ML20151Y920
| ML20151Y920 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek, Callaway, 05000000 |
| Issue date: | 02/11/1986 |
| From: | Petrick N STANDARDIZED NUCLEAR UNIT POWER PLANT SYSTEM |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19273A863 | List: |
| References | |
| SLNRC-86-3, NUDOCS 8602130220 | |
| Download: ML20151Y920 (10) | |
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SNUPPS Stendenfiaed Nucteer Unit Pouver Plant System 5 Choke Cherry Road Rockville, Maryland 20060 (301) 300w4010 February 11, 1986 SLNRC 86-3 FILE: 0278 SUBJ: Steam Generator Tube Rupture Analysis - SNUPPS 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
References:
1.
Facility Operating License NPF-30 for Callaway Plant, Unit No. 1 2.
Facility Operating License NPF-42 for Wolf Creek Generating Station, Unit No. 1 3.
SLNRC 86-1, dated January 8, 1985, Steam Generator Tube Rupture Analysis - SNUPPS
Dear Mr. Denton:
In accordance with License Condition 2.c.(11) of both references 1 and 2, SNUPPS submitted a report (reference 3) which demonstrated that the steam generator single'-tube rupture (SGTR) analysis presented in the Final Safety Analysis Reports (FSARs) for the SNUPPS plants is the most severe case with respect to release of fission products and calculated radiation doses, thus fulfilling the above license conditions for both Callaway Plant and Wolf Creek Generating Station.
Several corrections to the report enclosed with reference 3 have been identified and are tabulated in Attachment 1 along with the reason for the correction.
These corrections do not alter the conclusions stated
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0602130220 860211
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DR ADOCK 05000482 1 I PDR
f. ~. b 86-3 Page 2 in reference 3; namely, that the analysis presented in the FSARs for the SNUPPS plants is the most severe case with respect to the release of fission products and calculated radiation doses.
Also enclosed are new pages (proprietary version and non-proprietary version) for insertion in the reports enclosed with reference 3.
Note also that reference 3, dated January 8,1985 should have been dated January 8, 1986.
Very truly yours,
%l N cholas A. Petrick J0C/ NAP /mys22A21-22 Attachments cc:
G. L. Koester KGE J. M. Evans KCPL D. F. Schnell UE B. Little USNRC/ CAL J. E. Cummins USNRC/WC G. C. Wright USNRC/RIII E. H. Johnson USNRC/RIV
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Changes to Report Steam Generator Single-Tube Rupture Analysis For SNUPPS Plants - Callaway and Wolf Creek 1.
Typographical errors Figure 3 Steam is typed incorrectly as " stream" in two locations.
Table 4 Case 1 exclusion area boundary thyroid dose should have been 0.28 rem instead of 0.25 rem.
Table 4-4 Case 2 low population zone outer boundary thyroid dose should have been 3.2 rem instead of 2.2 rem.
Page A-1
- Three words were omitted from the first sentence. The sentence should read, "The scoping code for steam gen-erator tube rupture (SGTR) events was developed to evaluate in a scoping manner the effects of operator response times and various assumed single failures."
Page D-3
-Thesymbolforupstreamdensity,f,wasomittedinthe equation for GB and in the definition of the symbol.
Page F-4
- Analysis is typed incorrectly as " Analyzers" in Reference 2.
2.
Changes to Figure C-3 The times used in Figure C-3 to plot the scoping code reactor coolant system pressures were incorrect and should have terminated when the ARV was closed.
Figure 3-1 Partial Composite Schematic Otagram of SNUPPS Secondary Systes AAY 5 Spring-Loaded Safety Valves (TYP) 6 ivIV Fcy
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b TABLE 4-2 RADIOLOGICAL CONSEQUENCES OF-A STEAM GENERATOR TUBE RUPTURE WITH FAILED OPEN AUXILIARY FEEDWATER CONTROL VALVE Doses (rem)
Limiting Site (Callaway) 1.
Case 1 Exclusion Area Boundary (0-2 hr)
Thyroid, rem 0.28 Whole body, rem 0.007 Low Population Zone Outer Boundary (duration)
Thyroid, rem 0.06 Whole body, rem 0.002 2.
Case 2 Exclusion Area Boundary (0-2hr)
Thyroid, rem 3.3 Whole body, rem 0.01 Low Population Zone OuterBoundary-(duration)
Thyroid, rem 0.45 Whole body, rem 0.003 Case 1 - Accident induced iodine spiking per SRP 15.6.3 Case 2 - Pre-existent iodine spike per SRP 15.6.3 4-7 Rev. 1
TABLE 4-4 RADIOLOGICAL CONSEQUENCES OF A STEAM GENERATOR TUBE RUPTURE WITH A STUCK OPEN ATMOSPHERIC RELIEF VALVE Doses (rem)
Limiting Site (Callaway) 1.
Case 1 Exclusion Area Boundary (0-2hr)
Thyroid, rem 2.7 Whole body, rem 0.0'8 Low Population Zone Outer Boundary (duration)
Thyroid, rem 0.36 Whole body, rem 0.004 2.
