ML20002A865
| ML20002A865 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 11/03/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20002A860 | List: |
| References | |
| NUDOCS 8011210749 | |
| Download: ML20002A865 (5) | |
Text
...
o Sr Estg k
UNITED STATES i
NUCLEAR REGULATORY COMMISSION
. WA5rHNGTON, D. C. 20555 5
y
,8 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION l
SUPPORTING AENDMENT NO. 58 TO FACILITY OPERATING LICENSE NO. DPR-28 VERMONT YANKEE NUCLEAR POWER CORPORATION VERMONT YANKEE NUCLEAR POWER STATION DOCKET NO. 50-271 1.0 Introduction 7
By letters dated March 17,1980 an'd August 28, 1980, as supplemented May 9, August 13, September 23 and October 14,1980, Vermont Yankee.
i Nuclear Power Corporation (licensee) requested amendments to Facility Operating License No. DPR-28 for the Vermont Yankee Nuclear Power Station. The proposed amendments would revise the Technical Specifi-cations to authorize the licensee to:
A.
Replace existing pressure switches that sense reactor pressure and reactor water level with analog loops, B.
hstall an ATWS' recirculation pump trip (RPT) system of the Monticello design, to trip on low-low reactor water level or i
high reactor pressure, and C.
Require Type C testing of certain containment isolation valves to reflect changes brought about for post-accident hydrogen sampling resulting from NUREG-0578, "TMI Lessons Learned Task
. Force Status Report on Short Term Recomendations."
2.0 Evaluation 2.1 Analog Trip System The licensee has proposed certain modifications to Appendix A of the operating license for the Vermont Yankee Plant. These modifications involve installing a new design improvement'for safety system instru-mentation for General Electric Company (GE) boiling water reactors for the reactor protection system (RPS) and emergency core cooling system (ECCS). The proposed modification is referred to as the analog trip system.
1 son e2 o ?F
s
. This analog trip system is similar to that developed by GE and des-cribed in GE's Topical Report NED0-21617 of April 1977 and NED0-21617-1 of January 1978 entitled, " Analog Transmitter / Trip Unit System"(ATTUS).
GE submitted this topical report to the NRC staff for review and it was found acceptable by the staff as stated in the letter to GE dated June 27, 1978.
Since the licensee had not referenced this approved GE topical report, we requested the licensee to compare their analog trip system design to that described in GE's topical report. Our evaluation of the licensee's proposed design change (March 17, 1980) and responses to our inquiries (May 9,1980; and August 13, 1980 and September 23, 1980) is as follows:
The licensee identified the analog trip system to be a replacement of pressure and differential pressure switches which sense reactor pressure and reactor water level with analog channels each consisting of a transmitter, indicator and trip unit. This analog trip system is designed to increase plant reliability, reduce setpoint drift and improve safety of the plent. The equipment to be used includes Rosemount Model 1152 analog transmitters with "E" output codes and Rosemount Model 51000 trip units.
During our review of the it.ansee's proposed change and the results of the comparison of their design to GE's Topical Report NED0-21617, several items were identified as having possible safety implications.
The areas identified pertained to:
- 1) ATWS recirculation pump trip circuitry separation, 2) power supply, 3) arrangement of panel equip-ment, and 4) environmental qualification.
- 1) The ATWS recirculation pump trip circuitry is to be included within the ECCS cabinets for the analog trip system. The NRC staff was concerned with the adequacy of separation between the Class lE and non-class lE wiring for ATWS pump trip. After discussions with the licensee, we were informed that the ATWS recirculation pump trip circuitry will be a Class lE system and its wiring will be routed with the ECCS Class IE wiring.
Therefore, we consider this concern resol ved.
2)
The GE topical report states that DC power supplies can be connected in parallel to obtain a higher output current. However, the VerTnont Yankee plant design is such that a single power supply is sufficient to handle its associated divisional / channel load. Hence, the licensee does not plan to use parallel power supplie:;.
Each division / channel will be provided with a separate and independent power supply.
Therefore, we find this design acceptable.
. 3) The location'of equiprent and wiring of the trip unit panel differ from that described in the GE topical. The trip units and associated trip relays are to be arranged differently and all field wiring will be routed through the top of the cabinet instead of the bottom.
Separation will be maintained between Class 1E and non-Class 1E wiring. Therefore, this proposed change in physical arrangement does not pose a concern and is acceptable.
- 4) The licensee states that analog trip system cabinets are to be located in the reactor building. According to the GE topical report, the power supplies and trip relays located within these cabinets are qualified for use in control room environments.
Discussions with the licensee resulted in a coninitment from the licensee assuring that all equipment installed in the reactor building will be seismically and environmentally qualified for normal plant operation and the worst-case accident reactor building envimnment.
