ML18004B964

Notifies NRC of Intentions Re Conducting Generator Loss of Load Test from 100% Power.Util Has Adequately Demonstrated Compliance W/Reg Guide 1.68 & No Further Transient Testing Required

ML18004B964
Person / Time
Site:	Harris
Issue date:	10/12/1987
From:	Zimmerman S CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
RTR-REGGD-01.068, RTR-REGGD-1.068 NLS-87-226, NUDOCS 8710160251
Download: ML18004B964 (8)

Text

~U REGULA ( INFORMATION DISTR I BUT IOl YSTEM ( R IDS )

A ACCESSION NBR: 8710160251 DOC. DATE: 87/10/12 NOTARIZED: NO DOCKET FACIL: 50-400 Sheav on Harv is Nucleav Power Planti Unit ii Carolina 05000400 AUTH. NAME AUTHOR AFFILIATION ZIMMERMANiS. R. Cav'olina Power 5 Light Co.

RECIP. NAME RECIPIENT AFFILIATION Document Contv ol Bvanch (Document Control Desk)

SUBJECT:

Notifies NRC of intentions re conducting generatov loss of load test from 100/ poeer. Util has adequately demonstrated compliance w/Reg Guide i. 68 Zc no fuv thev tv ansient testing required.

DIBTRIBUTION CODE: A047D COP IEB RECEIVED: LTR J ENCL 0 SIZE:

TITLE: OR Submittal: Inservice Inspection/Testing NOTES: Application for permit renewal f i. led. 05000400 RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-1 LA PD2-1 PD BUCKLEY'S B INTERNAL: AEOD/DOA 1 AEOD/DSP /TP AB ARM/DAF/LFMB NRR/DEST/MEB NRR/DEST/MTB 1 NRR/ /ILRB OGC/HDS1 FILE 01 RES/DE/EIB 1 EXTERNAL: LPDR NRC PDR NSIC TOTAL NUMBER OF COPIES REQUIRED: LTTR 19 ENCL 0

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SRK0, Carolina Power & Light Company USNRC-DS 8810CI lq AtO:Oq OCT l p gg SERIAL: NLS-87-226 United States Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-000/LICENSE NO. NPF-63 PERFORMANCE OF GENERATOR LOSS OF LOAD TEST FROM 100 PERCENT POWER Gentlemen:

Carolina Power R Light Company (CPRL) hereby notifies the NRC of our intentions with regard to conducting a generator loss of load test from 100 percent power at the Shearon Harris Nuclear Power Plant (SHNPP). This test is identified in the SHNPP FSAR Section 10.2.12.2.18 and complies with Regulatory Guide 1.68, Appendix A, paragraph 5.nn. Based on the information presented below, CPdcL has adequately demonstrated compliance with this guidance, and no further transient testing is required.

Regulatory Guide 1.68, Appendix A, Section 5.nn specifies a power ascension test to "demonstrate that dynamic response of the plant is in accordance with design for the case of full load rejection." It also specifies opening of the generator output breakers via a method which will maximize turbine overspeed and to align the plant's electrical distribution system for normal full-power operation. It would appear that the primary consideration of this test is turbine overspeed. It is important to note that turbine overspeed was addressed by the ACRS for SHNPP. To resolve this issue, CPRL submitted its turbine overspeed test procedure to the NRC along with additional information via letters dated November 21, 1986 and December 5, 1986. This information was accepted by the NRC Staff and the ACRS to close out the turbine overspeed issue. The testing committed to in CPdcL's letters has been successfully completed; therefore, turbine overspeed is not a consideration for the generator loss of load test.

I The SHNPP was originally designed to accommodate a loss of load via turbine/generator and reactor runback to house load and remain ready for reconnection to the grid. Based on information from the NSSS vendor and the experience of other utilities who have tested this feature, it is not clear whether the plant will achieve this capability. It should be noted that this design feature is a non-nuclear safety control, and there are other safety-related plant systems designed to shut the plant down in the event of a loss .

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nited States Nuclear Regulatory Commission In its current configuration, the SHNPP is capable of a safe shutdown following opening of the generator output breakers; however, it is not expected to be able to run back to house loads. The expected response of the plant is that a reactor trip will occur on low-low steam generator level shortly after initiating a load rejection. The low steam generator level indication occurs because of an expected pressure spike occurring in the steam lines affecting the level transmitters. Since the turbine will trip on the reactor trip, and since the fast bus transfer from the main generator to offsite power is defeated when the test is initiated by the opening of the main generator output breakers, a loss of offsite power to the plant will occur. This will result in trip of all reactor coolant pumps and a startup of both diesel generators to re-energize the safety busses. The plant will stabilize in mode 3 in natural circulation. The capability of the plant to successfully respond in this manner has already been demonstrated during the power ascension test program. (Reference FSAR 10.2.12.2.21, Loss of Offsite Power Test Summary, FSAR 10.2.12.2.26, Natural Circulation Test Summary, FSAR 10.2.12.2.19, Turbine Trip From 100 Percent Power Test Summary). The integrated plant response, as described above, is nearly identical to a loss of offsite power, which was performed at the 10 percent power plateau. This expected plant response is such that the plant is well within analyzed conditions. Therefore, it is concluded that in the event of an inadvertent opening of the generator output breakers, the plant will shut down safely.

