ML20070S397
| ML20070S397 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek, Callaway, 05000000 |
| Issue date: | 02/02/1983 |
| From: | Petrick N STANDARDIZED NUCLEAR UNIT POWER PLANT SYSTEM |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| 83-0005, 83-5, NUDOCS 8302040374 | |
| Download: ML20070S397 (11) | |
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SNUPPS Standardized Nuclear Unit Power Plant System 5 Choke Cherry Road Nicholas A. Petrick Rockville, Maryland 20850 Executive Director (301) 869 8010 February 2,1983 SLNRC 83- 0005 FILE:
0278 SUBJ: Qualification-of Control Systems Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Docket Nos. STN 50-482 and STN 50-483
Dear Mr. Denton:
During the Instrumentation and Control Systens Branch review of the SNUPPS design, the commitment was made to submit to the NRC a plant-specific analysis of the effects on plant control systems of a steamline break in the vicinity of the main turbine impulse pressure trans-mitters.
The transmitters provide a signal input to the reactor rod control system.
Enclosed is an analysis of the postulated event which concludes that the consequences of the event are bounded by previous accident analyses described in the SNUPPS Final Safety Analysis Report (FSAR),
Submittal of the enclosed report also satisfies the requirements for resolution of Safety Evaluaticn Report Confirmatory Issues 13 and B.14 for the Callaway and Wolf Creek plants, respectively.
A summary of the analysis results will be used to update Question 420.3 in the SNUPPS FSAR in a future revision.
Very truly yours, l
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Nicholas A. Petrick MHF/nld/1a24
Enclosure:
Analysis of a Steam Line Break with Coincident Control Rod Withdrawal cc:
G. L. Koester KGE J. H. Neisler USNRC/ CAL C)O )
D. T. Mc Phee KCPL T. E. Vandel USNRC/WC D. F. Schnell UE G. Edison USHRC 8302040374 830202 PDR ADOCK 05000482 E
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ANALYSIS OF A STEAM LINE BREAK WITH COINCIDENT CONTROL R00 WITHDRAWAL O
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t DESCRIPTION OF EVENT The failure of main' steam piping in the Turbine Building could produce adverse environmental conditions in the vicinity of the turbine impulse pressure transmitters. The transmitters provide control signals to the reactor rod control system.
It has been postulated that the combined effect of the steam line break and a consequential failure of the turbine impulse pressure transmitters in a manner to adversely affect the rod control system could result in plant response more severe than those of previously analyzed transients and accidents. Of specific concern is the postulated control rod system failure which could result in increased reactivity addition to the core, i.e. a continuous control rod withdrawal.
The following is an analysis of this postulated event.
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AVAILABLE PROTECTION The following functions provide protection during this rod withdrawal type of transient:
Power range neutron flux instramentation actuates a reactor trip if two out of four channels exceed an overpower setpoir.t.
A reactor trip is actuated if any two out of four AT channels exceed the overpower AT setpoint.
A high pressurizer pressure reactor trip is actuated from any two out of four pressure channels which are set at a fixed point. This set pressure is less than the set pressure for the pressurizer safety valves.
I A high pressure water level reactor trip is actuated from any two out of three level channels when the reactor power is above approxim'ately 10 percent (Permissive 7).
A reactor trip is actuated subsequent to SIS actuation. SI may be actuated as a result of the steam line break.
d RCCA Withdrawal Blocks High neutron flux (one out of four power range) 0verpower AT.(two out of four)
'Overtemperature ai (two out of four)
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The following functions provide protection for the steam line break:
Safety Inje'ction Two out of four low pressurizer pressure signals
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Two out of three low steamline pressure signals in any one loop Feedwater Isolation Sustained high feedwatar flow would cause additional cooldown. Therefore, in addition to the normal. control action, w'h'ich will close the main-feed-water
. valves following a reactor trip, an SI signal will rapidly close all feed-water control valves and backup feedwater isolation valves, trip the main feedwater pumps, and close the feedwater pump discharge valves.
Steam Line Isolation Safety injection system actuation derived from two out of three low steamline pressure signal in any one loop (above Permissive-11)
Two out of three high negative steam pressure rate in any one loop (below Permissive-11)
All of the above functions may be actuated by a SLB/RCCA withdrawal transient.
ANALYSIS OF EFFECTS AND C'ONSEQUENCES Method of Analysis l
This transient is analyzed by the LOFTRAN code. This code simulates the' neutron kinetics, RCS, pressurizer, pressurizer relief and safety valves, The pressurizer spray, steam generator, and steam generator safety valves.
code computes pertinent plant variables, including temperatures, pressures,
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and. power level.
