ML20197C569
| ML20197C569 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 10/27/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20197C555 | List: |
| References | |
| NUDOCS 8611060261 | |
| Download: ML20197C569 (5) | |
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UNITED STATES 8 ' )^g%#( g
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NUCLEAR REGULATORY COMMISSION o
w.....j,E WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO.109 TO PROVISIONAL OPERATING LICENSE NO. DPR-16 i
GPU NUCLEAR CORPORATION AND JERSEY CENTRAL POWER & LIGHT COMPANY OYSTER CREEK NUCLEAR GENERATING STATION DOCKET NO. 50-219
1.0 INTRODUCTION
By letter dated September 11, 1986, GPU Nucleat (the licensee) requested an amendment to Provisional Operating Licentc No. DPR-16 for the Oyster Creek Nuclear Generating Station (0yster Creek). This amendment would authorize two changes to Section 4.4, Emergency Cooling, of the Appendix A Technical Specification (TS), which lists the surveillance requirements and the frequency of surveillance for the reactor emergency cooling systems. This amendment would change (1) the stated frequency and pressure conditions fcr the Automatic Depressurization System (ADS) valve opera-bility test in Item 4.4.B.1 of Section 4.4 to after each refueling outage and at system operating pressure prior to exceeding 5 percent power and (2) revise the Bases for Section 4.4.
This change was to clarify the surveillance requirements of ADS valve operability in the TS.
2.0 DISCUSSION The licensee has proposed Technical Specification Change Request (TSCR)
No. 140 to clarify the frequency and pressure conditions for testing the ADS valve operability required in the TS. The existing TS 4.4.B.1 on ADS valve operability is confusing. The TS refer to " low pressure" for the tests in the Bases of Section 4.4 but the pressure is not defined and refer to a test every refueling outage but the tests are run as the plant is going from the Refueling Mode into the Run Mode as the plant restarts from the refueling outage. The proposed words clearly state when and at what pressure conditions these tests are conducted.
The ADS consists of five automatically (or manually activated electromatic relief valves (EMRVs). The ADS is to 1) depressurize the reactor coolant system (RCS) during a small break LOCA to permit the low pressure core spray system to inject water into the core and (2) provide overpressure protection for anticipated plant transients. The ADS is automatically actuated by high drywell pressure and low-low-low reactor water level.
These also are indications of a large break LOCA; however, the large break LOCA will depressurize the RCS by itself and the ADS is not needed.
There are three EMRVs on one steam line and E EMRVs on the other steam line from the reactor vessel. The position of the EMRVs is shown in the 8611060261 861027 DR ADOCK 05000219 PDR
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. attached figure from the Oyster Creek Updated Final Safety Analysis Report. The EMRVs blow down to the torus suppression pool in the primary containment and not to the drywell.
Specification 3.4.B.1 states that the EMRVs shall be operable when the reactor water temperature is greater than 212 F and pressurized above 110 psig. The existing specifications could be interpreted to not allow EMRV testing at any steam pressure (steam does not exist below 212 F) and at steam pressures representative of those at which the EMRVs would operate. The EMRVs need a steam pressure above 50 psig to open. Testing the valves at representative operating conditions where they would be expected to operate provides the best assurance that these valves will operate satisfactorily if called upon to depressurize the RCS.
3.0 EVALUATION The licensee has proposed TSCR 140 to clearly state that the EMRVs may be demonstrated operable at RCS operating pressures prior to exceeding 5 percent power.
In addition, in order to remove a source of confusion from the TS the reference to low pressure testing of the EMRVs is proposed to be eliminated from the basis section for Section 4.4 of the TS.
The EMRVs are tested for operability as the plant comes out of every refueling outage when there is essentially no decay heat.
In the event of a leak or rupture coincident with the test and the failure of all five EMRVs, the Isolation Condensers can depressurize the RCS since there would be little stored energy or decay heat in the fuel. The depressurization capability of the Isolation Condensers is sufficient for testing the ADS following a refueling outage and as necessary during the operating cycle.
The proposed restriction that valve operability shall be demonstrated prior to exceeding 5 percent power adds a restriction to this surveillance requirement that is not in this existing TS.
The ADS is designed to depressurize the RCS during a small break LOCA to permit the low pressure core spray system to inject water into the core.
Testing the EMRVs at system pressure represents normal operating parameters and does not expose the plant to conditions beyond which it is designed to operate. All testing of the EMRVs at Oyster Creek has been at these pressures.
Because of its design, the EMRV cannot be tested below RCS steam pressures of 50 psig. The pressure is, by design, on both the front of the main valve disc which acts to open the valve and on the back of the disc, to close the valve. The unbalanced force to keep the valve closed is the enclosed spring (50 psig). To open the valve, an electrical signal to the solenoid assembly opens a pilot valve to bleed steam off the back of the disc and the higher RCS pressure on the front will open the valve. To close the valve, an electrical signal to the same solenoid assembly closes the pilot valve and st sam pressure builds up on the back of the valve disc equal to that on the front and the spring closes the valve. The valve manufacturer recommends testing the valve at the reactor operating pressures for which the valve has been designed.
. During the test of each EMRV, the RCS pressure could, if not properly controlled, drop in the RCS because the open EMRV is an open hole on the RCS. A significant drop in pressure would cause voiding and reactivity transients in the RCS which are not desired. The RCS pressure is con-trolled by the turbine pressure regulator controls.
The pressure regulator controls prevent unnecessary rapid depressurization of the reactor coolant system during the test. Either the mechanical or electrical pressure regulator controls reactor pressure during reactor startup, operation, and shutdown. The mechanical pressure regulator is used during reactor startup and shutdown and the electrical pressure regulator is used at reactor pressures in excess of about 980 psig. The mechanical pressure regulator response to changing pressure conditions is significantly slower than that of the electrical pressure regulator.
Therefore, the test at reactor operating pressures which are greater than 980 psig will have the better pressure control regulator (Ref. 2 and 3).
Based on the above, the staff concludes that the licensee's proposed changes to the TS in TSCR 140 are acceptable. The proposed changes to the Bases have been reviewed and found to be appropriate.
4.0 ENVIRONMENTAL CONSIDERATION
This amendment involves a change to a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes to the surveillance require-ments. The staff has determined that the amendment 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 cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement nor environ-mental assessment need be prepared in connection with the issuance of this amendment.
5.0 CONCLUSION
The staff has 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 this amendment will not be inimical to the comon defense and security nor to the health and safety of the public.
6.0 REFERENCES
1.
Letter from P. B. Fiedler (GPUN) to J. A. Zwolinski (NRC), TSCR No.
140, dated September 11, 1986.
+ 2.
Phone conference calls between M. Laggart and J. Kowalski (GPUN) and J. Donohew (NRC) on October 9, 1986.
3.
Oyster Creek Nuclear Generating Station, Updated Final Safety Analysis Report, Section 10.2, Turbine Generator.
Principal Contributor:
J. Donohew Dated:
October 27, 1986
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-hi GPU NUCLEAR CORPORATION OYSTER CREEK NUCLEAR GENERATING STATION REACTOR COOLANT SYSTEM BOUNDARIES UPDATED FINAL SAFETY ANALYSIS REPORT REV.O 12/S4 l
FIGURE 5.2-1 i
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