ML20071B034
| ML20071B034 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 02/23/1983 |
| From: | Longenecker J ENERGY, DEPT. OF, CLINCH RIVER BREEDER REACTOR PLANT |
| To: | Grace J Office of Nuclear Reactor Regulation |
| References | |
| HQ:S:83:217, NUDOCS 8302250292 | |
| Download: ML20071B034 (8) | |
Text
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O Department of Energy Washington, D.C. 20545 Docket No. 50-537 HQ:S:83:217 FEB 2 31993 Dr. J. Nelson Grace, Director CRBR Program Office Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Dr. Grace:
ADDITIONAL INFORMATION ON THE SECONDARY CONTROL R0D SYSTEM - CLINCH RIVER BREEDER REACTOR PLANT Enclosed are marked-up Preliminary Safety Analysis Report (PSAR) pages providing additional information on the secondary control rod system capability. This infonnation will be placed into Chapter 4 of the PSAR in a future amendment.
Any questions regarding the information provided can be addressed to S. Frye (FTS 626-6354) or K. Peterman (FTS 626-6186) of the Project Office Oak Ridge staff.
Sincerely, GMit JodnR.Longene er Acting Director, Office of Breeder Demonstration Projects Office of Nuclear Energy Enclosure cc: Service List Dl Standard Distribution nO Licensing Distribution U
B302250292 830223 PDR ADOCK 05000 A
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7.2 REACTOR SHUTDOWN SYSTEM
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7.2.1 DesertDtion
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7.2.1.1 Reactor shutdown System Description The Reactor Shutdown System (RSS) consists of two independent and i diverse systems, the Primary and Secondary Reactor Shutdown Systems, eithgr,%
of which is capable of Reactor and Heat Transport System shutdown.
All ed asiJ M@NW'M"4 events can be terminated without exceeding the specified limits by either system even if the most reactive control rod in the system cannot be inserted.
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.M M Vf = '4.- To assure adequate independence of the shutdown systems, mecha-nical and electrical isolation of redundant components is provided.
Fun;tional or equipment diversity is included in the design of instrumentation and electronic equipment. The Primary RSS uses a local coincidence logic con-figuration while the Secondary RSS uses a general coincidence.
Sufficient s.
s redundancy is included in each system to prevent sin degradation of either the Primary or Secondary RSS. gle random failure
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As shown in the block diagram of the Reactor Shutdown System, Figure 7.2-1, the Primary RSS is comoosed of 24 subsystems and the Se::endary {
RSS is composed of 16 subsystems.
Figure 7.2-2A is a typical Primary RSS instrument channel logic diagram.
Each protective subsystem has 3 redundant -
sensors to monitor a physical parameter.
The output signal from each sensor g is amplified and converted for transmission to the trip comparator in the e
control room.
Three physically separate redundant instrument channels are used. When necessary, calculational units derive additional variables from the sensed parameters with the calculational units inserted in front of the I
comparators as needed.
The comparator in each instrument channel determines if that instrument channel signal exceeds a specified limit and outputs 3 redundant signals corresponding to either tne reset or trip state.
The f
3 outputs of each comparator are isolated and recombined with the isolated outputs of the redundant instrument channels as inputs to three redundant i
logic trains. The recombination of outputs is in a 2 out of 3 local coin-t cidence logie arrangement.
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Operating bypasses are necessary to allow RSS functions to be bypassed durin
__1m,Wng main sodium coolant pump startup and ascent to power.
Operating bypasses are accomplished in the instrument channels.
For bypasses associated with normal three loop operation, the l
bypass cannot be instated unless certain permissive conditions exist which i
assure that adequate protection will be maintained while these protective functions are bypassed.
Permissive comparators are used to detennine when bypass conditions are satisfied.
When pennissive conditions are within the 57 allowable range, the operator may manually instate the bypass.
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7.2-1 Nov. 1980
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TABLE 7.2-2 (Continued)
Primary Reactor Secondary Reactor Fault Events Shutdown Systest Staatdmst Systm Failure of Stema Dap System Steam-Feedwater Flow Steam Drtan IAvel Mismatch Sodium Water Reaction in Steam (3)
Steam-Feedwater Flow Sodium-Water Reaction Generator Mismatch III.
Extree.1v thlikelv A.
Reactivity Disturbances a
Ibsitive Ramps IS2.0/sec Startup Primary Iow Flux Startup Nuclear 25-40% Power Flux-Delayed Flux or Modified Nuclear Rate or Flux-Pressure Flux-40tal Flow 4
40-100% lbwer Flux-Pressure Flux-Total Flow Full Ibwer High Flux Flux-Total Flow SSE IfTS Ptap Frequency IfrS Pump Voltage B.