Case 2 Exclusion Area Boundary (0-2 hr)
Thyroid, rem 24.6 Whole body, rem 0.04 Low Population Zone Outer Boundary (duration)
Thyroid, rem
'3.2 Whole body, rem 0.007 Case 1 - Accident induced iodine spiking per SRP 15.6.3 1
Case 2 - Pre-existent iodine spike per SRP 15.6.3 i
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Appendix A - Description of Scoping Code I.
General The scoping code for steam generator tube rupture (SGTR) events was developed to evaluate in a scoping manner the effects of operator response times and various assumed single failures. These analyses focus on the secondary side behavior of the SGs and are normmalized to RETRAN results.
The methodology of the scoping analyses may be summarized as follows.
The 0-2 hour time period following occurrence of a postulated SGTR is divided into six time intervals, corresponding to specific phases of the SGTR event and recovery actions.
The durations of the time inter-vals are established by inputting the time of reactor trip (as calcu-lated by RETRAN) and assumed operator response times for key actions:
isolation of auxiliary feedwater (AFW) flow to the f aulted SG, init-iation of cooldown of the reactor coolant system (RCS), depressuriza-tion of the RCS, and termination of safety injection (SI) flow. State points (system pressures, temperatures, etc.) are established for each time interval. These state points are based on RETRAN analyses and the emergency operating procedures for recovery from a SGTR event.
For each time interval, heat and mass transfers are integrated and the.
inventories of water and steam in each SG at the end of the time inter-val are determined. Water levels in the intact and faulted SGs, possible overfill of the faulted SG, releases of steam or steam / water, releases of radioactive iodine, and 0-2 hour offsite doses are then calculated.
Consistent with the FSAR analysis, loss of offsite power is assumed to occur at the time of reactor trip.
This is also the worst case with respect to offsite doses because steam dump and the condenser are unavailable for retention of some of the leaked radioactivity.
Various assumed equipment failures can be evaluated.
In addition, the tube break may be in the hot or cold leg of the SG and iodine spiking may be in accordance with Case 1 or Case 2, as defined by NRC Standard Review Plan 15.6.3.
The analysis procedure is altered slightly to consider the postulated failure of a stuck-open atmospheric relief valve (ARV). During the depressurization of the failed SG, three additional intervals are utilized, SG pressure and consistent RCS conditions are specified for each time interval, and the time to reach the pressure at the end of each interval is calculated in an iterative manner, using the ARV flow characteristics.
These analysis procedures are programmed in Microsof t BASIC for a personal computer.
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Modified Zaloudek (Reference 0-1),
b.
Burnell (Reference D-2), or c.
Henry, 1970 Model (Reference D-3).
Comparisons with data over the range-of interest for the SGTR have been perfomed.
It has been concluded that the Burnell correlation most conservatively estimates the critical flow rates with accept-able accuracy for the range of conditions encountered in the steam generator tube rupture event.
The Burnell equation is as follows:
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=
where:
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=
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=
PSAT = saturation pressure at upstream temperature, psia 144 = conversion factor, in2/ft2 Burnell constant (see Figure D-2)
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=
V.
Flow in the Long Hot Leg Tube For the long (hot leg) segment, heat transfer across the tube decreases the fluid temperature during transit.
In addition, the fluid will undergo a large pressure drop and a probable phase change while flowing through the long tube (1/d = 952).
It was determined that for the range of SGTR conditions exper-ienced, single phase, resistance-limited flow provided a con-servatively high mass flow rate. This conclusion was based on comparison to flow rates computed with two phase conditions.and heat transfer, as described above.
D-3 Rev. 1
(2) 100% of the iodine contained in the fraction of the break flow to th2 f aulted SG that flashes upon reaching the secondary side.
(This term is conservatively included even when the RETRAN analysis shows that no steam is released from the secondary side ARVs or safety valves).
The 1 gpm leakage flow to the intact SGs is assumed not to flash.
Thus, only the first of the above terms applies to the intact SGs.
Noble gas releases to the atmosphere are calculated to be equal to 100% of the noble gas contained in reactor coolant break flow or leakage flow that reaches the secondary side. This conservatively assumes no retention in the SG water.
6.
Doses The 0-2 hour site boundary and 0-8 hour exclusion boundary doses to the thyroid and whole body are calculated in accordance with Appendix 15A of the FSAR and adult conversion factors were utilized. That is, contributions of five iodine isotopes and thirteen noble gas isotopes are summed to obtain the total doses. Values of X/Q applicable to the Callaway plant have been used, because these values are higher than those for the Wolf Creek Station and result in higher calculated doses.
References 1.
Callaway Plant Chemistry Technical Procedure CTP-ZZ-02540, "Datermination of Isotopic Specific Gamma Activity in Liquids",
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Revision 5, dated 3/30/85 2.
Wolf Creek Generating Station Procedure CHM 0 3-052, Revision 2,
" Determination of Radioactive Iodine and Dose Equivalence Iodine Analysis."
F-4 Rev.1