed. For the other three design basis accidents, the calculated doses are directly proportional to
).
, reactor power level and the app 31 cable containment leak rate. At the i time of our POL review, we calculated that the maximum two-hour doses at 4 >
' the exclusion area boundary would result f rom the loss-of-coolant accident at 1779 MWt and were 150 Rem to the thyroid and 6 Rem whole body. The licensee has proposed reducing the technical specification limit for containment leakage rate such that estimated offsite doses will remain the same for operation at 1850 MWt as those originally calculated and accepted for 1779 MWt. The proposed containment leak rate reduction l
1 OFFHCHAL USE ONLY
e
~
OFFECHAL USE ONLY I . is fromthe present allowable 1.6%/ day to 1.5%/ day at 22 psig test pressure. The results of testing to date show a containment leak rate h of about 0.8%/ day which is well within the proposed technical specifi-cation limit. Our earlier conclusion that the calculated radiological i consequences of the design basis accidents were within the 10 CFR 100 guidelines is unchanged. 7.2 Transient Analysis Niagara Mohawk has reanalyzed all anticipated operational tran-sients af f ected by the power increase that might be expected from any single operator error or equipment malfunction. The results show g.
-t j that the design and perf ormance objectives will be accomplished under
( the proposed operation at 1850 MWt. Confidence in these analyses is [ li
! gained by the fact that test results have confirmed the predicted design i characteristics of the NMP reactor as described in the safety analysis i
i report. The addition of scrams from turbine stop valve closure and/or i from turbine control valve fast closure will reduce the pressure tran-I l. sient resulting from the turbine trip. We have concluded that the safety criteria regarding fuel damage limits applicable to current operation are satisfied and that these criteria will also be satisfied for oper- [ ation at the proposed power level of 1850 MWt. OFFHCHAL USE ONLY
OFFHCHAL USE ONLY 8.0 STARTUP AND POWER OPERATIONS
; 8.1 Startup and Power Testing at 1538 MWt , -l
( As indicated above, the startup and power testing program results I served to substantiate design predictions. The core thermal and hydraulic 1
; performances showed that the core operated within the specified thermal t
and hydraulic limits. The transients resulting from recirculation pump trips at one-half, three-fourths and full power were ccasisr.ent with I predictions, for example, the 5-pump trip at 1538 MWt showed a transient MCHFR of 2.3 at 2.75 seconds after the trip which is consistent with g the prediction. For the turbine trips at similar powers, the data e j showed maximum heat flux and MCHFR values to be within applicable
.f predictions and limits. Reactor system stability measurements were f +
within applicable criteria. Control rod reactivity worth measurements
', and rod insertion scram times were satisf actory.
I j The cold vibration test data from NMP indicated actual vibration f i levels within design limits. Measurements taken hot at zero power l f- were also indicative of performance in accordance with design. Readings taken under power conditions became progress:.vely less reliable due to instrument deterioration. Niagara Mohawk has no plans for replace-ment of the internal vibration instruments. An external vibration L e l i l i \ l l ( ! OFFHCHAL USE ONLY l 1 1 _ - _ _ _ _ _ _ _ _ - - _ _ _ _ _ 1
~
~.,,.-.....e.
l' s l_ 7 , OFFHCHAL USE ONLY
.p H
. monitoring system using accelerometers is being developed at NMP in response to a' comment by the ACRS during the POL review. The purpose of this system is - to detect excessive vibration and/or the presence of loose parts in the reactor vessel. Preliminary results obtained f.
from the accelerometer systems installed at NMP and Dresden-2 indicate that this type of system is incapable of detecting low amplitude vibration of-react'or vessel internals. It appears that such an accelerometer system may be .2ble to detect vibrations which result from impacting. The licensee indicated that this program evaluation and development is being continued. 1 The four main steam line 1 solation valves have ben leak-rate i ( r i tested twice and their closure time has been determined with the reactor (
! at one-fourth power and at full power. Closure time tests were n=rformed f
i j during startup testing and the results showed that all valves closed t
. within the 10 seconds permitted by the Technical Specifications; the
{
. .! slowest time recorded was 7.8 seconds. Leak-rate testirg of these .[ valves was initially performed in conjunction with the primary and f
i secondary containment leak rate test in August 1969, prior to fuel loading. The test results were within Technical Specification limits. Following the difficulties experienced at Oyster Creek in maintaining 2
'{.
