ML18142A067
| ML18142A067 | |
| Person / Time | |
|---|---|
| Site: | Surry, North Anna, 05000000 |
| Issue date: | 09/26/1984 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML18142A066 | List: |
| References | |
| NUDOCS 8410100694 | |
| Download: ML18142A067 (7) | |
Text
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ENCLOSURE CORE PERFORMANCE BRANCH SAFETY EVALUATION OF VEPCO EVALUATION OF THE CONTROL ROD EJECTION TRANSIENT VEP-NFE-2
(
- s410100694 s40_9_26~~-1 PDR ADOCK 05000280 P
e e Enclosure CORE PERFORMANCE BRANCH SAFETY EVALUATIO!i OF VEPCO-EVALUATION OF THE CONTROL ROD EJECTION TRAHSIENT
.VEP-NFE-2 1.0
SUMMARY
OF TOPICAL REPORT This report describes the methods developed by the Virginia Electric and Power Canpany (Vepco) for the analysis of a postulated control rod ejection transient in the Nor:th Anna and/or Surry Nuclear Power Stations.
- A description of the*rod ejection transient and a discussion of the acceptance criteria which must be met to assure the safe operation of the pl ant in the.~vent of sucn a transient is al so presented.
The RETRAN computer program is used for the analysis which is performed in two parts.
First, a point kinetics analysis is used to calculate the average core nuclear power history.
Next, a hot spot thermal-hydraulic calculation-is used to determine the--tmt-spot enthalpy and temperature transients from which the am9_ynt of fuel damage and radio-logical consequences are assessed.
Detailed descriptions of the calculational model and the techniques _employed in the analysis are presented.
The results of sensitivity studies used to quantify the effect of uncertainties in important core parameters and model,!_ng assumptions on the model's predictions are shown.
Also presented are comparisons of the results of the fuel vendor (Westinghouse) methodology with the Vepco methoctol ogy as wel 1 as comparisons of point kinetics results 1-1ith three-dimensional space-time kinetics mo'del results.
2.0 STAFF EVALUATION We have reviewed the subject report, including the mathematical models and analytical procedures and methods.
The RETRAN computer program is*
the principal calcu1ationa1 tool.. A point ~netics analysis is used
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to ca1culatE{the average core nuclear power during a. rod ejection transient.
The hot spot (hottest fuel pin) enthalpy and temperature transients are determined fro~ a hot spot thermal-hydraulic calcu-lation. This is the usual procedure used by.,the nuclear industry to analyze the spa.ti ally dependent transient with a point kinetics model and has been found to be acceptable and usually conservative.
Al though the RETRAI~ program is used to analyze the rod ejection transient,. the report emphasizes the implementation of the RETRAN models -since the detailed code description is available in a separate document.
Therefore, we did not review the RETRAN program, per se, but rather the qualification of its use in determining the consequences of a rod ejection transient.
The staff position, as well as that of most of the re~ctor vendors and licensees, has been to limit the average fuel pellet enthalpy at the*hot spot following a rod ejection transient-to 280 calories per gram (cal/gm).
This was based primac.ily on the results of the SPERT tests which showed that, in general, fuel failure consequences for uo2 have been insignificant below 300 cal/gm for both irradiated and unirradiated fuel rods as far as rapid fragmentation and dispersal of fuel and* cladding into* the coolant are concerned.
In this report, Vepco has chosen. more stringent design limits.
The limiting. fuel failure criterion has been reduced to 225 cal/gm for __ unirradiated rods and 200 cal/gm for irradiated rods.
Since this is a conservative revision, the staff finds these criteria acceptable.
Vepco proposes. a clad temperature limitation of 2700°F as* the temper-ature above which clad ernbrittlement may be expected.
Although this is several hundred degrees above the maximum clad temperature limitation proposed in the ECCS !;:riteria, the staff feels this is adequate in view of the relatively short time at temperature and the highly localized effect of this transient.
The staff has no limiting temperature criterion for rod ejection transients.
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The neutron{cs model including the reactivity insertion, neutron kinetics parameters, fuel and moderator temperature feedback, and reactor trip assumptions described in the report are in conformance
\\'.'ith the recommendations of USNRC Regulatory-,Guide 1.77 and are acceptable.
Since RETRAN uses a point kinetics neutronics model, the effect of locally peaked core flux shapes due to the rod ejectipn is not included in the Doppler reactivity calculation.
