ML19276E054
| ML19276E054 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 01/31/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19276E053 | List: |
| References | |
| NUDOCS 7903020245 | |
| Download: ML19276E054 (9) | |
Text
4 UNITED STATES
[4
'4 NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20566 t
%, * *... /
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 50 TO FACILITY OPERATING LICENSE NO. DPR-28 VERMONT YANKEE NUCLEAR POWER CORPORATION VERMONT YANKEE NUCLEAR POWER STATION DOCKET NO. 50-271 Introduction _
In conjunction with the Short Tenn Program (STP) evaluation of Boiling Water Reactor facilities with the Mark I containment system, the Vermont Yankee Nuclear Power Corporation submitted a Plant Unique Analysis (PUA) for the Vermont Yankee Nuclear Station. This analysis was performed to confirm the structural and functional capability of tiie containment suppression chamber and attached piping to withstand newly-identified suppression pool hydrodynamic loading conditions which had not been explicitly considered in the original design analysis for the plant. As part of the STP evaluation, specific loading conditions were developed for each Mark I facility, to account for the change in the magnitude of the loads due to plant-specific variations from the reference plant design for which the basic loading conditions were developed.
The results of the NRC staff's review of the hydrodynamic load definition techniques and the Mark I containment plant unique analyses are described in the " Mark I Containment Short Term Program Safety Evaluation Report",
NUREG-0408, December 1977. As discussed in this report, the NRC staff has concluded that each Mark I containment system would maintain its integrity and functional capability in the unlikely event of a design basis loss-of-coolant accident (LOCA) and, therefore, that licensed Mark I BWR facilities can continue to operate safely, without undue risk to the health and safety of the public, during an interim period of approximately two years, while a methodical, comprehensive Long Term Program is conducted.
79030202 T
. As discussed in Section III.C of NUREG-0408, of all of the plant parameters that were considered in the development of the hydrodynamic loads for the STP, only two parameters are expected to vary during nomal plant operation; these are (1) the drywell-wetwell differential pressure; and (2) the suppression chamber (torus) water level. Subsequent to the submittal of the PUA, the licensee was requested to submit proposed Technical Specifications which assure that the allowable range of these two parameters during facility operation would be it, accordance with the values utilized in the PUA.
The licensee has been operating this facility with differential pressure control to enhance the safety margins of the containment structure since early 1976 as a result of an Order issued by the NRC staff on February 13, 1976. This evaluation provides a more detailed basis for establishing the allowable range of drywell-wetwell differential pressure and torus water level, in order to quantify containment safety margins.
This amendment incorporates these parameters into the Technical Spec-ifications with the associatei limiting conditions for operation and surveillance requirements. Because these requirements replace the requirements of the February 13, 1976 Order, the Order is vacated.
By letters dated December 10, 1976, April 14, 1977 and May 16, 1978, the licensee proposed changes to the facility Technical Specifications to incorporate limiting conditions for operation and surveillcnce requirements for differential pressure control and torus water level.
Our evaluation of these proposed changes follows.
Evaluation The licensee has proposed certain Technical Specification requirements for the purpose of assuring that the nonnal plant operating conditions are within the envelope of conditions considered in their PUA. These Technical Specification changes establish (1) limiting condition for operation (LCOs) for drywell to torus differential pressure and torus water level, and (2) associated surveillance requirements. All other initial conditions utilized in the PUA are either presently included in the Technical Specifications or are configurational conditions which have been confirmed by the licensee and will not change during normal operation.
. Differential pressure between the drywell and the suppression chamber will result in leakage of the drywell atmosphere to the lower pressure regions of the reactor building and to the torus airspace. This leakage from the drywell will cause a slow decay in the differential pressure.
Therefore, surveillance requirements for the differential pressure have been included in the Technical Specifications. Surveillance frequency of once per operating shift for the differential pressure was selected on the basis of previous operating experience.
The torus water level is not expected to vary significantly during normal operation, unless certain systems connected to the suppression pool are activated. The torus water level will be monitored whenever such systems are in use. Therefore, we find that inclusion of periodic torus water level surveillance requirements in the Technical Specifications is not required.
We have reviewed the differential pressure and torus water level monitor-ing instrumentation systems proposed by the licensee with regard to the number of available channels and the instrumentation accuracy. This type of instrumentation is typically calibrated at six month intervals.
