ML20246F691
| ML20246F691 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 06/30/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20246F683 | List: |
| References | |
| NUDOCS 8907130360 | |
| Download: ML20246F691 (6) | |
Text
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y SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT N0.100 TO FACILITY OPERATING LICENSE NOS. NPF-4 AND NPF-7 VIRGINI A ELECTRIC AND POWER COMPANY CLD DOMINION ELECTRIC COOPERATIVE NORTH ANNA POWER STATION, UNIT NOS.1 AND 2 DOCKET NOS. 50-338 AND 50-339 INTRODUCTION By letter dated June 17, 1987, licensee) submitted proposed changes to the Technical Specificationsthe Virginia Ele (the the North Anna Power Station, Units No. I and 2 (NA-1&2). The changes support ~
the implementation of the licensee's Statistical DNBR (Departure from Nucleate Boiling Ratio) Evaluation Methodology as documented in Topical Report VEP-NE-2 of the same name. The principal Updated Final Safety Analysis Report (UFSAR)
Chapter 15 DNB events have been analyzed under the new DNB methodology. Some TS changes are needed as a result of these analyses.
Several NA-1&2 TS need to be changed to incorporate the revised DNBR ratio limit and the results of the associated transient analysis. The proposed changes to the NA-1&2 TS include a less restrictive negative moderator temperature coefficient (MTC) limit.
The new methodology employs Monte Carlo methods to evaluate the DNBR sensitivity to key parameters. The methodology is similar to those which have been approved by NRC for some of the vendors and uses the WRB-1 CHF correlation. The topical report "VEP-NE-2" was approved by the NRC on May 28, 1987. We found the methodology acceptable provided that four conditions were met when a plant-specific submittal was made. These four conditions are:
1.
The choice of " Nominal Statepoints" must be justified.
2.
For the statistically treated parameters, the uncertainty distributions must be justified.
3.
The model uncertainty must be substantiated.
4.
Topical Report COBRA /WRB-1 (VEP-NE-3) must be approved.
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2.0 _ DISCUSSION In December 1988, the MTC TS change was required for continued operation of NA-1 for the remainder of Cycle 7.
At that time our review of the licensee's sub-
.mittal had proceeded to the point that we could. approve the MTC TS change for L
the remainder of NA-1 Cycle-7. On January 3, 1989, Amendment No. 112 was issued L
to Facility Operating License No. NPF-7 for NA-1 which approved the MTC TS for l
the remainder of Cycle 7.
We have now completed our review for NA-2.
Each of the four conditions stated in the NRC Safety Evaluation (SE) for VEP-NE-2 is addressed below.
2.1 Nominal Statepoints L
To satisfy the condition on nominal statepoints, the choice of the nominal statepoints must be shown to maximize the DNBR standard deviation (and therefore, the DNBR limit) over the proposed range of applicability. The licensee's approach was to perform Monte Carlo calculations at core thermal limit and low
- flow statepoints. These conditions spanned the pressure range between the high and low trip setpoints, inlet temperature between a boundary cocidown event and a maximum heatup, power up to the 118 percent overpower limit and a boundary low flow event. The analysis consisted of 10 sets of 2000 calculations, each performed over the full range of normal operation and anticipated transient conditions.
The standard deviations were then plotted as a function.of state-point temperature. The data showed a clear dependence on temperature. A..
regression analysis was then performed'and the residuals were plotted showing no trends. This indicated that the standard deviation was a function of temperature only.
Therefore, the limiting statepoint was specified to be Statepoint 5 with power equal to 118 percent, inlet temperature equal to 538.6*F, pressurizer pressure equal to 1860 psia, and flow equal to 92 percent. The DNBR mean was 1.28 with a standard deviation of 0.1572. This satisfies the condition on selection of nominal statepoints.
2.2 Uncertainty of Statistically Treated Parameters The statistically. treated parameters are core power, pressurizer pressure, inlet temperature, vessel mass flow, core bypass flow and the nuclear and engineering enthalpy rise factors. The uncertainties for core power, pressurizer pressure, inlet temperature and vessel flow are quantified in WCAP-1203. The analysis was performed under the standard Westinghouse uncertainty analysis methodology.
