ML20024G650
| ML20024G650 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 07/31/1978 |
| From: | Brugge R, Henrikson P GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20024G412 | List: |
| References | |
| NEDO-24133-1, NEDO-24133-1-S01, NEDO-24133-1-S1, NUDOCS 9102140388 | |
| Download: ML20024G650 (10) | |
Text
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[uppletent 3 Class I July 197f SLTPLEMENT 1 FOR }f]NTICELLO RELOAD 6 SIMMER MARGI EVALUATION l
l l
P. H. Henriksen Licensing Enginect f>
f f/ f 'Y Approved: A . U I 'F- - t f~f <_
R.O. Brugge, Ma ter Operating Licenses II NU*LE AR ENE RGY PROJE CTS DIVISION
ADOCK 05000263 GENERAL ELECTRIC PDR
. 1 . i DISCLAIMER OF RESPONSIBILITY Th:s occument ses prepa'eu og or for the General Electnc Company, Neither tnt-General Electric Company nor any of the contnbutors to this document:
A Makes any serranty or representation, express or emplied. with respect to the accuracy. completeness cr usefulness of the rnformstron contarnedtn thos docu-Inent, of t*:st the ust of any snfo? mat +0n Cisclosed in this document may nct Infringt pnkaftty on net rigM:- c c
& Assu",es s'1v res Ons bahty for liabr!>ty or damage of any kind which may resu!!
ficm the use of any tr'torma Qn onsC!osto t1 this doc.ument
NEDO-24133-1 StTTLDINT 1 FOR M0hTICELLO RELOAD 6 SIM'IER MARG 7N EVALUATION l
1
- 1. INTRODUCTION AND StJMMARY ,
One event that has a significant inpact on boiling water reactor (B'a'R) availability is the spurious opening or failure to reclose of the dual function safety / relief valves. As described in Reference 1, the event from a safety standpoint has a relatively minor effect on the reactor core and reactor coolant pressure boundary. i l
The event does result in a significant maintenance outage since the reactor must l be shutdown, depressurized, and the valve repaired or replaced before the plant !
can be restarted and continue with power operation.
The cause of the majority of these spurious openings or failures to reclose of safety / relief valves is excessive leakage around the setpoint pilot vc1ve. Other causes of valva failures ha i been identified and corrective action has been taken. Operatint data demonstrctc that an increase in valve simmer margin (the differential pressure between the valve setpoint and normal system operating pressure at the valve) will reduce the probability of valve failure due to pilot leakage. A study was perforced for Reload 6 of Monticello Nuclear Generating Plant, to determine if the simmer margin of the safety / relief valve could be increased without imposing additional restrictions on plant operation.
This supplement provides the results of these evaluations. As demonstrated in Section 2, the operating limits derived in NEDO-24133 are still valid for a 28 psi increase in safety / relief valve setpoint. Therefore, additional saf ety/
relief valve reliability can be obtained for the next operating cycle without the imposition of any new limits on the plant.
- 2. SAFETY ANALYSIS 2.1 Introduerion The safety analysis for Reload 6 1s provided in NEDO-24133. The raising of the safety / relief valve setpoints only af fects those events which result in valve 1
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NEDO-24133-1 operstion to limit system pressure. The limiting events which require reanalysis are the test severe pressurization transient (turbine trip with failure of the bypa s valve), vessel overpressure protection analysis (closure of all main steamline isolation valve - flux scram) and loss-of-coolant accident (small .
break). In addition, the capability of the reactor core isolation cooling (RCIC) and high pressure coolant injection (KPCI) systems were re-evaluated for the higher safety /rclief valve setroints. The results of the analysis which demonstrate the acceptability of the increased simmer margin are given below.
All analyses were performed using the same input parameters as used in NEDO-24133 with the exception of safety / relief valve setpoint and capacity. The nominal safety / relief valve setpoint assumed was 1108 psig +1% using seven safety / relief valves. The capacity of the safety / relief valves i t their setpoint was 83.2%
of rated etcam flow. The increase in saf ety/ relief valve capacity is due to the increase in tass flow rato as a result of the higher pressure at the valve setpoint, an? sevcn safety / relief valver are used instead of sir.
2.2 Turbine trip With Failure of the bypass Valves This transient produces the most severe reactor isolation. The primary charac-teristic of this transient 1s a pressure increase due to the obstruction of steam flow by the turbine stop valves. The pressure increase causes a signiff-cant void reduction, which yields a pronounced positive void re&ctivity effect.
The net reactivity is sharply positive and causes a rapid increase in neutron flux until the net reactivity is forced negative by a scram initiated from position switches on the turbine stop valves and by a void increase after the safety / relief valves have sutomatically openad on high pressure. The results of thest analyser are given in Tatic 1 and shown in Figure 1.
The changc in critical power ratio caused by the change in setpoint is insignif-icant (0.002 ACPR). However, this change was enough to affect the roundoff of the third significant figure, so thct the LCPR for turbine trip without bypass with increased sicmer margin is 0.26. Theref o re , the MCPR Operating Limit with increased simmer cargin is 1.33 for both 8x8 and 8x8R fuel.
2
KEDo-24133-1 2.3 '.'es sel Overpressure Protection Analvst e The pressure r,elief system must prevent excessive overpressurization of the primary system process barrier and the pressure vessel to preclude an uncon-trelled release of fissien products.