To assure proper operation during such intervals, two monitoring channels for both differential pressure and torus water level have been provided, such that a comparison of the readings will indicate when one of the channels is inoperative or drifting. The errors in the instrumentation are sufficiently small relative to the magnitude of the measurement (i.e., a maximum differential pressure measurement error of 0.1 psid in a measurement of 1.0 to 2.0 psid and a maximum torus water level measurement error of 10% of the difference between the maximum and minimum torus water level) that they may be neglected, based on the expected load variation with differential pressure and torus water level.
There are certain periods during nonnal plant operations when the differential pressure control cannot be maintained. Therefore, provisions have been included in the Technical Specifications to relax the differential pressure / control requirements during specified periods.
The justification for relaxing the differential pressure control during these specific periods and the basis for selecting the duration of the periods are discussed in detail below.
1.
Startup and Shutdown During plant startup and shutdown, the drywell atmosphere under-goes significant barometric changes due to the variation in heat loads from the primary and auxiliary systems.
In order to keep the periods during which the differential pressure control is not fully effective as short as is reasonable, we have limited the relaxation
- of the differential pressure control requirements for the startup and shutdown periods to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following startup and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to a shutdown. The postulated design basis accident for the containment assumes that the primary system is at operating pressure and temperature.
During the startup and shutdown transients, the primary system is at operating pressure and temperature 'or only a part of the transient, during which the differential pressure is being established. These time periods have' been shown by previous operating axperience to be adequate with respect to the startup and shutdown transients, and at the same time sufficiently small in conparison to the duration of the average power run.
Since the principal accident event to which differential pressure control is important to assure containment integrity (i.e., with a factor of safety of two) is a large break LOCA, we have considered whether there is a significantly greater probability of a large break LOCA during the startup and shutdown transients. We have concluded that there is not.
Further, the operation of the plant systems is monitored more closely than normal during these periods and a finite magnitude of differential pressure will be available during the majority of these periods to mitigate the potential consequences of an accident.
2.
Testing and Maintenance During normal operation, there are a number of tests which are required to be conducted to demonstrate the continued functional performance of engineered safety features. The testing of certain systems will require, or result in, a reduction in the drywell-torus differential pressure. The operability testing of the drywell-torus vacuum breakers requires the removal of the differential pressure to permit the vacuum breakers to open.
Because of the configuration at Vermont Yankee, testing of the suppression chamber-reactor building vacuum breakers and the standby gas treatment system may result in a reduction of drywell-torus differential pressure. For testing of high-energy systems (e.g., high pressure coolant injection pumps) during normal operation, the discharge flow is routed to the suppression pool. This energy deposition will raise the temperature of the suppression pool, resulting in an increase in torus pressure and a reduction in the differential pressure.
_ Functional performance testing of engineered safety features is necessary to assure proper maintenance of these systems through-out the life of the plant. Some of these tests (i.e., pump operability and drywell-wetwell vacuum breakers) may require or result in a reduction in the differential pressure. We estimate that not more than four tests will be required each month which will result in a reduction in differential pressure.
In order to keep the periods during which the differential pressure control is not fully effective as short as is reasonable, we have permitted a relaxation of differential pressure control in order to conduct these tests, limited to a period of up to four hours. Again, we have carefully considered whether the probability of a large LOCA is significantly greater during these testing periods than that during normal operation. We conclude that it is not. Moreover, only the test of the drywell-wetwell vacuum breakers requires complete removal of the differential pressure.
Provisions have also been included in the Technical Specifications for performing maintenance activities on the differential pressure control system and for resolving operational difficulties which may result in an inadvertent reduction in the differential pressure for a short period of time.
In certain circumstances, corrective action can be taken without having to attain a cold shutdown condition. To avoid repeated and unnecesssary partial cooldown cycles, a restora-tion period has been incorporated into the action requirements of tne LC0 for differential pressure control,i.e., in the event that the differential pressure cannot be restored in six hours, an orderly shutdown shall be initiated and the reactor shall be in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The six hour restoration period was selected on the basis that it represents an adequate minimum period of time during which any short-tenn malfunctions could be corrected, coupled with the minimum period of time required to conduct a controlled shutdown. The allowable time to conduct a controlled shutdown has been minimized, because the containment transient response is more a function of the primary system pressure than the reactor power level. On this basis, we find the proposed restoration period and action requirement acceptable.
We conclude that the limits imposed on the periods of time during which operation is permitted without the differential pressure control fully effective provides adequate assurance of overall containment integrity, and the periods of time differential pressure control is completely removed are acceptably small.
. The torus pumpback system will exhaust about 1 cfm continuously from the torus to the Reactor Building vent to maintain the required torus-drywell pressure difference. The licensee predicted the potential exclusion area boundary dose resulting from pumpback venting prior to containment isolation after a design basis LOCA would increase only a fraction of a rem (the resulting dose would remain well within the 10 CFR 100 guidelines).