Therefore, a power weighting factor (PWF) is used to modify the Doppler reactivity versus core average fuel temperature data in order to better approximate the local reactivity effects.
The TWINKLE computer program is used by Vepco to justify the use of a PWF.
Since TWINKLE is a three-dimensional spatial kinetics**,code_, it acc;:ounts for the highly localized reactivity effects of a control rod ejection accident more realistically than a point kinetics calculat"ion and is, therefore, an acceptable method t'or verifying the PWF.
For cases initiated fran both hot.zero power and hot,.*
full power initial conditions, the RETRAN point kinetics predictions are shown to be conservative compared to the TWINKLE three-dimensional predictions and, therefore, the described RETRAN model is acceptable.
The radiological effects of a rod ejection transient have been addressed generically in the Surry. 1 and 2 and the North Anna 1 and 2 Updated ~inal Safety Analysis Reports a,s well as in the Westinghouse "Evaluation of the Rod Ejection* Accident in Westinghouse Pressurized Water Reactors..
Using Spatial Kinetics l~ethods" (WCAP-758.8):
In thi-s latter report, a
- departure from nucleate boiling (DNB) analysis was performed using the trans~ent THINC-III code for a worst-case three-dimensional rod eiection transient.
The results indicated that '.1ess than 10% of the fuel rods enter DNB.
Since one of the NRC acceptance criteria for this transient requires the assumption that all of the fuel rods which experience DNB release their entir_e gap inventory of fission pr~ducts to the coolant, the generic result indicates that less than 10% of the core will release fission products.
The source term for the radiological release calcu-lations for these and other Westinghouse designed plants is, therefore, based on 10% of the fuel experierking clad failure and releasing their
e entire gap activity.
This is in conformance with Regt:1latory Guide 1.77 and, hence, acceptable.
Since there is no fuel dispersal into the coolant, the pressure surge during the transient may be calculated based on conventional heat transfer methods.
Generic pressure surge calculations for the most severe excess addition of energy to the coolant indicate that any system overpressuri za ti on due to a rod ejection transient vri 11 meet the NRC criterion of being less than that pressure which would cause stresses*to exceed the Service Limit C as defined in Section III of the ASME Boil er *anct Pressure Vessel Code.
Vepco has analyzed the rod ejection transient using Westinghouse codes and techniques and comparect their results to the. results obtained by Westinghouse.
\\.le have reviewed these comparisons and concur that Vepco can adequately-replicate Westinghouse results.
We have also revi evte'd comparisons between results obtained with the NRC approved
\\.Jestinghouse methodology with those obtain-eel with the Vepco methodology presented in th-e topical report using RETRAN.
We concur that these comparisons demonstrate the acceptability of the Vepco methodology.
A comparison is also presented between the Vepco calculated results using both Westinghouse _and Vepco methodologies and results obtained by Vepco using the \\.Jes ti nghouse three-dimensional space-time kinetics..
code, TWINKLE.
The staff finds that this comparison.adequately demonstrates the conservatism of the Vepco rod ejection calculational method.
3.0 EVALUATION PR6CEDURE The staff has reviewed the report ~ithin the guidelines provided by Sections 4.3 and 15.4.8 of the Standard Review Plan (NUREG-75/087) and by Regulatory Guide 1.77.
Part of our review was based on our familiarity with and comparison of similar analyses for control rod
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ejection traris i ents provided in topi'cal reports by other P~JR vendors.
The staff al so had the benefit-of a meeting with VEPCO concerning the report.
4.0 REGULATORY POSITION The subject report (VEP-NFE-2) provides an acceptable method for analyzing the control rod eJecti on event for Vepco I s Surry and North Anna Nuclear Power Stations.
This methodology may be referenced in future Vepco license applications.
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SALP INPUT FOR VEP-NFE-2 *TOPICAL REPORT CORE PERFORMANCE BRANCH The following two functional areas are applicable to our review of the topical rep.art:
- 1.
Management Involvement.
The presentation to the NRC and the information contained in the
.topical report shows well. stated and thorough information relevant to the subject matter.
Rating:
Category 1
- 2.
Approach to Resolution of Technical Issues.
The docu1rientation demonstrates a clear understanding of the required assumptions and calculational methods and exhibits conservatism in v
those assumptions ahd methods.
Ka ti.ng:
Category 1