ML20054C321
| ML20054C321 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 04/13/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20054C320 | List: |
| References | |
| NUDOCS 8204200372 | |
| Download: ML20054C321 (5) | |
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SAFETY EVALUATION BY ThE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENTS NO. 39 AND NO. 20 TO FACILITY OPERATING LICENSE N05. NPF-4 AND NPF-7 VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATIOH, UNITS NO. 1 AND NO. 2 DOCKET NOS. 50-338 AND 50-339 Introduction By letter dated November 12, 1981, the Virginia Electric and Power Company (the licensee) requested a change to the Technical Specifications (TS) for the North Anna Power Station, Units No. 1 and 2 (NA-182). The requested change to the TS is based on the licensee's reanalysis of the Emergency Core Cooling System (ECCS) performance for the postulated large break Loss-of-Coolant-Accident (LOCA) assuming seven (7) percent uniform plugging of steam generator tubes. The reanalysis was performed with the NRC approved February 1978 ver-sion of the Westinghouse LOCA-ECCS evaluation model.
By letter dated February 12, 1982 the licensee provided supplemental information regarding non-LOCA accidents and transients which could be affected by the seven (7) percent uniform steam generator tube plugging.
The above reanalysis results in a newly adjusted overall heat flux hot channel factor of F equals 2.14 for which the licensee has requested a change to the l
I NA-1 & 2 TS Discussion Significar Input:
Certain.,onservative assumptions were made for the NA-1&2 LOCA-ECCS reanalysis as requi' ed by Appendix K to 10 CFR Part 50.
The assumptions pertain to the condittor. of the reactor and associated safety system equipment at the time that a L(,.A is assumed to occur and includes such items as the core peaking factors, the containment pressure, and the performance of the ECCS.
All pre-vious LOCA ECCS submittals for NA-1&2 have shown that the limiting double ended break size equals 0.4.
For this reanO ysis, the licensee has also explicitly determined ^ hat the limiting double ended break size is also 0.4.
All assumptions and initial operating conditions used in the licensee's reanalysis I
are the same as those used in the presently NRC approved LOCA-ECCS analysis for NA-1&2 with the following exceptions.
l 8204200372 820413 DR ADOCK 0500033g PDR
. Significant changes in the reanalysis reflect the operational conditions and limits necessary to allow full power operation for steam generator tube plug-ging levels up to seven (7) percent. The currently approved analysis allows for five (5) percent tube plugging. A core inlet temperature of 548.6 degrees Fahrenheit (F) was used in the reanalysis. This core inlet temperature value was adjusted from NA-1&2 operational data to encompass the five (5) to seven (7) percent steam generator tube plugging increase.
Several changes were made to containment parameters.
The thickness of one of the heat sinks was changed to better represent the as-built plant containment.
J Also, the previous value for the high-containment pressure setpoint was lowered t
to 18.5 pounds per square inch absolute to agree with the present value specified in the NA-1&2 TS.
The model calculations were performed assuming conservative generic 17 x 17 fuel parameters consistent with currently approved NRC methodology.
The previously
- Also, required 65 degree F uncertainty in fuel pellet temperature was removed.
a previous requirement of analysis using a spectrum of fuel heatup rates has been removed, which conforms with current NRC methodology for ECCS analysis.
When the above input changes were incorporated in the reanalysis, the assumed heat flux hot channel factor increased from 2.10 to 7.20.
A value of 2.20 The still ensures compliance with the 10 CFR Part 50.46 acceptance criteria.
increase from 2.10 to 2.20 is allowable from the NRC approved changes in the generic fuel parameters, the elimination of the fuel heatup rate spectrum, and the higher peak clad temperature resulting from the reanalysis.
Finally, an adjustment penalty of minus 0.06 must be applied to the overall heat flux hot channel factor F equals 2.20 which results in an adjusted heat flux hot channel q
factor F equals 2.14.
g The non-LOCA accidents and transients addressed in Chapter 15 of the NA-1&2 FSAR are affected in a variety of ways by increased steam generator tube plugging. The excess heat removal accidents tend to be slightly less severe because of the impaired heat transfer brought about by the two (2) percent in-crease in steam generator tube plugging.
Other accidents, such as overpressur-ization events remain essentially the same. The licensee's review of non-LOCA accidents has concentrated on those events to be judged adver:ely affected by the two (2) percent increase in steam generator plugging.
Fuel Pellet Stored Energy For LOCA analysis, Westinghouse methodology requires input be initialized with various steady state fuel parameters, one of which is a volumetric-average fuel To account for modeling uncertainties not explicity considered temperature.
elsewhere, a 65 degree F increase in temperature had previously been applied to l
l the steady state fuel performance calculated value.