1 ( 1 OFFHCHAL USE ONLY
ie' OFFHCHAL USE ONLY [. the leak-tightness of these valves, we requested that the NMP valves .4 be retested. The valves were retested on' June 24, 1970, at which time [ one valve leaked in excess of the limit. Investigation revealed that
}'
the leaking valve did not seat completely bece ise of an improper i l' limit torque switch setting. Af ter. adjustment of this switch, the leak rates of the valves ranged from 0.7 to 8.5 SCFH with all valves being
~
within the technical specification limit of 13.8 SCFH. 8.2 1850 MWt Power Test Program i' The licensee proposes a powe.r' escalation test program similar to 1 that performed during the initial approa:h to power. A set of base I 4 conditions will be measured at 1538 MWt before power escalation is I' l ( initiated to serve as a basis' for comparison with subsequent tests. 1 These base conditions will include chemical and radioactivity levels
]
ji at typical locations, radiation measurements, APRM calibrations , LPRM response characteristics, power distribution measurements and a core i I . performance evaluation. During the power escalation program, these tests will be repeated at an intermediate step of about 1700 MWt and
- p. again at the full-power level of 1850 MWt. Tests at 1850 MWt will l include induced transients, such as a step change of the pressure g
regulator set point, opening of one turbine bypass valve, shutdown L i i ( OFFHCHAL USE ONLY
o ' OFFECHAL USE ONLY b ( i i of recirculating pumps and turbine trip, to determine reactor response. j We have concluded that the proposed tests will provide the information j necessary to demonstrate the adequacy of the NMP facility to operate i j~ at the proposed power level of 1850 MWt.
.I 9.0 TECHNICAL COMPETENCE b
I The operating organization, its qualifications and responsibilities, operating procedures, records, maintenance, and review and audit functions are not changed from those we found acceptable during the POL review. I There has been only one change in the supervisory staff of the oper-m ating organization from that previously reviewed. The general tech-nical performance of this staff has been shown to be satisfactory during i o ( l the startup and power operations to date. However, the period of j demonstration has been short since sustained power operation commenced 1 ( j in July 1970. The subject of demonstrated technical competence will l* be evaluated more fully in our review of the application for a full-i term license in 1971. 10.0 TECHNICAL SPECIFICATIONS ] I. Several changes to the Technical Specifications will be necessary f for the proposed power increase. These changes involve references to power level or parameters associated with reactor pressure. In addition l 1 1 i ( OFFHCHAL USE ONLY ~ - i
~'
'T ' OFFHCHAL USE ONLY I i'
to the changes directly related to the power increase, we plan to update i d . i the Technical Specifications in some areas by incorporation of the current I i i 1 I 'l requirements for effluent releases, testing of instrument line flow check valves, reactivity anomalies, environmental monitoring program and reporting
- 1. - requirements.
I 1 r
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, u - , - m -.
;L T - D'tt &dL M& w u e ik f.: ., UNITED STATES ATOMIC ENERGY COMMISSION y
? '"" i *
' (,!
e*I ' ' wassmorow, o.c. aous January 27, 1971
. s , w -s Docket No. 50-220
'.y 1
.IN Dr. Spencer H.' Bush Chairman, Advisory Committee
. ,i n on Reactor Safeguards
"' U. S. Atomic Energy Commission Washington, D. C. 20545 j .
/ .; ]
Dear Dr. Bush:
Sixteen copies of our Report No. 2 on the Niagara Mohawk Power Corporation's application for a power in-I ' crease for the Nine Mile Point Nuclear Station are f enclosed for review by the Committee. Sincerely, 4 b\ i M ,u. Peter A. Morris, Director
~4 I
Division of Reactor Licensing Enclosures
-l DRL Report No. 2 to ACRS l
j' t O v l' When separated from enclosure handle this document as 12iCIASSIFIED f [M/_ klM A wl DL .UA% M-,_ s 4c9,1
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..,, January 27, 1971 Docket No. 50-220 4
1 1
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Report No. 2 to the ACRS
. ;, . . 1. a
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. .l 4
l' NINE MILE POINT NUCLEAR STATION { Power Increase _
, {
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. 1 J
\
' U. S. Atomic Energy Conunission Division of Reactor Licensing i 2 i s
J am a = ne fw;g y HA=Ak %JQ & %uayant-S- l I
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- ABSTRACT _
)
- Our November 23, 1970 report to the Committee, regarding the increase in the licensed power level of the Nine Mile Point Nuclear Station from 1538 We to 1850 We,
" indicated that our evaluation of the performance of the emergency core cooling system at the proposed power level of 1850 We had not been completed. We have performed extensive additional reviews of the calcula-
' tional models now used by GE and a model developed
%. ..
