ML20211N826
| ML20211N826 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 02/18/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20211N805 | List: |
| References | |
| NUDOCS 8703020231 | |
| Download: ML20211N826 (6) | |
Text
p 4'
UNITED STATES NUCLEAR REGULATORY COMMISSION E
.E WASHINGTON. D. C. 20665
%*...j SAFETY EVALLATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 14 TO FACILITY OPERATING LICENSE NO. NPF-38 LOUISIANA POWER AND LIGHT COMPANY WATERFORD STEAM ELECTRIC STATION, UNIT 3 DOCKET NO. 50-382
- 1. 0 INTRODUCTION By applications dated June 24, July 15, and August 29, 1986, as supplemented by letters dated October 3, November 3, and November 12, 1986, Louisiana Power and Light Company (the licensee or LPL) requested changes to the Technical Specifications (Appendix A to Facility Operating License No. NPF-38) for the Waterford Steam Electric Station, Unit 3.
The proposed changes would:
(1) bypass the non-safety related high steam generator level trip; (2) add the Reactor Vessel Level Monitoring System; and (3) change the location of the seismic monitors inside containment.
2.0 DISCUSSION The proposed changes to the technical specifications requested by the licensee are in three areas, as described below.
2.1 Steam Generator Level Hi Trip (NPF-38-23)
The proposed change would revise Table 3.3-1 of the Waterford 3 Technical Spacifications to include the capability of manually bypassing the high steam generator level trip during Modes 1 and 2 at power levels of 20% full power or less.
The revision would be made by adding footnote (g) to the Table on Reactor Protective Instrumentation.
2.2 Reactor Vessel Level Monitoring System (NPF-38-28)
The proposed change would revise Technical Specification 3.3.3.6, " Accident Monitoring Instrumentation", to add the Reactor Vessel Level Monitoring System (RVLMS) to Tables 3.3-10 and 4.3-7, and the associated Bases.
0703020231 ET70210 ft.ADOCK0500gG
e.
2.3 Seismic Monitors (NPF-38-37)
The proposed change would reflect the relocation of seismic monitor YR-SM 6020 and correct an error in the location of another seismic monitor, YR-SM 6021.
YR-SM 6020 and YR-SM 6021 are passive devices that allow for evaluation of reactor coolant system response after a seismic event.
During the first cycle of operation of Waterford 3, YR-SM 6020 suffered heat damage. The licensee proposes to relocate this monitor during the first refueling outage, moving it from the high temperature environment at its present location on the pressurizer to Safety Injection Tank (SIT) 18. The proposed relocation uses a mount virtually identical to the original mount and which performs the same function.
3.0 EVALUATION The proposed changes to the Technical Specifications requested by the licensee and described in three areas above, are evaluated below.
3.1 Steam Generator Level Hi Trio (NPF-38-23)
In Waterford 3, a high steam generator water level trip is provided to trip the reactor when measured steam generator water level rises to a high preset value, ncminally 87.7 percent of the distance between the lower and up)er instrument nozzles. Since the turbine is automatically tripped when t1e reactor is tripped, the high steam generator water level trip is provided to protect the turbine from excessive moisture carryover. The trip is an equip-ment protective trip only and, thus, is non-safety related. No credit was taken for operation of this trip in the plant safety analyses nor does the trip setpoint correspond to a Technical Specification safety limit. Likewise, the design and reliability of the reactor protection system is unaffected by the proposed change.
The high steam generator water level trip bypass will be administrative 1y controlled such that the bypass cannot be enabled above 20% power.
For power levels at or below 20%, existing automatic and manual controls assure that the main steam piping does not become water filled.
Since the potential does exist for the initial steam generator water level to be as high as 80.5% (narrow range), LPL has analyzed the potential impact of the increased steam generator inventory on the radiological consequences of an inadvertent opening of a steam generator atmospheric dump valve. This event was previously analyzed for Cycle 1 at zero power and with an initial steam generator water level just below the high steam generator water level trip setpoint (87.7%) in order to maximize secondary side water inventory and
e' A.,
the radiological consequences.
The reanalysis confirms that the total dose remains well,within the guidelines of 10 CFR Part 100.
The consequences of a steam line break event with a higher steam generator inventory could be more severe than the consequences with a lower inventory, particularly with regard to peak containment temperature and pressure.
LPL analyzed steam line break events for Cycle 2 occurring at full power initial conditions, with a loss of off-site power coincident with the reactor trip and an initial steam generator water level governed by the high water level trip setpoint (87.7%).
These results of the analysis indicate that the highest containment pressure and temperature occur within 60 seconds following a break in the main steam line and blowdown of the affected steam generator ends at 200 seconds.
The peak containment pressure and temperature, therefore, occur well before the end of steam generator blowdown.
Thus, the additional inventory that may be present in the steam generator at 20% power or less will not change the maximum containment pressure or temperature presented in the FSAR.
Based on the above, the staff finds that the capability of manually bypassing the high steam generator level trip at power levels of 20% or less is not a safety concern and is acceptable.
