ML19220B312
| ML19220B312 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 01/16/1978 |
| From: | Ross D Office of Nuclear Reactor Regulation |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| References | |
| TASK-TF, TASK-TMR NUDOCS 7904250625 | |
| Download: ML19220B312 (6) | |
Text
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Distribution: @
.h nog, riie Etd 161979 RSB File Watt cnron Occket No. 50-320 h3 0RANDLH FOR:
D. B. Vassallo, Assistant Director for LWRs, CF!i 9/H:
D. F. Ross, Jr., Assistant Director for Rector Safety, CSS SL'BJECT:
S E t LINE BREAX Plant Name:
Three Mile Island, Unit No. 2 Docket Number; 50-320 Licensing Stage:
OL Responsible Branch LWR-4 and Project Manager:
H. Silver Systems Safety Branches Involved: Reactor Systeras Branch Analfsis Branch Core Perfomance Branch Description of Review:
Additional Questiens Review Status:
Awaiting Information The Reactor Systems Branch coordinated the review of a steam line break analysis submitted by the applicant November 18, 1977. The preliminary staff review indicated that the limits of ICCFR100 would be exceeded.
The applicant and their NSS3 supplier met with the staff December 6 and 9, 1977 to discuss the analysis. Infomation rqquests as a result of that meeting are presented in the enclosure. Outstanding issues other than the steam line break analyses were discussed during the meeting. Questions relating to these topics have also been included in the enclosure.
D. F. Ross, Jr., Assistant Director for Reactor Safety Division of Systems Safety
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O 212.0 REACTOR SYSTEMS BPX CH 212.1 Steam line break analyses with reactor coolant pumo coastdown and retention of reactor coolant pump cperation are provided. The description suggests that offsite pcwer was retained in both cases.
Discuss if the analysis of the cumo coastdown case is equivalent to loss of offsite power case.
If not, discuss wnat was the bases for assuming operation of any other ccmponents or systems dependent on offsite power during the analyses for Cycle 1.
Justify the use of these components and show how they were considered in the analyses.
212.2 Turbine trip was assumed not to occur for the Cycle 1 analyses.
Show how ficw to the turbine, feedwater turbine, and the break were considered. Provide the total mass flow -ate frcm the secondary system as a function of time and the contributions of each exit path. Discuss the primary to secondary heat removal that was effected by consideration of the total mass flow rate frcm the secondary system.
212.3 Figure 6, page II-17; the plot shows the MONBR recovers rapidly at 6 seconds. The staff assumes that these minimum CNBRs are for rods not exceeding MDNBR previously and bears no relationship to the cladding temperature presented.
Is this the procer interpretation?
212.4 For the 15". of the rods experiencing CNB, discuss the fluid, themal, 1
and heat transfer conditicns causing a turnaround in peak cladding temcerature and subsequent cooling to near saturation temperatures.
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212.5 The applicant has presented short-term (up to 15 seconds) analyses showing clad temceratures where the hot red has gone into film boiling. These results are derived frcm the initial axial power distribution (pre-accident) and the rapid deoressurization. The applicant has not shown that acceptable temperatures would be main-tained on these rods at icnger times (20 to 60 seconds) when the power distribution caused by 4.he stuck centrol red becomes significar*.
It is recognized that the axial power distribution caused by the stuck red combined with voiding in the core will be different than the pre-accident distribution and that the axial location of initial CNB may not correspond to the peak pcwer elevation produced by the stuck control rod.
The acclicant should provide clad temcerature for the hot red at times up to minimum subcriticality (M0 to 60 seconds).
These calculations should be based on the most limiting combination of initial power distribution (preaccident) and cost-accident power distribution (with stuck centrol rod). He should justify why the power distributions chosen result in the highest post-accident cowers (at 20 to 60 secs) at the location exteriencing film boiling because of short-term effects (at accut 5 secs),
212.6 For attachment II, discuss hcw clad swelling is considered in the analysis cerfor ed with the RACAR code.
Include in the discussion how changes in cladcing surface area affected hot channel flow and heat transfer. Drovide basis for determination that a coolable core gecmetry is retained.
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. 212.7 Isolation of Makeuo Tank Following ECCS Actuation A potential problem was identified during reactor operator tests for Three Mile Island Unit No. 2.
