ML19352A976
| ML19352A976 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek, Callaway  |
| Issue date: | 05/27/1981 |
| From: | Petrick N STANDARDIZED NUCLEAR UNIT POWER PLANT SYSTEM |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| SLNRC-81-36, NUDOCS 8106020472 | |
| Download: ML19352A976 (8) | |
Text
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- e sNUPPS Stenderdised Nueleer Unit Power Mont System 5 Choke Cherry Road Nicholas A. Petrick v
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\\9 Y,78 May 27, 1981 7.k 3@$ ((G/
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- SLNRC 81-36 FILE: 0541 SUBJ: SNUPPS FSAR - NRC Request for Additional Information y1 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Comission Washington, D. C.
20555 Docket Numbt rs:
STN 50-482, STN 50-483, STN 50-486
Reference:
NRC (Tedesco) letter to Union Electric (Bryan) and Kansas Gas and Electric (Koester), dated April 21, 1981: Same subject
Dear Mr. Denton:
The referenced letter requested information in the area of reactor thermal-hydraulics.
The enclosure to this letter provides the requested information and will be incorporated into the SNUPPS FSAR in a future revision.
Very truly yours, R-4<sc Nicholas A. Petrick RLS/mtk j
Enclosure 5
i cc:
J. K. Bryan
//
l G. L. Koester iE D. T. McPhee KCPL f
T. E. Vandel USNRC/WC W. Hansen USNRC/ Cal g
i 81060204
SNUPPS Q492.2 The effects of fuel rod bowing must be included in the thermal-hydraulic design.
The predicted extent of rod bow (gap closure) versus exposure and the effect of rod bowing on DNBR must be addressed.
Use of the staff report " Revised Interim Safety Evaluation Report on the Effects of Fuel Rod Bowing on Thermal Margin calculations for Light Water Reactors,"
February 16, 1977, represents an accept-ably conservative treatment of rod bowing.
RESPONSE
The DNB analyses described in the FSAR of the SNUPPS 17x17 core were performed such that generic DNBR margins described in the " Revised Interim Safety Evaluation Report on Effects of Fuel Rod Bowing on Thermal Margin Calculations for Light Water Reactors (Revision 1)," February 16, 1977, are avail-able for offsetting rod bow penalties.
The appropriate rod bow penalty and any operating restriction in the technical specifications, if required, will be addressed prior to the issuance of the Operating License of this core:
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492.2-1
SNUPPS Q492.3 operating experience on two pressurized water reactors (not of the Westinghouse design) indicate that significant reduction in core flow rate can occur over a relatively short period of time as a result of crud deposition on the fuel rods.
In establishing the Technical Specifications for callaway and Wolf Creek we will require provisions to assure that the minimum design flow rates are not exceeded.
Therefore, provide a description of the flow measurements capability for Callaway and Wolf Creek as well as a description of the proce-dures to measure flow and the actions to be taken in the event of an indication of lower than design flow.
RESPONSE
Operating experience to date has indicated that a flow resistance-allowance for possible crud deposition is not required.
There has been no detectable long-term flow reduction reported at any Westinghouse plant. ' Inspection of the inside surfaces of steam generator tubes removed from operating plants has confirmed that there is no significant surface deposition that would affect system flow.
Although all of the coolant piping surfaces have not been inspected, the small piping friction contribution to the total system resistance and the lack of significant deposition on piping near steam generator nozzles support the conclusion that an allowance for piping deposition is not necessary.
The effect of crud enters into the calculation of core pressure drop through the fuel rod frictional component by use of a surface roughness factor.
Present analyses utilize a surface roughness value which is a factor of three greater than the best estimate obtained from crud sampling from several operating Westinghouse reactors.
The operator has at his disposal several methods of detect-ing significant RCS flow reduction, these are:
a.
Flow meter on each RCS loop.
b.
If operating in an automatic control rod mode (T held constant) a reduction in reactor power wou18 be present for sigr.ificant reductions in RCS flow.
If operating in a manual control rod mode (power c.
held constant) an increase in AT across the core would be present for significant reductions in flow.
492.3-1
SNUPPS d.
Local changes in flow could be indicated by incore flux maps (assuming significant changes in local power), and e.
Core exit thermocouple readings.
The operator will verify flow, perform calorimetric power checks, and r=4.- incore flux maps as required by the Tech-nical Specifications.