- independently by INC. Differences between the two models have not yet been fully resolved.
,. Q4 The applicant has provided assurance that the NMP core spray system can achieve rated flow reliably in 35 seconds
,.I or less, instead of the value of 60 seconds assumed in
;,'g7, j
'T M previous calculations. Taking this change into consid-eration, we have determined that, for a loss-of-coolant
.g accidsnt resulting from a recirculation line break, the peak clad temperature calculated by either the w '
GE method or the more conservative INC method is less i.y; ' than 2300*F. For accidents involving small breaks, GE calculates peak clad temperatures of 2226'F. M' ;1 ' We have concluded that the ECCS performance is accept-able for operation of Nine Mile Point at a power level d of 1850 W t. N y. .. . .-
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' - Oi?'$ L Q .M.k U@&a %va G -
. l EVALUATION OF ECCS FOR 1850 We
$ NINE MILE POINT j
In our evaluation of the Oyster Creek power increase to 1690 We, we used the FLECHT BWR test results (Zr-2K) as an experimental basis for assessment of the adequacy of the ECCS. The indicated test conditions c, for the aircoloy bundle test (Zr-2K) were essentially upper limits for
~v, the 1690 Wt power level, so that the test results could be regarded
~.
l as a demonstration test. In that case, we did not require additional j analytical model formulation as a basis for accepting the adequacy of j
- A the ECCS for the Oyster Creek application.
The power increase to 1850 Wt for the Nine Mile Point reactor, l however, appears to exceed the test conditions so that a similar i approach is not possible. The relationship of the Nine Mile Point conditions to the FLECHT test (Zr-2K) conditions is indicated in I the accompanying tables. The total power levels and some of the spray initiation , temperatures obtained in the FLECHT bundle test are exceeded by those 1 I expected in the Nine Mile P' o' int core at the higher power level, although f 3 the linear power density appears favorable. On balance, the indicated dif ferences between the conditions of the test and of the Nine Mile a Point reactor at 1850 We require a systematic extrapolation of the test ! results by means of an appropriately developed model formulation based on the full rangi of experimental data available. GE had previously presented a spray cooling model development for the Oyster Creek application for 1690 We based on the FLECHT 1 I 1 w,.- - - - . annt u Sb
=_-_:__-__
-.-_ _ _ . _ wrmrrm c n ~~.~ ~n- my
%vf d' h% a ,. . .
l ) tests. At that time, we concluded that there had not been demonstrated sufficient bas'is on which to completely accept the GE model for spray 1 cooling. Several subsequent meetings have been held with representatives In addition,
.e of GE to discuss their model formulation in greater depth.
. independent efforts to develop a model have been initiated by INC.
4^ 9/@ Recently, a meeting was held at GE in San Jose between members of the
)*} regulatory staf f, INC, and GE to discuss detailed calculations obtairied from the models and to explore conceptual differences that have been form-
. ' ' 'w f.A ulated in the two models. Details and differences involved in the models
.i are discussed in the attached appendix.
Both methods represent attempts to produce rational and systematic The approaches to a very complicated thermal-hydraulic phenomenon.
.9 models have helped to understand the importance in the spray cooling
.'h phenenenon of certain parameters, such as channel quench time, channel
.e...
film coefficient prior to quenching, heat transfer to the spray fluid,
.)-}'S
> .h
t. radiation heat transfer, and' grouping of the fuel rods. However, several differences in the treatment of these parameters exist between the models that have not been resolved at this time. Numerical eval-uations of the Nine Mile Point reactor conditions have been made with
- both models for a recirculation line break with an assumed 60-second
- (, a initiation time for the spray system (i.e., achievement of rated flow). These evaluations predict a peak temperature of 2180'F by the GE method and e
4
.e
[: .