The likelihood that the Harris Plant would experience an actual event similar to the generator loss of load test is very low. The only events that lead to the scenario addressed by the loss of load test are a complete loss of offsite power or the opening of both generator output breakers without a generator lockout signal which is indicative of an adverse condition in the switchyard..

The Harris Plant offsite distribution network consists of six separate transmission lines radiating in multiple directions from the site, with any one line capable of supplying the safety related loads at the plant. FSAR Section 8.2 provides a complete description of the offsite power system including the design features that make the complete loss of offsite power a highly unlikely event. Section 3 of NUREG 1032, Evaluation of Station Blackout Accidents at Nuclear Power Plants, addresses loss of offsite power frequency and duration as applied to station blackout. Table 3.1 of NUREG 1032 concludes that from.a historical perspective, the frequency of total loss of offsite power at United States nuclear power plant sites is 0.088 per site-year. In addition, there are important, although less quantifiable, considerations which make this probability even lower. These considerations include:

~ NUREG 1032 uses historical data back to 1968. Utilitygrids have become more reliable over time as evidenced by Table 5.1 in NUREG/CR-3992; therefore, a historical frequency of occurrence is conservatively high compared to frequency of occurrence for a state-of-the-art design.

The number of offsite transmission lines serving the Harris Plant is higher than a typical plant.

The Harris Plant is located in a region of the country which is not as susceptible to hurricanes, tornados, or snow and ice storms as coastal, midwest, or northern locations.

The design of the Harris Plant switchyard as described in FSAR Section 8.2 is highly reliable and uses a complete breaker-and-a-half scheme with the unit connected in a double breaker scheme.

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United States Nuclear Regulatory Commission NLS-87-226 / Page 3 During the startup test program and the initial months of commercial operation, the SHNPP has experienced a number of plant trips which serve to demonstrate the dynamic response of the plant. In that the expected plant response is a reactor trip from 10096 power, the dynamic safety systems response can be expected to be similar to that of previous reactor trips from 10096 power. Reactor trips from 10096 power are discussed in the following LERs: 87-031 dated 3une 23, 1987;87-035 dated 3uly 13, 1987;87-001 dated September 3, 1987; and 87-002 dated August 10, 1987.

In establishing the course of action to be taken with regard to the generator loss of load test, CPRL has considered the continued safe operation of the plant and the severity of conducting the test. It is CPRL's position that it is safe and prudent to rely on completed testing for confidence that the dynamic response of the plant has been demonstrated.

Regulatory Guide 1.68 establishes the basis for an initial test program for a nuclear power plant which is acceptable to the NRC. Clearly, there is a grouping of plant transients which the plant is designed to safely accommodate; however, because of their severity, a specific plant test is not performed to demonstrate the capability.

Correspondingly, there is a second grouping for which specific tests are conducted. For the SHNPP design, the performance of a generator loss of load test is equivalent to a loss of offsite power at 100 percent reactor power. While Regulatory Guide 1.68, Appendix A, Section 5.jj calls for a loss of offsite power test in the 10 to 20 percent power range, it does not require that this test be done at 100 percent power. It is CPRL's position that while the SHNPP is designed to accommodate a loss of offsite power from 100 percent reactor power, this transient is of sufficient severity to warrant not intentionally conducting such a test.

An additional issue to consider relative to conducting a generator loss of load test has been raised by Westinghouse with the Westinghouse Owners'roup (WOG). A scenario is postulated wherein a loss of forced reactor coolant flow occurs at elevated reactor coolant system (RCS) conditions as a result of loss of external load combined with failure of a reactor trip on turbine trip and failure of an automatic fast bus transfer. One possible, although not expected, sequence of events during conduct of generator loss of load test at SHNPP could lead to a situation similar to that postulated by Westinghouse.

In the SHNPP case, if the reactor and turbine do not trip quickly and if the steam dump control system and rod control system fail to operate after the opening of the generator output breakers, the RCS will begin to heat up. During this time, the reactor coolant pumps (RCP) will operate off power supplied by the generator. When the reactor does trip, the turbine will also trip and power to the RCPs will be lost as fast bus transfer is blocked by the open generator output breakers. The result is a loss of forced reactor coolant flow at elevated RCS conditions. The WOG is currently evaluating this generic issue and has concluded that continued operation is justified; however, it is CPRL's position that it is neither prudent nor conservative to deliberately expose the plant to this possible sequence of events by running this test.

t United States Nuclear Regulatory Commission NLS-87-226 / Page 0 In summary, CPRL believes it has satisfied the intent of the large load reduction from full-power test. This position is justified based on: l) the low probability of a load rejection occurring at the Shearon Harris Plant, 2) the expected response to such an event is a safe shutdown which has been previously demonstrated in other tests, 3) the requirement to demonstrate maximum credible turbine overspeed has been resolved, and

0) deliberately causing this severe transient is not a prudent course of action.

Should you have ariy questions concerning this subject, please contact Mr. 3im Kloosterman at (919) 836-8055.

Yours very truly, S.. Zim erman ager SRZ/3DK/pp Nuclear Licensing Section Cco Mr. B. C. Buckley Dr. 3. Nelson Grace Mr. G. F. Maxwell Mr. A. R. Herdt