The total core bypass flow and its uncertainty was confirmed as being bounded '
by the NA-1&2 core uprating Improved Thermal Design Procedure (ITDP) analysis values.
Because of the difficulty in characterizing the form of the uncertainty distribution, the implementation analysis assumed that the probability was uniformly spread over a much lar0er range than was justified by the sum of the components.
The nuclear enthalpy-rise factor uncertainty was based on availabic measurement /
predictive data, which consisted of over 11,000 points taken from 9 cycles of operation at both NA-1&2. The error of prediction relative to the radial power factor measurement ([P-M]/M) yielded a mean value of 0.1 percent with a standard
deviation of 1.55 percent. The non-zero mean is conservatively positive and as a conservative measure, a standard deviation of 2 percent was used in the l
implementation analysis. The data was shown to be a normal distribution curve.
The engineering enthalpy rise uncertainty factor consists primarily of the uncertainty in hot channel power and flow. These factors were quantified by means of a closed-channel calculation, in which boundary values of high hot channel power and low flow were employed. A uniformly distributed 2 percent uncertainty was found to conservatively bound the results. This satisfies the condition on the uncertainty of statistically treated parameters.
2.3 Model Uncertainty The model uncertainty was included to account for differences between the 6 channel COBRA model, which was used for the Monte Carlo calculations, and the 25 channel COBRA production model, used for performing DNBR calculations.
Comparisons show that the 6 channel model DNBR standard deviation is consis-tently much larger than the standard deviation produced by the 25 channel production model. However, a model uncertainty was quantified as an upper confidence limit on the S (model) where the S (model) is the standard deviation on the ratio of the 6 to 25 channel model DNBR using 100 random statepoints. This satisfies the condition on model uncertainty.
2.4 COBRA /WRB-1 Verification (VEP-NE-3)
Our review of VEP-NE-3 has been completed. Our SER approving the topical COBRA /WRB-1 (VEP-NE-3) was dated June 14, 1989.
2.5 TS Changes As noted above, implementation of the licensee's DNBR evaluation methodology requires that the NA-1&2 TS be updated to reflect the change in the plant DNBR licensing basis. These TS changes are described below.
2.5.1 TS 3/4.1.1 Most Negative MTC Limit The change in the Limiting (SR 4.1.1.4) for the MTC are intended to provide an Condition for Operation (LCO 3.1.1.4) and the Surveillance Requirements end-of-cycle and associated trigger values which are appropriate for current NA-1&2 fuel cycles.
The revised limit and trigger values are based on a revised safety analysis of the UFSAR Chapter 15 transients which are sensitive to the most negative MTC parameter.
It is noted that the DNB design limit was not violated in any of the analyzed transients.
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2.5.2 TS 3/4.3.1 Pressurizer Water Level _ Response Time The change to Table 3.3-2, which is referenced by LC0 3.3.1.1, reflects a new requirement to have the pressurizer water level response time be less than or equal to 2.0 seconds. A response time requirement has been added to protect against filling the pressurizer prior to the actuation of the overtemperature delta T reactor trip.
2.5.3 TS B 2.1.1 Full Core DNB Probability Criterion The change in bases Section 2.1.1 adds a new criterion in the establishment of the DNBR limit for the licensee's Statistical DNB Methodology.
Previously, traditional analyses and ITDP analyses considered only peak pin DNB probability.
Plant operation required that, for normal operation and Condition II operation, the peak pin avoid DNB with 95 percent probability at a 95 percent confidence level.
The new methodology retains this criterion, and adds an additional criterion that the DNB probability of every rod, when summed over the whole core, shows that at least 99.9 percent of the core is expected to remain in the nucleate boiling regime.