.e . - m. ~: e., n -y a a:a w.s_
J, %pi; w s /* \ [h I V ! ' ' '- - V :< h u s 3 i calculations and 2220*F by the INC method and calculations. This comparison of peak temperature alone, however, is not suf ficient to
! N.rf i 'q..
appraise the models as the time at which the peak temperature 4.s pre-dicted to occur may also be a significant indication of the intrinsic << .~ validity of the model formulation. In the case cited Love, the peak is predicted to occur 4 minutes after the break by the GE model and
.4
' . ;'E i
-10 minutes by the INC model, a substantial difference. A similar dif ference has been exhibited between the model predictions for the
[' " .
,r, . ~~ " Er-2K FLECHT data; the GE prediction underestimates the tia.e at which the experimental peak temperature occurs while the INC model appears to be substantially better in enis. regard. Both models predict the peak test temperatures reasonably well but the prediction of an earlier peak temperature is less conservative. -n. I If somewhat more conservative values of the channel heat transfer kk,*h'! coefficient (h) and channel quench time are considered (reducing h i.',.; '
from 20 to 10, and increasing the quench time from 3 minutes to 4 minutes), the'INC model predicts a peak clad temperature of about
' 2500*F. GE ha: made a maparable calculation with the same result, but
, .t GE prefers to calculate a "best estimate" by reducing some of the specific conservatism it.cluded in their calculation of a 2180*F peak.
The "best estimate" peak temperature is about 2000*F. If the reduced conservatism in this latter calculation were included in the INC cal-enlations, the conservative INC result would be reduced to approximately 2320*F. Although these countations are useful, the uncertainties in 1l MddE Of68" f' p a _~ - . _ _ _ _ _ _
. , - . . . , , .. ~
, ,,y v .. o .: ..
4
..e 1
! the rodel formulations require additional consideration.
1
.3 Rere are significant difficulties in properly modeling the complex phenomens involved in the spray cooling phase of the Nine Mile Point BWR.
The dif f erent methods of excracting and applying generalized heat transfer i
.i parameters from the FLECP.T tests as represented by the INC and GE approaches are not unreasonable, but some fundamental dif ferences appear to exist at
<j
; this time which have not yet been resolved. At present, then, we do not
.4 rely solely on the GE model for the evaluation of the spray cooling phase of the recirculation line break in the Nine Mile Point LOCA, but also rely on estimates of anticipated performance bar,ed on the more conservative results obtained by the INC calculations.
W e calculated peak clad temperatures can be reduced by taking into account the f act that the Nine Mile Point reactor ECCS is to be operated 3 and maintained so'that the core spray system can achieve rated flow in 35 seconds or 1. ss instead of the value of 60 seconds used in the calcul-
- 6 I ation. Representatives of the Niagara Mohawk Power Corporation have assured q
the staff that the required startup and operation of the diesels for this
.l
! Under these conditions, we have concluded purpose is feasible.and reliable.
I that the peak clad temperature calculated by either calculational model ( will remain below 2300*F; GE estimates a peak temperature of approximate 1:r
! L I
I 2000'F, and INC approximately 2200'F. l On the basis of a spray system which achieves rated flow in 35 seconds or less, we conclude that the ECCS for Nine Mile Point reactor sh?uld provide adequate core cooling performance in the event of a recirceJacion line br2ak at power levels up to 1850 MWt. M . M. C " 4 E E T_ n-O
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_ _ . . . , .m -, y
%'w . -r~. .,
- 5
\'ls1 b,'d In Amendment No. 5 to the application for power increase (received
- 1/26/71), the applicant has documented calculations by GE for small-break i
LOCAs. The calculated peak clad temperature is 2226*F. We are reviewing
- these calculations and will report our conclusions orally to the sub-committee.
iC
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I APPEN'0IX
&;4 NDE !GLE PODE IDC.A 70. _0_.1... .
-w For Nine Mile Point, the loss-of-roolant accident is divided
.'., -1 into thrve tir.e periods: (1) blowdcym, (2) core teatup, (3) core i,-
%. a '
spray. 7be blowdown and core heatup periods of the LOCA pasented
.. .e j by E appear to be reasonable. That is,, the use of the 1.8 second
%'s- .
dryout thre, the experimentally based dryout rest trEnsfor, and no cIsdit for steam cooling are warrarted dr. this case. 'iha deriva-
- l ..