Sodium Flow Disturbances HfTS Leak Manual Trip (4)
Manual Trip IffTS Leak IHX Primary Outlet Primary to Intermediate Flow Ratio Temperature The maxinum anticipated reactivity fault results from a single failure of the control system with a maxinem insertion rate of (1) a p oximately 4.1 cents per second.
We maxinem unlikely reactivity faults result fran multiple control system failures leading to withdrawal of six rods at (2) normal speed or one rod at the maximum mechanical speed.
The PPS is required to terminate the results of these extremely unlikely events within the umbrella transient specitled as (3) emergency for the design of the major components.
No automatic PPS protection is required for the DBE Rf!S leak of 8gpa since the required response time is significantly (4) greater than 30 minutes and safety-related information systems are provided to inform the Operator of the presence of a Hrts For additimal protection (margin), a reactor sodium level trip subsystem is included in the Primary RSS to provide leek.
protectim for anS leaks beyond the design basis.
7.2-22 Amend. 76 Feb. 1983
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... 'llMity AND SEC0flDAltY SilUT00im SYSTEM DAHAGE SEVERtTf LIMITS
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gl Primary System (Inly functioning Secondary System Only Functionin't lfi tliout With WithouL Wi th Stuck Rod St 'k Rod Stuck Rod Sturt Rod 45 Norsul: Operatiosial {2) riot Applicabic ifot Applicable riot Applicable Not Applicable i
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'Dperational Operational Minor Jncident (1)
Hinor incident (1) lacident incident T.
Emergen. y: Unlikely Faults.. " Hinor incident Minor incident Ibjor Incident Major Incident 1
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(1) Failure of the primary system to scram w!ica required for an anticipateil fault is defined as an cxtremely unlikely event (faulted condiLion). Hcuever, the damage 3,cverity limit for the second.try shutdown system is conservatively specified to assure fuel piti integrity cyca for tPee concurrent anticipated fault and failure of the prisury shutdown system.
j (2) fio action required by Plant Protection System during Normal Operation.
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-15.3.3.4 Primarv naat Tranamrt System Pipe rmk 15.3.3.4.1 Taantification of canaaa and Accident naccriotion Small sodium leaks have occurred several times in sodium testing facilities and in operating reactors. As a result, HffS leaks are considered in the design and evaluation of the plant to assure that the design has adequate cagabilities from the standpoint of core thermal transients. Wis partimlar section will address the NTS pipe leak as an undercooling event while section 15.6.1.4 provides a detailed discussion of the NTS pipe leak and its consequences with regard to cell pressure and temperature transients and radiological effects.
Based on a detailed evaluation of the NTS piping structural integrity, presented in Reference 2 of Sectim 1.6 of the PSAR, a 4-inch crack was chosen -
to eri.ablish the design basis leak 690) for the functional performance of the heat transport system (see PSAR Section 3.6.1.1).
We maximum leak rate corresponding to the 4-inch crack is 8 gal / min. As indicated in Section detection capability for leaks as small as 100 gg/hr (- 5.3x10p to provide 7.5.5.1, the liquid metal-to-gas leak detection system is desi-gal / min).
15.3.3.4.2 Analysis of Effects and Consequences A 8 gal / min leak would not result in any measurable core transient. An automatic reactor trip would not be required and adequate time (significantly greater than 1/2 hour) would be available for the operator to manually shutdown the reactor. W erefore, a leak from the NTS is not a design basis event for the Plant Protection System. A normal reactor shutdown would be accomplished following indications from the leak detection system. We primary PPS includes reactor vessel sodium level and flux / pressure trip
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functions, that would provide margin capability to scram the reactor in the event of a leak significantly greater than the IBL.
Following an indication of a leak, the reactor would be shutdown and the coastdown of the pumps would re&ce the system pressure. After pump coastdown l
(<1 minute), the leak rate would be reduced to a fraction of the 8 gal / min l
leak rate used for the event because of the pressure re&ction and the system would then continue to drain until static equilibrium of the fluid in the '
system is reached, asuniming no operatur action to re&ce the amount of sodium released. Se quantity of sodium whidi could potentially leak from the system
& ring this period is dependent on the location of the leak and the action that the operator takes. Once the plant is shutdown, the leakage rate becomes so small that the operator would have several days to select a method for further re&cing the sodium leakage. Even if no further action were taken, l
the system design (gucrd vessels and elevated piping) would assure that long term core cooling would be govided.