2.5.4 TS B 3/4.1 MTC Surveillance Requirements The changes to bases Section 3/4.1 are related to the new MTC limit and surveil-lance requirements. Additionally, the reference to the moderator density coefficient (MDC) is deleted because it is no longer related to the safety analyses performed by the licensee. The MDC parameter was used in the previous safety analyses performed by the licensee's fuel vendor, Westinghouse Electric Corporation. Since the safety analyses performed by the licensee uses tempera-ture instead of density to specify moderator reactivity feedback, it is preferable to use MTC in the TS.
An additional benefit of this approach is that the relationship between the TS limit and the safety analysis limit can be more clearly defined.
In fact, the two limits differ only by the measurement uncertainty and a correction for Bank D insertion. The revised bases section makes the connection between the two limits clearer.
2.5.5 TS B 3/4.2.3 DNBR Limits TS B 3/4.2.3 has been modified to reflect the revised DNBR limit n's obtained with the new methodology. The new safety analysis DNBR limit is 1.26; the l
addition of 13.7 percent retained DNBR margin yields a design DNBR limit of j
1.46.
Separate values were not derived for the typical and thinble cell, since i
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the single limit was shown to be bounding for both. The retained margin is used for such applications as compensation of the rod bow penalty, for example.
The rod bow penalty is Westinghouse proprietary information and is not listed in the FSAR.
However, a thorough discussion of the rod bow penalty is proviced in the FSAR, and the sources of the appropriate numerical values are referenced.
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b 2.5.6 TS B 3/4.2.5 DNB Parameter Surveillance The change to bases Section 3/4.2.5 clarifies the treatment of measurement uncertainties. The licensee has performed analyses to show that the measurement uncertainties on the DNB parameters can be offset by the retained 1
DNB margin and need not be accounted for by the plant operations staff.
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i 3.0 EVALUATION i
. Based on our review of the licensee's June 17, 1987 submittal, we conclude that the requested TS changes are acceptable. This conclusion is based on the fact that the plant-specific iluplementation submittal for use of the licensee's Statistical DNBR Methodology adequately addresses three of the four conditions specified in the NRC Safety Evaluation dated May 28, 1987 for the staff's review of VEP-NE-2, " Statistical DNBR Evaluation Methodology."
In addition, the staff's review of the fourth condition, the approval of " COBRA /WRB-1 (VEP-NE-3)," is now complete and the topical has been approved. Therefore, based on all of the above, the TS changes described above are acceptable for NA-2.
In addition, these TS changes, as noted above, were previously approved for NA-1 for Cycle 7 only and continued use of these TS is hereby approved for NA-1 Cycle 8, Cycle 9, etc.
4.0 ENVIRONMENTAL CONSIDERATION
This amendment involves a change in the installation or use of a facility com-ponent located within the restricted area as defined in 10 CFR Part 20 and changes to surveillance requirements.
The staff has determined that the amend-ment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cunulative occupational radiation exposure. The Commission has previously published a proposed finding that the amendment involves no significant hazards consideration and there has been no public connent on such finding. Accordingly, the amendment meets the eligibility criteria for categorical e.xclusion set forth in 10 CFR 551.22(c)(9). Pursuant to 10 CFR 551.22(b), no environmental impact statement or environmental assess-ment need be prepared in connection with the issuance of the amendment.
CONCLUSION We have concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations, and the issuance of the amendment will not be inimical to the connon defense and security or to the health and safety of the public.
Cate: June 30, 1989 Pr_incipal Contributor:
Margaret Chatterton
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DATEDi #" " ' "
AMENDMENT NO.100 TO FACILITY OPERATING LICENSE NO. NPF-7-NORTH ANNA UNIT 2
- CM dilhFA NRC & Local PORs PDII-2 Reading S. Varga,14/E/4 G. Lainas, 14/H/3 H. Berkow D. Miller L. Engle OGC D. Hagan, 3302 MNBB E. Jordan, 3302 MNBB B. Grimes,)9/A/2 T. Meek (4
.Wanda Jones, P-130A J. Calvo, 11/F/23 ACRS (10)
GPA/PA OC/LFMB B. Sinkule, R-II OFoi I,
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