The operator is repuired by precedure to close a valve isolating tha makeup tank following automatic actuation of the ECCS. The first concern is that by closing the valve (MU-V-12) too soon, before supply is assured frcm the BWST, the high pressure injection pumo would fail due to cavitation. The second concern deals with operator failure to isolata the makeup tank.
In the event of a LOCA, water would continue to be drawn from both the BWST and the makeup tank. This could lead to the makeup tank hydrogen cover gas being drawn through the pt,np and into the primary system, leading to either pump failure or unacceptable hydrogen concentration in containment.
The solution appears to be achicvable by either autcmating the isolation of the makeup tank or by limiting conditions in the makeup tank.
In response to this question the applicant must describe his modification and demonstrate that the function of the HPI will net be comprcmised.
Vilve alignment associated with varicus HPI pump combinations is
'ot clear. Provide diagrams shnwing valve positions for the "arious pump combinations used for makeup and safety injection.
Identify the vital bus providing power to ecch component for each combination of pumps.
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- 232.0 CCRE PERF0r2ANCE 3 RANCH 232.1 Given that a revised fuel failure T.cdel will not be accepted (in a suitable time fra.r') and that the extent of CNB (fuel failure) is not acceptacle, aat changes can be made in the model or input (that can have reasonable ex::ectation of racid approval) to alter the results? For examole, what is the sensitivity of the (DNB) results to the moderator coefficient? If this parameter is significant, what operating limits can be employed to lessen the consequences of the accident? Auxiliary feedwater is assumed to start at 2 seconds. Would a more realistic starting time retuce the number of rods experiencing CNB?
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222.0 ANALYSIS BRANCH 222.1 For hign pressure safety injection of cold water and ficw of borated water from the core reficcd tank, provide details about the reactivity feedback, flow path in a the core, and time these fluids reach the center of the core.
222.2 Provide the results of TRAP-2 code which were benchmarked against RELAP-3B comcuter code.
222.3 Provide the results of CRAFT code which were benchmarked against experimental results.
222.4 Cescribe in more detail t&
coolant flow reduction in the hot channel folicwing a stec.n line bresk accident.
(Comoine :esponse wi th that for 212.6. )
222.5 Discuss how time-dependen; radial and axial core pea.ing factors were generated in evaluating the thermal-hydraulic effects following a steam line break accident.
222.6 Describe in detail how time-dependent pressure drop in the primary system was calculated folicwing a steam line break accident.
222.7 Show the change in core coolant density following a steam line break which accounted for the positive reactivity insertion (0 to 16 seconds) for the BOL conditions, 222.3 Cescribe in detail hcw DNBR was calculated and the assumations used.
Include the inout parameters and correlations used.
(Ccmaine response with that for 212.3.)
222.9 Cescribe hcw the percentage of core which experienced CNSR was calculated.
Identify a reference supporting figure 9 (Page II-20).
222.10 At what reduced initial pcwer level would CNS be avoiced folicwing a steam line break?
222.11 Cescribe in detail hoe ficw rate, pressure drop, and ceaking factors in the hot channel were determined. Was tha same method acplied in other channeln?
5 231.0 CORE PERFORMANCE BRANCH 231.1 Your analysis of the main steam line break (MSLB) a;cident indicates that uo to 15" of the fuel rods could experience ONB (using " worst-case" assumptions). Nevertheless, you contend that no fuel damage will occur. Please present evidence te surport this contention; in particular, provide experimental ver1 T: cation for the assertion that rods can depart frem nucleate boiling for the time intervals and peak cladding temperatures calculated for the MSLB accident, without experiencing cladding breach.
Discuss the magnitude of cladding embrittlement, softening, localized burnout, etc., expected as a result of DNB during MSLB.
231.2 Since MSLB is, for a time, a power-increasing event, there ray be a posential for fuel rod failure via a pellet / cladding interaction PCI mechansim. This has not been addressed in your analysis so fa r.
Therefore, please address the potential for PCI failure.
231.3 Provide further discussion of the methods used (1) to convert cladding failure stress to aP across the cladding as a function of temperature and (2) to calculate the fuel pin internal pressure.
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