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'Q492.4 The NRC approval of the THINC-IV code, for use in the thermal-hydraulic design, indicates that the pressure gradient at the core exit must be modeled.
I Provide a revised THINC-IV calculation at the steady state reactor design conditions including the modeling of the core exit radial pressure gradient.
Provide the following specific infor-macion from that calculation:
1.
minimum DNB ratio (value and location) 2.
hot channel flow vs. axial position 1
3.
hot channel enthalpy vs. axial position 4.
hot channel void fraction vs. axial position i
5.
the assumed core exit pressure gradient.
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RESPONSE
On October 25, 1977, Westinghouse met with the NRC to discuss the effects of nonuniform upper plenum pressure distribution as part of the NRC staff's review of REGAR-414.
The Westing-house material presented at that meeting was tfansmitted to the NRC via letter NS-CE-1591, dated November 2, 1977, from C. Eicheldinger (Westinghouse) to J. F. Stolz (NRC).
This letter addresses the THINC-IV information requested by Question 492.4, and is applicable to all Westinghouse 4-loop plants, including the SNUPPS units.
In addition, this issue was pursued further by the NRC during the McGuire FSAR review.
The McGuire fuel is iden-tical to the SNUPPS fuel, and the same thermal-hydraulic models and correlations were used.
As a result of this review, the staff concluded that+his issue was adequately resolved.
This conclusion is equally applicable to the SNUPP5 units.
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492.4-1 L
SNUPPS Q492.5 Insufficient information has been provided to justify the design power level of 2389 Mwt (70% of full power) during three-loop operation.
Tempera-ture differences in the active cold legs of a few degrees could exist during three-loop operation.
Therefore a radial power tilt and an increase in enthalpy rise factor could result.
As a result, we request that a complete detailed description of the following items will be provided:
1.
The method of determining the temperature distribution among the cold legs and the associated radial power tilt; 2.
The method of accounting for differences (if any) in the three-loop thermal-hydaulic design; 3.
The instrumentation available and monitoring procedures during three-loop operation; 4.
The DNBR Technical Specification and how it will be implemented for three-loop operation; 5.
The reactor protective system setpoints related to DNBR protection and how.they are generated; 6.
The effects of anticipated operational occur-rences on the cold leg temperature distribu-tion and how this effect is included in the design.
RESPONSE
This question is not applicable to the SNUPPS Plants, since they do not currently plan to operate in the N-1 mode.
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492.5-1
SNUPPS Q492.6 Please state your intent regarding the use of Westinghouse optimized fuel assembly in your plant.
If the use of this design is being con-sidered, provide a discussion of the status and schedule for any revised submittals.
RESPONSE
The SNUPPS Plants do not currently plan to incorporate Westinghouse opt'.mized fuel for the first fuel cycles.
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Q492.7 Please state your intent regarding the use of the Westinghouse " Improved Thermal Design Procedure" described in WCAP-8567, dated July, 1975.
If you intend to use these methods, responses to the following questions will be required:
(a)
Provide a block diagram depicting sensor, process equipment, computer, and readout devices for each parameter channel used in the uncertainty analysis.
Within each ele-ment of the block diagram, identify the accuracy, drift, range, span, operating limits and setpoints.
Identify the overall accuracy of each channel transmitter to final output and specify the minimum acceptable accuracy for use with the new procedure.
Also identify the overall accuracy of the output value and maximum accuracy requirements for each input channel of this final output device.
(b)
Discuss the method (s) for incorporating environmental effects (e.g., noise, EMI) on instrument channels into the uncertainty analysis.
(c)
Provide data to verify that the plant instru-ments will perform with a high degree of confidence, within their design accuracies.
This information may be obtained from oper-ating history of identical instruments in-stallet in other plants.
This request per-tains to the instruments affecting the un-certainties in the design procedure (as identified in question 1 above), the over-temperature AT trip, the high flow trip, the low pressure trip and the pump voltage trip.
(d)
Provide the ranges of applicability of sen-sitivity factors.
t (e)
Demonstrate that the linearity assumption of equation 3-8 in WCAP-8567 is valid when the WR3-1 correlation is used.
PESPONSE
'he Westinghouse Improved Thermal Design Procedure is not l
'u.rantly planned to be used in the SNUPPS applications.
l 492.7-1
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