' ':P tion of acceptable core spray models and their application to core
. f f.' .' [
heatup calculations is of particuhr concem $n tMs plant. Sm i
#~ I. .Cr4.E SPRAY !ODELS
~
T Sta;t3ng with the MECHT SS-2N. data, Censtul Electric and icato Nuclear Cwpotution have embar!cd on similar procedures. u , a. . 5
,3 Each has dew 2cpet cuputer programs to "enract generalized
- x. ,; . '
heat transfer t._c. meters from the data. 7tese parameters were then
'. "iN I
I used to develop heat transfer correlations, hese correlations for f,<),. s heat Mr Par coeffiolent and channel wetting time are used in
-:.r.:
. ., J N, ,
com.rbr cde to predict the msults 6f the FIECHT '.tainless and
* ( '. ,
cirqaloy te'acs, as well as the IDCA for NMP. u.
..m..2--)
[p.tgitjyjare Spray Heat Transfer Parameters [.. 3 A. a,e first step in E's extraction procedure is to calculate
;?9. .
.g ppey bcdy factors for the 49 rod ph2s channel box array for a
..g-connant emissivity using their GhEY code (for SS-2N, ( =0.6).
hy.p p..
' itese grey body factors are then cambined to match the desired g
groupirg, of rods for which grey body factors are r.eeded. l A
^
T _^ G j T't
- dgT*_% -wwas va vna B
.. g p,- - -- ---.'. .}
i va a~aw%d wa m ku a ~l s ..
- 2-e Next, the code FID00 is used to solve the rod energy
]
I balance, the only unknown now being the heat transfer coeffi- ) cient, since rod surface temperatures am known and the fluid temperature is assumed to be at saturation. FID 00 neglects
- - r conduction and does not calculate chanm1 box heat transfer
- coefficients. For extracting coefficients from SS-2N the 16
&.y " northwest" rods in the bundle were considered and gmy body
-f.
factors cabined accordingly.#
.. }
i
'Ihe INC extraction pmcedure begins by using the DATAR o - J,- code to solve for a " total" heat transfer coefficient from each rod and channel surface. 'Ihe coefficient includes all It is energy leaving the surface including any radiation.
i the total heat flux divided by the tenperatum dmp between
, . ,. j the surface and the same saturation tenperatum that GE uses.
1 l
,- ), An inverse conduction model is used in DATAR. Sirce all W ', 1 experimental temperatures are not recorded simultaneously, the values used in ' calculating are interpolated from the data to the time desired. .
8
'Ibe RADff code uses the DATAR results and the experimental !
surface temperatures to extract coefficients from surface to fluid. 'Ibe rod matrix used in RAnlT is the entire bundle plus a
* '1his entire procedure is explained more fully in the appendix h
of GEAP-13086 " Heat Transfer in a Simulated BWR Fuel Bundle g Cooled by Spray Under Icss-of-Coolant Conditions," June 1970. u MM A W WWM M h Wjy wi
. .was uns Va#G w=w "
l L
's ~A,M& Q u**Cj,;,Y, _ ^ ' .
t* l
- 1 I-I .
i f the channel box. If temperatures are not available for each rod, its " mirror image" along the northwest to southeast diagonal is
9 used. A temperature-dependent emissivity is progransned into
'" RADHT and the matrix solution for body-to-body radiation solved for each buiy for each time step. The emissivity calculated f;"if.j _
i for stainless in RADHT is almost always about 0.89.
*1 The INC method is somewhat more rigorous than the approach l
adopted by GE since it includes a conduction model, a temperature interpolation routine, a variable emissivity, a radiation a solution for the entire matrix for each time step, and extracts g j channel box heat transfer coefficients. In general, the INC method yields lower h's than th;. GE method especially for the
] ,
outer rods and the channel box. For these reasons, it seems a 5 more conservative method and, therefore, preferable at this time. , B. Correlation of Extracted Heat Transfer Parameters 1
' GE was able to correlate their results as a temperature
- function (Roger's correlation) strongly dependent on rod loca- l tion, the outer rods yielding higher coefficients. INC did not I i
find a strong temperature or location dependency but rather a i i l dependence on channel wetting time. Af ter can quench, INC h's were between about 0.5 and 1.5; GE values were between
)
I about 0.5 and 10. The ability of GE to correlate well with the temperature function (T4 - T,4)/(T - T,) indicates a 4 I 1 l
. ,g;(C= . -c z.x y_ , ,-
-, =_ _ uwu -- -m;ewa
. .I ,
. .s
- strong radiation component of the heat transfer coeffic.ient (T, is the sink temperature, assumed to be saturation). 'this component is too large to be accounted for as radiation to steam or droplets, so GE claims it is radiation to the water j film around the wetted channel. Even though some radiation is now absorbed by water surrounding the channel, an appro-priate decrease in the amount of radiation to the channel te not made. That is, body-to body radiation is not properly decreased to account for radiation to the film. Conservation of radiant energy requires that the sum of all shape f actors 1
for a given body to all other bodies is 1.0. This is main-tained prior to chsnnel wetting. liowever, af ter chann-e )
. .,/ . .