We have reviewed the service environment for this equipment and conclude that it is a mild environment. The qualification of safety-related electrical equipment to function in environmental extrenes not associated with accident conditions is the responsibility of the licensee to evaluate and document in a form that will be available for the NRC to audit. Qualification to assure functioning in mild environments must be completed by June 30, 1982.
Based on our prior review and approval of GE's Topical Report NED0-21617 and our recent review of the licensee's submittals which included a comparison of Vermont Yankee's design to the GE topical, we conclude that the proposed modifications (analog trip system) meet the applicable provisions of IEEE 279-1971, GDC 13, and GDC 20 as described above.
2.2 ATWS Recirculation Pump Trip Installation of an RPT, is for the purpose of providing a partial backup system to the Reactor-Protection System (RPS).
In the unlikely event that the plant experienced an operational transient without a subsequent scram, the RPT would reduce core power generation by rapidly reducing core flow, thus mitigating the short term consequences of the transient. The basis for the system's design is the Monticello type RPT with a time delay on low-low reactor water level which is described in the letter fmm NRC to Vermont Yankee, dated January 8,1979.
Because of the increment of safety added by installation of an RP,T of such a design, on February 21,'1980 the NRC ordered that this installation be completed by December 31, 1980. We find this modifi-cation acceptable.
l
l 4-2.3 Containnent Isolation Valve Testing The licensee has requested certain changes in Type C containnent 1
valve testing in order to meet the requirements of NUREG-0578, "TMI Lessons Learned Task Force Status Report and Short-Term Recommendations,"
with respect to post-accident hydrogen sampling.
It should be noted that we are currently in the process of reviewing all generic issues pertaining to Appendix J leak rate testing for operatir.g nuclear power plants at the Franklin Institute.
The licensee has proposed that certain valves of the hydrogen monitoring system (part of the CAD system), be deleted from Table 4.7.2b of the Technical Specifications. The hydrogen monitoring system was designed and installed to' meet Seismic Category I requirements and as an integral part of containnent. The system, which does not require an isolation system, has valves 109-75A-D; 1 & 2 and 109-76 A & B.
These valves do not receive PCIS signals and are remotely operated from the main control room and have not previously been subject to Type C leakage tests. The licensee regards this system as an extension of primary containment not requiring a system isolation and remaining open in a post-accident condition. The containment hydrogen monitoring system would therefore operate continuously to satisfy the associated short term lessons learned requirenent.
l General Design Criterion 56 explicitly states that each line that connects to the containment atmosphere and penetrates primary reactor containment shall be provided with containment isolation valves unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other basis.
In this regard, the Standard Review Plan at Section 5.2.4, paragraph II.3.b permits the deviation from GDC 56 for lines that~ are part of engineered safety systems. These systems may have marual valves, but provisions should be made to detect possible leakage from these lines outside of-containment.
On this basis we are permitting the use of remote manual valves in the hydrogen monitoring svstem for those lines which penetrate the prinary containment, since they are an integral part of an engincered safety system. However, we require that for this system one valve in every line penetrating the primary containment be regarded as an isolation valve and be appropriately included in Table 4.7.2b.
With regard to
', these valves, we currently are reviewing at the Franklin Institute all requirements for Appendix J testing, which will address this system as well as other systens in the Vermont Yankee Plant. Therefoce, until this review is completed, we will defer judgment on requiring Type C
.esting for these isolation valves. However, the licensee has agreed that the hydrogen monitoring system will be subject to a Type A Inte-grated Leak Rate Test'.
1 1
-,w w--
The licensee has also proposed that solenoid valves VG-23, VB-26 and 109-76 A&B,' which were installed in the radiation monitor inlet and outlet lines to close.on receipt of a primary containment isolation a
~
signal and provide containment isolation, be :added to Table 4.7.2b and thus be subject to Type C_ leakage tests.
Appendix J,.10 CFR Part 50, requires that valves designed to operate subsequent.to a design basis accident, which may become a part of the containment isolation system barrier during post-accident operation, be subject to Type C leakage rate tests.
By letter dated October 15, 1980, the liter,see indicated that the isolation valves VG-23 and VG-26 are located as close as practical to the hydrogen monitoring system, which is considered an extension of containment and that valves 109-76 A&B are located as close as practical 1
to the torus wetwell. Based on the above considerations we find this change in containment isolation valve testing acceptable.
3.0 Environmental C_onsideration i
We have determined that the license amendment does not authorize a i
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
^
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 envimnmental impact statement or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of this amendment.
1 4.0 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 pmviously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, 4
(2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed i anner, and (3) such activities _will be conducted in compliance with the Commission's i
ragulations and the issuance of this amendirent will not be inimical to the common defense and security or to the health and safety of the public.
Dated: November 3,- 1980 2
4
,