3.2 Reactor Vessel Level Monitorina System (NPF-38-28)
In response to NRC Generic Letter No. 83-37, "NUREG-0737 Technical Specifications," dated November 1, 1983, the Combustion Engineering (CE)
Owners Group proposed a generic Technical Specification for the RVLMS in their letter from R. W. Wells to H. L. Thompson (NRC), dated February 19, 1985 (RWW-85-12).
The proposed Technical Specification was reviewed by the NRC staff and it was found to be acceptable for application to CE designed reactors such as Waterford 3 which uses the heated junction thermocouple system (HJTCS).
For non-System 80 CE plants, a channel (eight sensors in a probe) is defined to be operable if at least four of its eight sensors are operable (one or more of the upper three [ upper head] and three or more of the lower five
[ plenum region]).
The accepted Action Statements permit seven days to repair a failed channel if repairs can be made without shutting down.
If repairs are not feasible during operation, a special report must be filed with the NRC within 30 days.
If less than one channel is operable, then 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> are permitted to restore operability or an alternate method of monitoring reactor vessel inventory must be initiated, a special report filed with the NRC within 30 days, and operability restored at the next refueling shutdown.
'N The proposed change to the Waterford Technical Specification 3.3.3.6 in-corporates the above provisions which have been approved previously by the staff.
In addition, since the specific purpose of the proposed change is to enhance accident and transient monitoring capability, the staff finds the proposal acceptable.
3.3 Seismic Monitors (NPF-38-37)
The requirements for seismic monitors in the containment are given in 10 CFR Part 100, Appendix A, Section VI(a)(3): " Required Seismic Instrumentation.
Suitable instrumentation shall be provided so that the seismic response of nuclear power plant features important to safety can be determined promptly to permit cornarison of such response with that used as the design basis. Such a comparison is needed to decide whether the plant can continue to be operated safely and to permit such timely action as may be appropriate."
The proposed relocation of YR-SM 6020 is due to damage induced by high tempera-tures experienced by the monitor during the first cycle of operation. On the pressurizer, temperatures exceeded the 555*F rating of the thermal barrier installed with the monitor, resulting in damage to the monitor. For the new monitor location, the SIT maximum design temperature is 200 F - well below the thermal barrier rating. The expected SIT temperature will approximate the containment ambient temperature of 120'F, well below the design rating for the seismic monitor. Other environmental stresses, i.e., acceleration, nuclear irradiation, relative humidity, and chemical exposure, expected at the new location in the containment will be less than the values of these stresses for which the monitor was designed. Thus, the monitor is expected to be available to perform its intended function in the unlikely event of an earth-quake.
Because no safety-related equipment is located below the YR-SM 6020 monitor, and there are no electric, hydraulic or pneumatic lines of any kind from YR-SM 6020 to any other system, adverse systems interactions are not a factor in the proposed relocation.
The staff concludes that the proposed relocation of seismic monitor YR-SM 6020 to the lower liftina lug of safety injection tank 1B is acceptable.
The location of seismic monitor YR-SM 6021 is given incorrectly in the Tc.chnical Specifications in Tables 3.3-7 and 4.3-4.
These tables are revised per this proposed amendment to correctly describe the seismic monitor location.
The staff concludes that Tables 3.3-7 and 4.3-4 of the Technical Specifications l
for Waterford 3 should be corrected to reflect the actual location of YR-SM 6021.
s 4
- 1 4.0 CONTACT WITH STATE OFFICIAL The NRC staff has advised the Administrator, Nuclear Energy Division, Office of Environmental Affairs, State of Louisiana of the proposed determination of no significant hazards consideration. No comments were received.
5.0 ENVIRONMENTAL CONSIDERATION
This amendment involves changes in the installation or use of facility components located within the restricted area. The staff has determined that the amendment involves no significant increase in the amounts of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued proposed findings that the amendment involves no significant hazards consideration, and there has been no public comment on such findings. Accordingly, forth in 10 CFR 51.22(c)(9)gibility criteria for categorical exclusion setPursuant to 10 C the amendment meets the eli impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
CONCLUSIONS Based upon our evaluation of the proposed changes to the Waterford 3 Technical Specifications, we have concluded that:
there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed menner, and such activities will be conducted in ecmpliance with the Commission's regulations and the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
We, therefore, conclude that the proposed changes are acceptable, and are hereby incorporated into the Waterford 3 Technical Specifications.
Principal Contributors:
L. Kopp, C. Morris Deted:
February 18, 1987
February 18, 1987 ISSUANCE OF AMENDMENT N0. 14 TO FACILITY OPERATING LICENSE NP. NPF-38 FOR WATERFORD 3 DISTRIBUTION Docket File 50-382 /
NRC PDR Local PDR P8D7 Reading FMiraglia JLee (5)
JWilson Attorney, OGC - Bethesda LHarmon EJordan BGrimes JPartlow T8arnhart (4)
WJones WRegan ACRS (10)
OPA RDiggs, LFMB DCrutchfield CThomas LKopp NLauben WRegan JCalvo Wermeil CMorris l
l l
l l