The Monticello pressure relief systec includes eight dual function safety / relief valves located en the main steamlines within the drywell between the reactor ,
vessel and the first isolation valve. These valves provide the capacity to l l
lindt nuclear system overpressurization (analysis assumes 7 S/RV's).
l The ASMI Boiler and Pressure Vessel Code requires that each vessel designed to meet Section III be protected from the consequences of pressure in excess of the vessel design pressurc:
(a) A peak allowable pressurt cf 110'. cf the vess(1 design pressure is allowed (1375 psit for a vessel with a design pressure of 1250 psig).
(b) The lowest qualified safety / relief valve setpoint must be at or belov vessel design pressure.
(c) The highest safety / relief valve setpoint must not be greater then 105%
of vessel design pressure (1313 psig for a 1250 psig vessel).
Monticello's safetyIrelief valves will be set to self-actuate at a nominal set-point of 1108 psig, thereby satisfying C and (c) above. Requirement (a) is evaluated by considering the most severe isolation event with indirect scrat.
The code does not require failure ef reactor protective systems; however, General Electric provides a conservative analysis for licensing purposes which take credit only for reactor protective signals which are indirectly derived.
The event which satisfies this specification is the closure of all main steamline isolation valves with indirect (flux) scrac. The initial conditions assumed are those spccified in Section 6 of NEDD-24133. Figure 2 graphically illustrates the event. An abrupt pressure and power rise occurs as soon as the reactor is isolated. Neutron flux reaches scrac leve] at about 1.7 seconds, initiating 3
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l reactor shutdewn. The safety / relief valves open to limit the pressure rise to 1248 psig at the bottom of the vessel. This response provides a 127 psi margin to the vessel' code limit of 1375 psig. Thus, requirement (a) is satisfied and .,
adequate overpressure protection is provided by the pressure relief syster.
I 2.4 Loss-of-Coolant Accident Analysin Analysis of the design basis loss-of-coolant accident demonstrates that the !
pressure decays during the event, and the change in safety / relief valve setpoints !
vill have ne (ifcet or. the results. However, f or smaII breaks, the reactor will remain pressurized until the initiation of the automatic depressurization system (assuming the single failure of the HPCI). The change in safety / relief valve setpoint will result in a slight increase in inventory loss of the treak durint this perdc'.
ECCS analysis predicts a I'Ci of approxi=ately 1760'F which is 40*F higher that.
that for the case of the old SRV setpoint, for the most limiting small recircu-2 lation line break of 0.07 ft . This small increase in PCT is due primarily to the fact that the higher SRV setpo1nt results in higher vessel pressure which increases inventory loss and delays ECC systems initiation slightly.
I 2.5 RfCI and RCIC Capability One of the design requirements for the HPCI and RCIC systems is that they be capable of providing design flow at the lowest safety / relief valve setpoint.
These systers still meet the design requirement with the increase in lowest safety / relief valve setpoint to 1106 psig, the nominal setpoint.
4 4
l NEDO-24133-1 i
Table 1 EVENT DATA SD24ARY (EOC7)
Peak Peak ..
Core Peak Neutron Peak Surface Steamline Vessel Power TIow Flux Heat Flux Pressure Pressure Event (%) (%) (% of Ref) (% of Ref) (psig) (psig) '
Turbine Trip v/o 100 100 312 115 1168 1207 Bypass - Trip Scram MSIV. Closure, 100 100 602 127 1199 1248 Y1ux Scram 5
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'1 REGULATORY INFORMATION DIGTRIDUTION SYSTEM (RIDS)
, n'<.Tv 1r: m mt , r ,l,1 LR I ra S0.263 i
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ORG: MAYER L 0 DOCD ATE: 08/16/7Fl NRC N OTAlES PWR DATE RCVD: 08/18. l DOC 7YPE. LElTLR NOT ARI ZED: YEO
SUBJECT:
COPIES RECEIVED LTR 3 ENCL 40 l FORWARDINO LII. f)O DPR-22 APPL FOR AMEND: APPENDIX A TECH SPEC PROPOSED CI CONCERNING hEVISION 10 TiiE PERMISSIBLE SETPOINT OF THE EIGHT SAFETY / RELIE{
VALVES INSTALLED AT SUBJECT FACILITY TO 1100 PSIO NOTARIZED 00/16/78.
W/ATT NEDO-24133 M M i
PLANT NAME: MONTICELLO REVIEWER INITIAL: Xuffj DISTRIBUTOR INITIAL:#C
- u*H M*+****+ 1 DISTRIBUTION OF THIS MATERIAL IS AC FOLLOWS **************+++<!
GENERAL DISTRIBUTION FOR AFTER ISSUANCE OF OPERATING LICENCE.
(DISTRIBUTION CODE AOO1) 1 FOR ACTION: BR CHIEF ORBil? EC**W/7 ENCL J NT ERt: c ; f ILEe U OIf?: NRL PDR**W/ ENCL i I b E**W/2 ENCL DELD**LTR ONLY HANAUER**W/EtKL CORE PERFORMANCE BR**W/ ENCL l AD FOR SYS & PROJ&&W/ ENCL ENGINEERING BR**W/ ENCL '
REACTOR SAFETY PR* pW/ ENCL EEBe*W/ ENCL PLANT OYSTEMS BR**W/ ENCL
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J. MCGOUGH*nW/ENLL EFFLUENT TREAT SYS**W/ ENCL j EXlERNAL LPDR'S MINNEAPOLIS, MN**W/ ENCL TERA **W/ ENCL l
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