ML19331B933

Forwards Response to Open Items Re Fuel Loading & Low Power Testing Resulting from 800801 Meeting W/Nrc.Items Include Shift Manning,Training for Mitigating Core Damage & Independent Safety Engineering

ML19331B933
Person / Time
Site:	Farley
Issue date:	08/06/1980
From:	Clayton F ALABAMA POWER CO.
To:	Schwencer A Office of Nuclear Reactor Regulation
References
800606, NUDOCS 8008130555
Download: ML19331B933 (1)

Text

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6 6 Alabama Power Company 600 North 18th Street

. Post Office Box 2641 .

Birmingham, Alabama 35291 Telephone 205 250-1000 P. L CLAYTON, JR.

senior Vice President Alabama Power August 6, 1980 Docket No. 50-364 Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Mr. A. Schwencer JOSEPH M. FARLEY NUCLEAR PLANT - UNIT 2 RESPONSE TO OPEN ITEMS TO SUPPORT FUEL LOADING AND LOW POWER TESTING Gentlemen:

The NRC Staff in the August 1,1980, meeting concerning the above referenced subject requested that Alabama Power Company address several questions. Alabama Power Company hereby submits the enclosed response to the NRC's questions for Farley Nuclear Plant - Unit 2.

If you have any further questions, please advise.

Yours very truly, 7

MCLSh<ah-C Clayton, Jr.

RWS/rt Enclosure cc: Mr. R. A. Thomas Mr. G. F. Trowbridge Mr. L. L. Kintner Mr. W. H. Bradford }Ohj 5

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l ENCLOSURE I.A.1.3 SHIFT MANNING Position .

(1) Modify plant administrative procedures to require that a second senior reactor operator (rather than a shift supervisor) remains in the control room.

(2) Establish an administrative procedure for limiting overtime of personnel performing safety related functions.

(3) Modify proposed technical specifications to reflect current requirements for minimum number of licensed personnel during various modes of Unit 2 operation.

Response-t Alabama Power Company will respond to the NRC position by August 8,1980.

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II.B.4 TRAINING FOR MITIGATING CORE DAMAGE Position Provide a description of program for training operators in the use of.in-stalled systems to monitor and control accidents in which the core may be severely damaged.

Response

The content of the licensed operator requalification program has been modi-fied to include instruction in heat transfer, fluid flow, thermodynamics and mitigation of accidents involving a degraded core, as outlined in our response dated July 29, 1980. In addition, all Shift Technical Advisors (STA's) are required to participate in the requalification program. Examina-tions will be administered upon completion of the lecture series. The new grading criteria of 80% overall and 70% minimum for each section will be utilized to demonstrate adequate knowledge of the material. Training for those persons who fail to meet the new grading criteria is described in Alabama Power Company's requalification program.

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t' I.B.1.2. INDEPENDENT SAFETY ENGINEERING Position Establish an independent, dedicated, safety engineering group onsite that reports at high engineering management level offsite.

Response

Alabama Power Company will respond to the NRC position by August 8,1980.

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II.E.1.2 AUXILIARY FEEDWATER INITIATION AND INDICATION Position Commit ~to modify the power supply to auxiliary feedwater control valves to meet single failure criterion. .

Response

Alabama Power Company, in order to meet single failure criteria, will provide power supply train separation for the solenoid valves ' associated with the auxiliary feedwater flow control valves prior to exceeding 0% power.

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II.F.2 INADEQUATE CORE COOLING INSTRUMENTATION

. Position For the proposed reactor vessel level measurement, provide: ,

(1) a tangible plan for installation of a specific vessel level .

measurement system; (2) a schedule for installation, testing and calibration and implementation; and

, (3) a commitment to provide procedures and related analyses for the use of the system for staff review and approval prior to implementation of the system.

Response

The non-invasive watgr level measurement system consists of a set of externally mounted luBF3 neutron detectors above and below the reactor vessel. The principle used is the detection of photoneutrons from the reaction of high energy ganinas with the deuterium impurity present in the reactor coolant system. A simplified diagram of the system is shown in Figure 1. Each detector set above or below the reactor vessel consists of

, eight -(8) 2-inch diameter, 24-inch active length 10BF3 filled thermal neutron counters. These detectors are made from stainless steel and are filled to 70 cm. Hg. pressure. .They are shielded by a 1/2-inch thick lead sleeve, and they are surrounded by a plastic moderator. Additional 1/2-inch lead shielding is provided between the vessel and the detectors. The ratio of

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the count rates from these two sets of detectors is used to detennine the water level-in the vessel above the core.

The detector assembly above the reactor vessel consists of two counters sheathed in 0.065-inch thick steel. Figure 2 shows the preliminary sketch of the cross-section of the top detector assembly. Four detector assemblies are then mounted in position above the vessel head as shown in Figure 3. The top detectors are expected to'have a combined sensitivity of approximately 500 counts per NV. The bottom detectors are mounted in the area below the reactor yessel. All eight of these detectors are consolidated into a single assembly as shown in Figure 4. Since these large detectors are too sensitive -

for-use during operating power levels, a pair of small, less sensitive 10B lined detectors will be used to measure level during power operation, if necessary.