wetting, the radiation portion of the Roger's correlation H causes the sum to be larger than 1.0 for all tod groupee 1%C
. . f. ;
estimates that in the GE case of a corner rod, the sum of
..4 ..
the shape f actors can be as high as 1.5 to accommodsb the additional rod-to-film radiation. It can be art,jnh eusever, that a high total view factor is compensation for a .kw value of er.issivity. But without a calculation to dehstratu that hypothesis, the issue remains unresolved. Tlu proce&.'rt. of
~ 5,, q allowing an excess amount of radiant energy to br. transferred to a constant sink appears non-conservative.
..%'a It!C did a parametric study of the effect of e.missivity <
using their method. It appears that about I.alf the difference
- a.
$$'t .
ah
- l I
n~~._ - =_-- --__
"M a AMumarma . -l__ + m e l
t, .
. . W/
., I in extracted h's between the GE and INC methods can be attri-buted to the values chosen for Laissivity. The intrinsic
*i method of extraction t. hen appears to account for the balance of i *
'.' the difference.
GE has correlated channel wetting time as a function of M3.,. t- 4 a Yamanouchi* parameter containing channel wall temperature and temperature gradient. The GE correlation with this para-
. i meter is linear, although the data are insufficient to be
.at a M:W,
" definitive and a less f avorable correlation conceivably could be supported by the limited data. On the other hand, the approach developed by INC requires that the wetting time be i
treated arbitrarily and handled parametrically in their pre-
'I dictions. In formulating the heat transfer correlation af ter
..4 spray initiation but before channel wetting, SE calculates the radiation to the water film component and uses a fraction of the fully developed coefficient as expressed by the ratio of (t - t,)/
l (t g- t,), where t is the time at which the coefficient is being calculated, t, is the time of spray initiation, and eq
~ is the channel quench time. Note that t,4.t4.tg . Neither the INC nor the GE data correlate to) well with time or temperature l before channel quench. This is of special concern where tempera-ture turnarounds are predicted to occur at about t.he same time l
I *Ya .anmchi, A. , "Ef fects of Core Spray Cooling and Stationary State Af ter LOCA", Journal of Nuclear Science and Technology, October 1968. l t
=- - . - y_m n -gy
~ . d% . . .- =
%#lDu a d " ;- u w = - V l
}-- -
L-
- j. .
- 6-
...l thj as ch' 'l quench; that is, most of the transient has occurred during this period before channel quench. 'Ihe diffemnees in correlation have given rise to different interpretation of the mechanism operating during this period. Inasnuch as the heat
~'
transfer inproved prior to and just'at channel quench, E believes that the phenomenon is due to inproved emissivity of Nh-the wetted channel. INC, on the other hand, attributes the
.. ,4,;
s t. . u improvement to improved convection due to splashing and slaking
- e. ~
of the water from the quench fmnt. Without resolving the Q.. differences in the two calculational nodels, it would be diffi-d cult to ccznent on the interpretation of meca.anics. y i3 If just the inner can face is cons'.dertd, E estLnates
*J heat transfer coefficients of about 20 for the channel pricr tc hi5'M can quench. INC actually extracts coefficients that are alnest always less than 10. The entire E cormlation which allows an %ir; additicrial amount of radiatic.n to be transferred to a low tempera-I The tum sink prior to and after quench is non-conservative.
j , higher channel coefficients which result in lower channel
'.~.i : temperatures provide a better radiation sink for the hot rods.
3 (. i Most subsequent calculations including some where red coeffi- :, I' * !..
.g,.r < .
