ML19330C138
| ML19330C138 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 07/25/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19330C137 | List: |
| References | |
| NUDOCS 8008080003 | |
| Download: ML19330C138 (6) | |
Text
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'g UNITED STATES p,
NUCLEAR RE'GULATORY COMMISSION
,E4 WASHINGTON. D. C. 20666 d
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 46 TO FACILITY OPERATING LICENSE NO. NPF-1 PORTLAND GENERAL ELECTRIC COMPANY THE CITY OF EUGENE, OREGON PACIFIC POWER AND LIGHT COMPANY TROJAN NUCLEAR PLANT DOCKET NO. 50-344 Introduction By letter dated January 5,1979, Portland General Electric Cor.pany, iojaii Nuclear et al., requested changes to the Technical Specifications (TS) for operation of the T Plant in Columbia County, Oregon. The proposed changes relate to the use of the Westinghouse Improved Thermal Design Procedure and the WRB-1 Critical Heat Flux Cor-relation. Both relate to methods and procedures for detemining safety limits and reactor protection system settings for protection from departure from nucleate boili_ng (DNB) in the reactor core.
Discussion and Evaluation One of the safety limits which must be met by every operating PWR is based on the requirement that the fuel rods must not experience DN8 during Condition 1 (normal operation) and Condition 2 (moderate frequency) events. This safety limit is expressed as a DNB Ratio. When the Trojan reactor was licensed, the method used to calculate the DNB Ratio was based on the Westinghouse method existing at that time (Reference 1).
Since that time, Westinghouse has proposed (Reference 2), and the staff has approved (Reference 3), a new method of calculating the DNB Ratio for Condition 1 and Condition 2 events.
It is this new method that Portland General Electric (PGE, the licensee) has proposed to use for the Trojan reactor.
In the original method the accident analysis is done with conservative values (resulting in low DNB Ratios) assigned to the significant variables (such as power, power distri-bution, reactor coolant system pressure, coolant flow, etc.). This method implies that the values of these input variables will be at their most conservative values simul-taneously. The conservative values are detemined by adding or subtracting the appro-priate uncertainty in the variable to/from its nominal value.
In the Improved Themal Design Procedure the safety analysis is performed with these same variables at their nominal values. A new DNB Ratio safety limit is defined which includes the uncertainties in these variables. The details of this procedure are described in Reference 3.
In addition to these differences in the method of including uncertainties in the cal-culation of the DNB Ratio, the Improved Thermal Design Procedure employs a different criterion for the safety limit. The fixed-value method used a safety limit which met the following criterion as stated in the Trojan FSAR:
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Trojan j DNB will not occur on at least 95% of the limiting fuel rods during normal operation and operational transients and any conditions arisi.ng from faults of moderate frequency at a 95% confidence level (Section 4.4.1.1).
This criterion was met with the W-3 Critical Heat Flux (rHF) correlation with appropriate multipliers which gives a DNB Ratio safety limit of less than the traditional value of 1.3.
However, the licensee has used the 1.3 value.
The Improved Thermal Design Procedure uses the following criterion (from Refer-ence2):-
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Considering plant parameter uncertainties, there must be at least a 95%
probability that the minimum DNB Ratio of the limiting power rod during Condition 1 and 2 events is greater thart or equal to the DNB Ratio limit of the DNB correlation being used.
The DNB Ratio limit for the corre-lation is established based on the variance of the correlation such that there is a 95% probability with 95% confidence that DNB will not occur when the calculated DNS Ratio is at the DNB Ratio limit.
j The s.ignificance of the change in the safety limit criterion is shown in Figure 1 (taken from Reference 3). While the fixed value method used the 1.30 value (with the W-3 correlation) as a limit, the Improved Thermal Design Procedure allows this limit to be exceeded with a 5% probability. The difference is subtle but signi-ficant.
While the fixed value method establishes a safety limit which gives a 95% probability the DNB will not occur with a 95% confidence, the limit of the Improved Thermal Design Procedure gives a 95% probability that the safety limit of tne fixed va Le method will not be exceedec witn a 95% ccnfidence.
i,a.im:: of tne Improved Thermal Design Procedure is called the design limit.
In addition to the use of the Improved Themal Design procedure, the licensee also employed a new DNB correlation called WRB-1. The derivation and description of this correlation is given in Reference 4.
The staff Safety Evaluation Report approving use of this correlation is given in Reference 5.
The value of the safety limit for this correlation is 1.17. This value meets the criterion for a DNB safety limit as stated in the Trojan FSAR given previously.
The value of the design limit determined for Trojan by the Improved Themal Design Procedure with the WRB-1 correlation is 1.36 for thimble coldwall cells (three fuel rods and a guide tube) and 1.38 for typical cells (four fuel rods) without con-sideration of fuel rod bowing.
In addition, the design limit has been conservatively increased by the licensee to incorporate an arbitrary 20.2% margin in DNB Ratio.
The design DNB Ratio values which include the 20.2% DNB Ratio margin are 1.71 and 1.73 for thimble wall and typical cells, respectively. These design limits were use.d in the accident analysis and set point determinations.
