ML19326E218

Responds to NRC Request for Evaluation of Plant RCS Susceptiblity to Overpressurization Events.Mass Input & Energy Input Analyses Sections Were re-evaluated Using New Valve Discharge Coefficients

ML19326E218
Person / Time
Site:	Fort Calhoun
Issue date:	07/15/1980
From:	William Jones OMAHA PUBLIC POWER DISTRICT
To:	Clark R Office of Nuclear Reactor Regulation
References
NUDOCS 8007280530
Download: ML19326E218 (6)

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Omaha Public Power District W/

1623 HARNEY a OMAHA. NEBRASKA 68102 e TELEPHONE 536-4000 AREA CODE 402 July 15, 1980 Mr. Robert A. Clark, Chief U. S. Nuclear Regulatory Commission Operating Reactors Branch No. 3 Washington, D. C. 20555

Reference:

Docket No. 50-285 4

Dear Mr. Clark:

In response to the Commission's request for an evaluation of the Fort Calhoun Station's reactor coolant system's susceptibility to over-pressurization events, the Omaha Public Power District submitted a rcport entitled " Low Temperature Overpressure Protection for Fort Calhoun Unit 1". In forwarding this report to the Commission, the District's letter of May 10, 1977, stated that the capability of the power operated relief valves (.PORV) for preventing low temperature overpressurization incidents would be confirmed by means of a qualifi-cation test program. Performance and evaluation of this test program has been completed, with the following results:

(1) The test PORV was cycled a total of 24 times during two-phase and subcooled flow conditions. Upstream temperatures ranged from 1200F

, to 4600F; upstream pressures ranged from 300 psig to 600 psig. The valve opened and closed properly during each cycle and no degrada-tion of performance was noted.

(2) Valve discharge coefficients were computed to range from .46 to

.61, depending upon the amount of upstream subcooling, over the area of interest discussed in the referenced report.

(3) Following performance testing, a seat leakage test was conducted using nitrogen at 2266 +/- 23.85 psig. After one hour of pres-surization, leakage was measured at 4.95 cc/ minute.

Based on the results of this test program, the mass input analyses and energy input analyses sections of the " Low Temperature Overpressure Protection" report were reevaluated using the new valve discharge coefficients. As a result of this reevaluation, forty (40) copies of the attached replacement pages are forwarded as an update to our original report.

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Mr. Robert A. Clark,' Chief July 15, 1980

Page Two It _is the District's position that the PORV's are acceptable for -

low temperature overpressure protection.

Sincerely,

.fb gqW W. C.j uones Division Manager Production Operations WCJ/KJM/BJH/PRT:jmm

- Attachments cc: LeBoeuf, Lamb, Leiby & MacRae 1333 New Hampshire Avenue, N. W.

Washington, D. C. 20036

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k.0 RESULTS'0F ANALYSES Results of the solid water RCS mass and energy input analyses which were discussed in Sections 3.1 and 3.2 are shown in Figure 8. These tran-sients indicate that the most severe overpressurizations are:

1) a RCP start with hot steam generators; and

2) an inadvertent SI actuation.

These incidents are considered as design bases for Fort Calhoun over-pressure protection. Any means which vill mitigate the design base events are sufficient for any of the less severe incidents. This section vill discuss the results of analyses which examined the effectiveness, upon terminating the design base transients, of a procedurally controlled pres-surf:er steam bubble or a single low setpoint PORV.

The results of these analyses are based upon the assumption that the existing PORV's are satisfactory for low-pressure relief. This assumption has been verified with a valve test program.

h.1 RCP Start Transient The RCP start transient which results from a positive secondary to prinary AT is shown in Figures 8 and 9A to be one of the potentially most severe transients. This is due to the rapid increase of pressure. The transient may, however, be easily controlled by a pressurizer steam bubble.

The required size of this volume would be achieved by forming a pressurizer steam space of approximately 60 percent. The design criterion requiring 10 minute operator inaction is satisfied for this incident, provided an overpressuri=ation alarm is incorporated in plant design which alerts the operator of the pressure excursion. In addition to a RCP start transient, other overpressurization incidents which can be more than adequately controlled by' maintaining a pressurizer bubble are:

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B.1 Retetor Coolant System (Continutd) of decay heat rates at initiation of shutdown cooling, after attaining refueling water temperatures (30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, minimum), and after one week are 1.0, 0.6, and 0.3 percent, of full thermal power, respectively.

Reactor coolant pump heat is conservatively considered the thermal equivalent of the pump power rating. When a reactor coolant pump is started while the steam generators are hot (i.e., secondary inventories are at higher temperatures than the reactor vessel coolant), a positive secondary to primary temperature differential is created, and thus heat transfer to the RCS. This heat input vill result in a severe overpres-surization in a solid system. RCP start / hot steam generator transients parameters for Fort Calhoun are.found in Table 4. Examples of pressure transients initiated by a reactor coolant pump start with a 500F AT between reactor vessel coolant and steam generator secondary inventories are given in Figures 9A and 9B.

Located on the pressurizer are two electromatic solenoid operated relier valves, also designated as power operated relief valves (PORVs).

These are electrically operated relief devices which may be operated  ;

manually by closing a svuch, or may be aligned with a pressure sensi-tive element to relieve pressure automatically. The flow area for the Fort Calhoun PORVs is 0 9h in2; inlet and outlet diameters are 2 and k inches, respectively. Valve discharge rates were determined by assuming a liquid discharge coefficient between .46 and .61 (depending upon degrees of subcooling), a pressurizer water temperature of 4000F and a backpressure l

of 100 psi. Hence, relief capacities varied with the upstream pressure; e.g., 400 and 500 psia upstream pressure yield 388 and 451 gpm, respectively.

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