1 cients wem zero show that radiation ultinately to the channel j i even before channel quench is what arrests the temperature transient. Lerefore, any mechanism which asults in lower v.
- et Amme. A er er emuneum a m ,,, gy
- % rihrsam8 ~ % F A 7 lied [
~
~m' W _ = = _:.
\* :*. .
~'-
\.l
. ,. l chanr.el terrperatures prior to quench should be evaluated nest conservatively.
II. HEATUP PEDICTIONS
~ E uses the code CHAST for bundle heatup calculations. The code can be used for predicting a IOCA or results such as FIECHT.
/t!';
'Ihe 49 rod bundle is lutped into 4 groups plus the channel box
- for ease of calculation. Group 1 is the 4 corner Iods, group 2 1s the outer rows minus the corner rods, group 3 is the second
-l rows, and group 4 is the center nine. Grey body factors are
. *g appropriately Ocmbin2d so that each group and the channel has one interchange factor with each other group, the peaking factors are h averaged, and each group has a corresponding heat transfer corre-l l
1ation extracted from the BWR ILECh7 tests (Focer's correlation). INC's revised ! OXY code does the heatup calculation. 'Ihe 49 rods are treated individually in contrast to the simplified E approach, i.e., indivictual peaking factors are used and the entire 9 i radiation matrix i's solved for each time step. 'Ihis allows varia-tion of emissivity with tinie for each body. E uses a single The INC emissivity for the entire matrix for the entire time. method then is more flexible and rigorous. 'Ihus far, INC has treated GE's groups 1 and 2 as a single Poup with respect to heat 5 transfer coefficient correlations. f
^*
f%iBiB1Tf'11 *~ ~ f &t Q v r- r a w as n ama % 767 Ed wr a v - - E u_--______ _m..
-~7 -
--m: w_ _7=-f -a"r w aww w_ 1
-8, 7:..
,ri A. FLECHT Predictions e
Using their respective codes, both organizations have done l a good job of predicting temperatures for the stainless steel
' 16 -
FLECHT tests from which the data were extracted. Agreement in I. ,. this "mfitting" process is necessary but not sufficient or
^i
=
}. surprising. In trying to predict the Zr-2 testse both do
,k(Aj
'~-
surprisingly well in view of the extensive expenne., al diffi-W culties discussed previously. INC's peak temperature predic-
,}
tions are generally closer to the experinent than N's, usually c slightly higher than the exper$3nent but lower than JE's prediction.
'Ihe time of turnaround is also usually better predicted by INC.
GWs turnaround tires are almost always too early by significant g
-: factors. 'Ihe shape of the IliC temperature curves are generally
~
; flatter around the peak as are the Zr-2 data, but usually this
.' - I is not the case with the E predictions which tend to be mon
. ;., - ] .
sharply peaked near,the maximum temperatures. B. Nine Mile Point Predictions In predicting cladding temperatures for NMP, E used the OIASf code by first applying their expa::tenta11y based dryout correlations to the blowdown, then an " adiabatic bundle" heatup
- ' from end of dryout until the core spray Itaches rated flow.
I
" Adiabatic bundle" inplies radiant energy interchange .mong the rod groups and the channel walls but with a convective coeffi-cient of zero. 'Ihe E Roger's correlation was then used to l
__._._u_______._.___ 9
- b.a bb Wws I
,f e .e s; <
1
. l model the period of spray cooling which includes the channel !
e wetting time correlation. The results are presented in the NMP addendum 3 cited earlier which indicates the peak clad
. v.