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f DNBR VARI ATtou FOR THINBLE CELL CONSIDERING VARI ABLE VALUES OF DESIGN PARANETERS 5%
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a DNBR (VARI ABLE) = !.30 LlWIT DNSR = 1.46 i
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Figure 1 Interpretation of DNB Ratio Safety Limit With The Improved Thennal Design Procedure l
-4.
The staff Safety Ev41uation Report on the Improved 7hermal Design Procedure
'(Reference 5) states several requirements which must be met in order to use the new procedure. The first requirement is that all Condition 1 and Condition 2 events which must meet the DNB Ratio safety limit should be reanalyzed to show that they satisfy the design ' limit of the Improved Themal Design Procedure.
The second requirement is that the uncertainties assumed in determining the design-limit must be consistent with the instrumentation and design of the parti-cular reactor.
The licensee's submittal (Reference 2) meets the first require-ment. All Condition 1 and 2 events listed in Table 3 of the staff Safety Evalua-tion Report have either been reanalyzed or are bounded by some other analysis.
The second requirement concerning applicability of the uncertainties used in the analysis was discussed in detail by the licensee in Reference 6.
In particular, the licensee provided in Reference 6 a description of sensors, process equipment, computer and readout devices for pressure, reactor coolant flow, reactor power and reactor coolant temperatures.
For each element, the accuracy, drift and range were given. The standard deviation for each measurement was determined by summing the variances of each component to obtain the total system variance. The standard deviation is the square root of the variance.
These standard deviations were compared with the corresponding values used in the analysis.
The analysis values were bounding.
The staff, with our consultants at Idaho Nuclear Engineering Laboratory (INEL),
reviewed Reference 6.
We agree with the licensee's analysis of the design accuracy of the pressure, temperature, power, and flow measuring systems used at the Trojan "uclear phnt.
We also reviewed the frequency and methods of surveillance used by the licensel to assure that the values he stated for his instrumentation would actually be. met.
The TS require periodic checks of pressure, temperature, power, and flow instru-ments to ensure proper operation. A functional channel check done monthly on each of these instruments verifies instrument performance by the injection of a signal into the circuitry to ensure operability including alarm and trip initiation features.
The licensee submitted data based on operating history to demonstrate a high degree of confidence that the required instruments would perform within their design accuracies. While data on reactor coolant flow instrumentation was not included, data on temperature, pressure, and power shows that only one out-of-limits condi-tion occurred over a 39-month period which included 504 functional channel checks of these instruments.
This data represents a success rate of 99.8..
The licensee also submitted data on several instruments showing calibration da'ta.
These data indicate that the design accuracies submitted by the licensee are reasonable based on the actual calibration tests.
-5 Based on the design instrument accuracies p'rovided by the licensee, the pressurizer pressure, average temperature, reactor powc., and reactor coolant flow instrument accuracies are within the limits assumed by the Improved Thermal Design Procedure.for preventing a DNB occurrence. Data provided on instrument Performance and the requirements of the TS indicate that adequate measures exist at the Trojan Nuclear Plant to monitor these instiuments and to ensure the instruments will perform with a high degree of confidence within their design accuracies.
Based on the data on instrument performance provided by the licensee and the requirements of the TS, adequate measures exist at the Trojan Nuclear Plant to monitor the required instrumentation to ensure continued operation within the limits specified in the safety analysis.
The licensee has provided adequate analyses -to show that the core is protected from DNB for all Condition 1 and Condition 2 events.
For the above' reasons the application of the Westinghouse Improved Thermal Design Procedure in conjunction with the WRB-1 correlation is acceptable.
Environmental Consideration We have determined that the amendment does not authorize a change
-in effluent types'or total amounts nor an increase in power level and.will not result in any significant environmental impa:t. Having made this determination, we have further concluded that the amendment involve: an 1:tien whi:5i insigni":ar.t f-:r the stand:: int of environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement, or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of this amendment.
Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not.be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to.the common defense and security or to the health and safety of the public.
Dated: July 25, 198'O
Refersaces
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1.
Final Safety Analysis Report, Trojan Nuclear Plant, Section 4.4 2.
Letter from C. Goodwin, PGE, to A. Schwencer, USNRC, License Change Application 49, January 5,1979.
3.
CheYemer, H., et. al, " Improved Thermal Design Proceuure", Westinghouse Elsctric Corporation, WCAP 8567, July'1975.
4.
Motley, F., et. al, "New Westinghouse Correlation WRB-1 For Predicting Critical Heat Flux in Rod Bundles With Mixing Vane Grids", Westinghouse Electric Corporation, WCAP 8762, July 1975.
5.
Letter from J. F. Stolz, USNRC, to C. Eicheldinger, Westinghouse Electric Corporation, April 19, 1978.
6.
Letter from C. Goodwin, PGE, to A. Schwencer, USNRC, " Additional Information Required for Trojan to Justify Use of Improved Thermal Desigr, Procedure",
November 5, 1979.
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