Releases from accidents inside containment will be limited by contain-ment isolation, which also isolates the torus pumpback vent lines.
In addition, the torus air removed by the pumpback system is released through the Reactor Building vent, which is monitored by radiation detectors and closed on a high radiation signal from those detectors.
We have performed an independent analysis of the effect of the proposed torus pumpback system on the potential dose consequences of a LOCA.
Our assumptions for that analysis and the resulting potential dose consequences are presented in Tables 1 and 2.
We presented potential dose consequences of 148 Rem to the thyroid and 5.7 Rem to the whole body for the LOCA in the Safety Evaluation for Vermont Yankee dated June 1971.
The potential dose consequences presented in Table 2 represent an insignificant increase in the consequences presented above, and the resulting potential dose consequences are still well within the guide-lines of 10 CFR Part 100.
Routine flow from the torus will release some radioactivity to the environment because torus air contains some radioactivity after ce/tain reactor operations, such as a safety-relief valve lift.
The licensee has estimated that during normal operation, including antici-pated operational occurrences, 625 microcuries of I-131 per year will be released from the torus pumpback system.
The licensee based this estimate on measurements of the torus airborne radioactivity.
Based on cost-benefit analyses, the licensee found that the cost of installing and maintaining a charcoal filter in the torus vent line would exceed the benefit to the population from reduced radioiodine exposure.
We have reviewed the bases and methods the licensee used to estimate the routine radioactivity releases from the torus pumpback system.
Based on that review and on our knowledge of effluents from boiling water reactors, we conclude that the licensee's estimate of 625 microcuries of I-131 per year is reasonable. We have also reviewed the licensee's cost-benefit analysis for adding a charcoal filter to
7-the torus vent line. We have performed an independent analysis using more conservative assumptions, and we agree with the licensee's conclusion that this would not be cost-beneficial to the public.
In addition, the release limits given in the Technical Specifications fce the Vermont Yankee station are not being changed by this action; hence, the releases from torus-pumpback system operation, when added to the other normal plant releases, must remain within current effluent limits.
We are currently evaluating Vermont Yankee, as well as all other operating light water reactors, to determine compliance of Vermont Yankee with the requirements of 10 CFR Part 50, Appendix I.
At the conclusion of that evaluation, the Technical Specifications will be amended as necessary to implement Appendix I for the operation of Vermont Yankee.
The releases of radioactivity from torus pumpback operation will be included in the evaluation and the implementing changes to the Technical Specifications.
On November 29, 1978, we requested information from the licensee related to purging during operation. The information that the licensee provides in response to this request will include consideration of the torus pumpback vent lines. We will review this information generically according to the guidelines described in our letter of November 29, 1978.
We believe the licensee has adequately considered the potential impact on the public of accidental and normal releases from the torus pumpback system venting and believe torus pumpback system operation will not pose an undue risk to the public health and safety.
Environmental Consideration We have detennined 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 impact. Having made this determination, we h:ve further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR Section 51.5(d)(4), that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
Conclusions The proposed Technical Specifications will provide the necessary assurance that the plant's operating conditions remain within the envelope of the conditions assumed in the Plant Unique Analysis (PUA) performed in conjunction with the Mark I Containment Short Term Program. The PUA supplements the facility's Final Safety Analysis Report (FSAR) in
~.
. that it demonstrates the plant's capability to withstand the suppression pool hydrodynamic loads which were not explicitly considered in the FSAR. We therefore conclude that the proposed changes to the Technical Specifications cre acceptable.
We further conclude, 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 Comission's regulations and the issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the p.blic.
Dated:
January 31, 1979
TABLE 1 ASSUMPTIONS FOR AND POTENTIAL CONSEQUENCES AT THE EXCLUSION AREA BOUNDARY OF LOCA CONTRIBUTION BY TORUS PUMPBACK SYSTEM FOR VERMONT YANKEE Assumptiens Value Valve Isolation Time 10.5 seconds Primary Coolant Iodine Activity, 1.1 microcuries per gram Dose Equivalent I-131 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> X/Q Value, Exclusion 4.0 x 10-4 seconds per Area Boundary (94 meter stack release) cubic meter Primary Coolant Released from Torus 350 pounds Pumpback System Prior to Isolation
~.
TABLE 2 EXCLUSION AREA BOUNDARY (EAB) CONSEQUENCES FROM ACCIDENT (CONTRIBUTION FROM TORUS PUMPBACK SYSTEM)
Doses, Rem Thyroid Whole Body s<l.0
,< 0.1