temperature of 2180'F for the largest double. ended break and 2225'F for the worst intermediate break of 0.14 fg2,
, ,;, , (
INC used the M0XY code to parametrically study spray cooling effectiveness in NMP. The temperatures at time of
}
' spray initiarion as calculated by GE were used as initial
. 1
,N temperatures in these calculations. The results are shown
) in Table 3 The result most nearly equivalent to the GE predictions are for a 60 sec spray initiation time,180 sec l
quench time, and a channel h of 20. GE predicts a peak of 2178'F with a quench time of 172 see af ter epray and a time of peak at 154 seconds after spray. The peak temperature by
) #m M0XY is 2220*F at 520 seconds af ter spray. The longer times predicted by,INC are consistent with their Zr-2K predictions and.the Zr-2K data. .The longer times and flatter profiles
! i are consistent with lower heat transfer coefficients which are not strongly temperature dependent. Since INC's extrac-l tion procedure yielded channel h's generally of 10 or less, I
whereas CE asserts that, these values should be in the order of 20, identical calculations were made with hannel h's of 10 and 20. In the Zr-2 test, 5 of 6 T/C's on the hot can faces ranged from approximately 1200*F to 1400*F prior to { gyp;Wr=rm"--*w
- ,n . - y ,
i i 1
7 h[ u C Y f,
] ,14
.' quench, 3 of those 5 quenched at between 3.7 and 4 minutes, the other 2 quenched at about 2.5 minutes, one of those was just barely at 1200'F at quench. We sixth T/C which was about 1000'F quenched at about 3.5 minutes. GE calculates a channel temperature of 1467' at spray initiation for Nine
..}.
Mile Point reactor which f alls to about 1350' at quench under the influence of the constant channel coefficient of 20. When INC uses a constant coefficient of 10 during this period,
.1
' .}
the can temperature does not fall. Under these circumstances I then, a quench time of 4 minutes appears quite reasonable. i The H0XY prediction for this case of a 4-sinute quench and a l can coefficient of 10 yields a peak of 2496*F. Re-calculation ' l of this worst case using the calculated GE clad temperatures { at 30 seconds rather than 60 seconds to simulate earlier spray b initiation yields a peak value of 2340*F. his INC value is artificially high for a 30-second spray since the channel 1 I wetting time was not reduced from 4 minutes to a more appro-priate value of about 2 minutes for a 30-second spray. There-fore, it would appear that a 30-second spray initiation will successfully arrest the clad temperature rise with reasonable choices of channel parameters. In f act, the INC prediction g 4 using a 3-minute channel wetting time is only 2268'F at i 49 sec. I i j _ _ _ _ _y n ,
- hs. j7--7_ -r~".%a w 3 ; ik *" "- * '~ * . _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
_hmsa. _ = --d.!-- 11 - .M . e. M III. NINE MILE POINT MODEL EVALUATION
'
- As was the case with the Er-2 predictions, the INC temperatures e ,- *
,3 for NMP have a flatter profile near the peak than the GE predictions.
y .1 Also, as in Er-2, the predicted times are longer in the INC-MOXY
- predictions. Although a peak may be predicted at about 500 seconds M % in IOXY, it typically may have been within 100'F of that peak at 1.';T.i 250 seconds or about the time the GE peak is predicted. Experimental vork at Oak Ridge
- and INC** suggests that extended periods of time, like 8 to 10 minutes, at modest temperatures, if ke 2200'F, may be
-i[Si' just as likely to cause brittle frseture as s short time at a more 4
' ' elevas d temperature such as 2700*F.***
If the total amount of heat transferred at the surf ace of a rod is about the same in the INC and GE extraction methods and both r l have about the same amount for " pure" coniection component, then s the GE correlation which has a larger body-to fluid radiation 4.%. , component is non-conserv'ative. As temperatures are elevated, the
!.[-.'
strong temperature dependence (roughly 3T ) for the Roger's corre- , lation is removing more heat from the bundle. Although the body-r.. .
.[ to-body component would be larger for the INC extraction method, ;
'$ > both prediction methods used the same emissivity for zircaloy. i *Rittenhouse, P. L., Progress in Zircaloy Failure Modes Research, ORNL-TM-3188 December 1970. I
**Herzel & Meservey, Trans. An. Nucl. Soc. _12 (1), 355-356,1969.
***Using Baker-Just Kinetics, 8 minutes 9 2200'F or 40 seconds at 2700'F locally reacts about 10% of BWR cladding.
N b O --E~ OED
- GWilp3-.h_" " A*U^ --
- @ 4. YM 6 I
I j
#h:2 FLR. " -- .a .
--nJ,w.
' __ ~ J' p . u.m*:
m
. .!L .
12 - r: Therefore, the body-to-body component is not as such larger as the
. INC predictions, if at all, than it was in the extraction procedure.
At any rate, a large body-to-body component does not remuve heat from the bundle but merely redistributes it within the bundle. A
,' strong temperature dependence for the body-to-fluid radiation ~9 - 4t!I ,.
component seems to be a non-conservative self-limiting feature. e p. d 6 '
- b. .
~ i:g.,
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