The licensee has deleted this uncertainty from the present LOCA reanalysis for 7 percent steam generator plugging.
. We have previously approved the deletion of this uncertainty for this stored energy conservatism in our review of WCAP-8720, " Improved Analytical Models Used in West-inghouse Fuel Rod Design Computations," dated March 27, 1980.
Therefore, we find removal of the 65 degree uncertainty in the fuel pellet temperature for the licen-see's reanalysis to be acceptable.
Supplemental ECCS Analysis:
We have been generically evaluating three cladding material models that are used in ECCS evaluations.
These models predict cladding rupture temperature, cladding burst strain, and fuel assembly flow blockage.
We have discussed our evaluation of these medels with vendors and other industry representatives in our " Summary Minutes of Meeting on Cladding Rupture Temperature, Cladding Strain, and Assembly Flow Blockage," dated November 20, 1979, and in our published NRC report NUREG-0630, " Cladding Swelling and Rupture Models for LOCA Analysis," dated April 1980, wherein we concluded that licensing cladding models were, in some areas, non-conservative. Our letter dated November 9,1979 required licensees to confirm that operating reactors would continue to be in conformance with 10 CFR Part 50.46 when substituting NUREG-0630 cladding material models in presently approved ECCS evaluations.
In our letter to Westinghouse dated December 1,1981, we stated the completion of our generic review and approval of new acceptance criteria for Westinghouse clad-ding models.
For licensees using old Westinghouse ECCS evaluation models, the ECCS analyses should be accompanied by supplemental calculations using the clad-ding material models specified in NUREG-0630.
The licensee has referenced the old Westinghouse ECCS evaluation and has provided the supplemental ECCS calculations specified in NUREG-0630 for its reanalysis with seven (7) percent steam generator tube plugging.
The licensee's reanalysis also accounted for a non-conservatism identified by Westinghouse in its February 1978 ECCS evaluation model which used a fast-heatup-rate rupture-temperature correlation for slow transient analysis.
Based on a heat flux hot channel factor of F0 =2.20, the licensee's reanalysis for seven (7) percent steam generator tube plugging assessed the ' combined impact of the fuel-heatup-rates and the NUREG-0630 models to be worth 8550F peak cladding temperature above that previously calculated.
Subsequently, Westinghouse calculated that a reduction in the total peaking factor Fg.of 0.21 would offset the portion of the 855 degree F increase in peak cladding temperature that exceeded 2200 degrees F.
However, Westinghouse also identified a margin in Fg available through the use of upper-head-injection thermohydraulic models that we have generically approved for the NA-1&2 type of three-loop plant This margin is worth 0.15 in F 0 Therefore, a F0 reduction of minus 0.06 (0.15-0.21) is required and an overaT1 heat flux channer factor F is determined to be applicable for NA-1&2.
Basedontheabhve,we.14(2.20-0.06) of 2 conclude that the licensee's reanalysis has adequately addressed our concerns related to the clad swelling and rupture issue.
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. Non-LOCA Accidents and Trans'ients:
The licensee has evaluated non-LOCA accidents and transients adversely impacted by increased steam generator tube plugging.
The licensee has determined that steam generator tube plugging up tc, seven (7) percent would not reduce the primary system flow below the thermal design limit.
Therefore, analysis of departure from nucleate boiling (DNB) events, such as rod withdcawal at power, would not be affected.
Tube plugging affects pump coastdown characteristics and could adversely affect loss-of-flow accidents. The licensee has evaluated this matter, and the change in loop resistance is so small that the impact is negligible.
Boron dilution events could be affected by the reduced volume.
However, the two (2) percent reduction is not considered significant since more than an hour is still available for diagnosis and correction of such an event.
Evaluation:
Based on our review of the above matters, we conclude that the results of the ECCS-equal to 2.14 meets the criteria of CFR Part 50.46 and the LOCA analysis with a F0 In addition, we analysis was performed in accordance with 10 CFR 50 Appendix K.
have reviewed the licensee's evaluation of non-LOCA transients that might be affected by tube plugging, and we find that these transients are not adversely affected by a steam generator tube plugging increase from five (5) to seven (7) percent. Also we have determined that the licensee has adequately addressed our concerns regarding the cladding material models addressed in NUREG-0630. We therefore conclude that the proposed technical spec fication changes for NA-182 are acceptable for seven (7)%
steam generator tube plugg' g.
Environmental Consideration We have determined that the amendments do not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendments involve an action which is insignificant from the standpoint 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 these amendments.
, Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendments do not involve a significant increase in the probability or consequences of accidents previously considered and do not involve a significant decrease in a safety margin, the amendments do 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 these amendnents will not be inimical to the common defense and security or to the health and safety of the public.
Date: April 13, 1982 Principal Contributors:
N. Lattben
". Pov ers
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