Burnett, T. W. T., et. al., "LOFTRAN Code Description," WCAP-7907, June 1.
1972. Also supplementary information in letter from T. M. Anderson, NS-TMA-1802, May 26,1978 and NS-TMA-1824, June 16, 1978.
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,s A detai,le,d themal and hydraulic digital-computer code, THINC, has been used to detemine if DNB occurs for the core.coiiditions computed by the LOFTRAN code.
The following assumptions were made for this transient.
Initial conditions of maximum core power and reactor coolant average tem-a.
perature and minimum reactor coolant pressure, resulting in the minimum initial margin to DNB are used.
The most End-of-life shutdown margin and equilibrium xenon conditions.
b.
reactive RCCA stuck in its fully withdrawn position is assumed for conditions following reactor trip.
A negative moderator coefficient corresponding to the end-of-life unrodded c.
This maximize the reactivity insertion caused by the core is used.
cooldown during the steam line break.
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Minimum capability for injection of boron (2,000 ppm) solution corres-d.
ponding to the most restrictive single failure in the safety injection 1)
The emergency core cooling system consists of three systems:
system.
the passive accumulators, 2) the residual heat removal system, and 3) the, Only the safety injection system is modeled for safety injection system.
this analysis.
The reactor trip on high neutron flux is assumed to be actuated at a e.
The AT trips conservative value of 118 percent of nominal full power.
include all adverse instrumentation and setpoint errors; the delays for
, trip actuation are asstaned to be the maximum values.
The RCCA trip insertion characteristic is based on the assumption that the f.
highest worth assembly is stuck in its fully withdrawn position.
2 This is the The break size assumed for this transient is 1.2 feet.
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. largest for which a low steamline pressure signal will not occur prior to Prior to the eventual steamline isolation on the reactor trip on OPAT.
low steamline pressure, this break is fed by all four steam generators.
Fol1owing steamline isolation the break will be fed from one steam f
generator causing an assymetric transient.
h.' In computing the steam flow during a, steamline break, the Moody Curve for
- fL/D=0 is used.
Results The calculated sequence of events for the SLB/RCCA withdrawal transient is shwn on Table 1.
Figures 1 and 2 show the RCS transient and core heat flux following the stece-lin re with coincident RCCA withdrawal.
The steamline break affects the turbine impulse transmitters and causes the control rods to withdraw at the initiation of the transient. This causes an increase in reactor power and core heat flux to,the point at which the over-power delta-T trip setpoint is reached. This indrease in core power geMrates a reactor trip which tenninates the most adverse part of the transient. The steamline. break causes an increased heat removal and consequent de::rcase in primary pressure simultaneous with the increase in reactor power. Secondary pressure also decreases until the low steamline pressure setpoint is reached l
initiating steamline and feedwater isolation.
- Because of the lower RCS pressure coincident with the increase in reactor l.
power, the consequences at the point of peak heat flux may be more adverse i
than the Rod Withdrawal at Power transient analyzed in the FSAR. The most limiting part of this transient pertinent to this study is immediately prior to reactor trip;,for this reason the analysis is tenninated at 50 seconds.
The modelling of Engineered Safeguards Features (SI, SLI, FWI~)'is not needed l
since they will not be generated prior to reactor trip. The ret 0rn to power following reactor trip and steamline isolation is bounded by the* transient for the larger break presented in the FSAR. The FSAR analysis assumed a larger break size and initial conditions corresponding to no-load tenperatures (i.e.
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less stored energy in the RCS and reactor fuel).
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Margin to Critical Heat Flux A DNB analysis was perfomed for this transient. The DNBR was found to be great.t than the limit value at all times.
CONCLUSIONS The analysis demonstrates that DNBR does not decrease below the limit value
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and no fuel or clad damage is predicted. Additionally no system overpressur-ization is expected, thus all applicable acceptance criteria are met. There-fore there is adequate protection on the SNUPPS plants to ensure plant safety for this transient.
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FIGURE 1 NUCLEAR POWER, CORE HEAT FLUX, AND RCS PRESSURE FOR THE SNUPPS SLB/RCCA WITHDRAWAL TRANSIENT
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TIME SEQUENCE OF EVENTS OF THE STEAM LINE BREAK WITH 'A COINCIDENT CONTROL ROD WITHDRAWAL" b-Time (sec)
Event 0.
Steam, line ruptures -
Overpower delta-T reactor trip setpoint reached 5.0 8.0 Rods begin to fall 21.4 Lew steam line pressure setpoint reached Steam line isolation occurs 28.4 28.4 Feedwater isolation. occurs a
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FIGURE 2 avg. REACTOR VESSEL INLET TEMPERATURE, AND SG CORE T
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