Se 8 gal / min leak rate is orders of magnitude below the leak rate that could cause a significant core transient. Conservative analysis indicates that for 3-loop operation, a transient maximum loss rate of over 50,000 gal / min would be required for the core sodium temperature to approach the saturation value, and this would require a rupture of more than 1 square foot at the reactor 15.3 -50 Amend. 76 Feb. 1983
inlet nozzle. At other postulated primary heat transport system locations, even larger rupture areas would have to be postulated to challenge core cooling. Segrate best-estimate margin analyses have demonstrated that even leaks as large as a double ended rupture can be acconmodated without a loss of core coolable geometry. We results of this analysis were confirmed in Reference 16 of Section 1.6 of the PSAR.
15.3.3.4.3 conclusions
%e improbable occurrence of a leak, on the order.of 8 gal / min in the HfTS piping would lead to an inconsequential transient in the reactor. Activation of several leak detection systems would result in corrective action including manual plant shutdown. % e consequences would be limited to an economic penalty for plant downtime, sodium cleanup, and piping repair. Moreover, a leak several orders of magnitude greater than the 8 gal / min leak would not cause hot channel coolant temperatures to approach saturation.
l 15.3-51 Amend. 76 Feb. 1983
4 9.
W e only credit for operator action in mitigation of postulated sodium spills is shutdown of the Na overflow system makeup pumpt 30 minutes after plant scram for a postulated leak in the Primary Heat Transport System (see Section 15.6.1.4).
- 10. Analyses of liquid metal burning in inerted cells assumes burning of all oxygen in the cell in which the liquid metal is postulated to leak and burning of all the oxygen contained in cells which are environmentally connected to the cell with the liquid metal leak.
- 11. %e analysis of postulated liquid metal fires in air-filled cells does not include reaction of the liquid metal with postulated water released from concrete. %e validity of this approach is presently being verified in conjunction with the large scale sodium fires test program discussed in Section 1.5.2.8 of the PSAR.
If the test program does not a pport the present analysis approach, the appropriate effects of water release from concrete will be included in subsequent analyses.
- 12. %e Moderate Energy Fluid System (MEFS) leak is used in this section to conservatively establish the CRBRP cell structural design performance. For purposes of assessing 1the functional performance of the heat transport systems, a leak rate corresponding to a 4-inch crack ( 8 gal / min)'was selected based on a detailed evaluation of the IETS piping integrity (see Section 15.3.3.4).
Table 15.6-1 provides a sunmary of the initial conditions for each fire considered and the maximum off-site dose as a percentage of the 10CFR100 guideline limits. As the table indicates, a large margin exists between the EM ential off-site doses and 10CFR100. A discussion of the pressure /
temperature transient for each event is provided in the following sections; in no case do the fires-result in conditions beyond the design capabiity of the cell / building.
%e Project is assessing the imEacts of NaK spills in the Reactor Service Building and will provide the results of aerosol released from the Reactor Service Building when the assessments are completed. The aerosols released from the RSB as a result of NaK spill will be controlled so as not to affect safety-related equipnent.
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15.6-2a Amend. 76 Feb. 1983
TABLE 15.6-1 EDIUM SPILL EVDirS Max.
Off-Site Max. Cell Section Sodium Spill tocation*
Dose t of Gas No.
Events Gallons Temp (F)
Atmospere Bldg.
Cell 10CFR100 Press /Tesp 15.6 Sodium Spills 15.6.1 Extremely Unlikely 15.6.1.1 Primary sodium in 35,000 400 Normal n
Overflow 0.8 pig contairunet stor-Air Tank Cell 0.19 1380F**
age tank failure
& ring maintenance Design Press 10 pig 15.6.1.2 Failure of ex-vessel 15,250 600 Inerted RSB Ex-Vessel 3.8 pig sodium cooling sys-Sodium Tank 0.48 2540 ***
F tem & ring operatim Cell Design Press 12 psig 15.6.1.3 Failure of ex-con-50,000 450 Inerted SGB/IB Storage Tank 3.5 pig tairunet primary Cell 2.13 2600 ***
F sodium storage tank Desip Press 4 psig 15.6.1.4 Primary Heat 35,100 1015 Inerted n
arts Cell
<10-4 14.4 pig Transport System (HrrS Cell) 6800F***
piping leak ****
Design Press 30 psig 29,200 (Reactor Cavity) 750 Inerted n
Reactor Cavity 10.3 paig 6500F***
Desip Press 35 pig 15.6.1.5 Intermediate Heat 39,000 8000F Normal SGWIB IB 3
0.4 pig Transport System Air 6300F***
pip ng leak i
Design Press 3 pig
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- Reactor Service Building SGB/IB - Steam Generator Bldg / Intermediate Bay HrrS - Primary Heat Transport System
- In Contairunent
- In Affected Cell
- Although considered to be beyond the HrrS desip basis, the MEFS leak is included in the extremely unlikely category for cell structural evaluatims only. See Section 15.3.3.4 for systems effects fran leaks.
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15.6-3 Amend. 76 Feb. 1983 g