While details of electrical wiring diagrams for the system have not yet been finalized, Figure 5 shows the features that will be incorporated in this system.

The 10BFr detector assemblies are divided into pairs with their associated preamp 1.ifiers so that the redundancy requirements will be satisfied. The

. amplifiers drive two-counter timers. One counter operates on the top detectors,. set-up for a preset number of counts. Operation of this counter

gates the second counter which accumulates counts from the bottom detector

during the period when the first counter is counting a preset number of counts. Consequently, the counts from the second scaler is proportioned to:

Bottom Count Rate Top Count Rate The reduction in the amount of water above the reactor core increases"the top count rate, thereby reducing the ratio. Thus, the second counter display counts increase with increased water level.

Background:

During the sumer of 1979, under EPRI sponsorship, the National Nuclear Corporation (NNC) conducted tests using a 252Cf neutron source in a water tank mockup to determine whether neutron measurements outside a reactor vessel would provide an unambiguous measure of the water level inside the i vessel. The method chosen is shown schematically on Figure 1. An array of neutron detectors is placed above and below the vessel, and the ratio of the counts from these arrays is related to the water level. Figure 6 shows j count rates measured above the vessel as a function of water level. The initial rapid drop off for about four feet is due to shielding, by the '

water, of fission neutrons from the source (or reactor). Beyond four feet, neutrons are principally produced from the action of high energy (over 2.2 MeV) gamma rays on the deuterium within the water. Since these gamas travel further than neutrons in water, these photoneutrons predominate when the water in the tank is deeper than four feet over the core. When the counts from the lower detector are divided by counts from the upper detector (as in Figure 7),

a relationship roughly proportional to water depth is obtained, except just before core uncovery when a much greater effect is observed.

As shown by this data from tank tests at NNC, for water levels over four feet above the core, most of the neutrons detected arise from interaction of high energy gammas with the deuterium impurity in the water, while below five feet (where there is a danger of core uncovery) the neutron level above the reactor rises very rapidly due to neutrons produced by fission in the reactor core. Thus, this system provides a vivid warning well before core uncovery. Less dramatic indications are provided to gauge water level in the range between five feat and full. This is shown in the correlation on Figure 7. In this test, top counts were referenced against a side detector, instead of the bottom detectors used in the actual installation.

Following these tests and additional tests at Prairie Island and Rancho Seco, equipment was built and used to demonstrate successfully the operation of this system at Trojan during initial drain-down. Data from the Trojan test is shown on Figure 8. Based on this data, and on Trojan side counter data, the curve on Figure 9 has been projected to indicate the performance of the actual system one day after shutdown.

Following the Trojan test, analytical studies have been made to further investigate the system's capabilities. It has been brought out that the system reads weight of watar above the core, thus giving a valid indidation I before core uncovery.

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Farley - Unit 1 Program:

Based on the results of the demonstration program undertaken by EPRI and NNC thus 'far, APC0 feels that the non-invasive water level measurement system could offer a viable alternative to the other developmental ( AP) system. Some of the additional incentives for considering this system are:

(a) more easily installed in existing reactor (b) measures water weight per unit area above the core and responds to voiris above the core (c) not affected by voids below the core (d) not affected by pressure drep in the system (e) gives large unambiguous signal before core uncovery (starting abour 4 feet above the core)

Consequently, Alabama Power Company, in coordination with EPRI, will under-take a program on Farley - Unit 1 to demonstrate the capability of c. .irototype system on an experimental basis. EPRI will continue to fund the analytical investigations associated with the prototype demonstration program and will assist APC0 in establishing the necessary test programs and evaluation of the collected data.

The prelimir.ary system discussed in " General Description" above will be installed in Farley - Unit 1 during the October 1980 outage. The system will be calibrated and part of the necessary tests will be conducted during this 4

outage. During startup, after completion of the refueling, effects of density change with temperature will be investigated. Also, durina a full-power operation period, the small detectors will be evaluated for their performance. Additional necessary water level measurement tests will be performed during unplanned outages, but no later than the next refueling outage currently planned to commence during the fall of 1981. During this entire period continued engineering evaluations of the system will be performed towards improvement of the cesign, installation, and qualification of the system. If the system proves to be viable and reliable during the Unit 1 tests, then the same system will be installed in Unit 2 during the first refueling outage currently planned to commence mid-1982. However, if the system fails to meet performance expectations (determination to be made by the end of Unit 1 fall 1981 refueling outage) then APC0 will install, in both units, an alternate system having been proven or having been accepted as having the capability of meeting the requirements.

It is recognized. that the non-invasive water level system is still in its l developmental stage and the effort undertaken by APC0 could be construed l as the first prototype field testing program. However, it is our feeling

- this it is no less proven, at this stage, than the A P alternative used by others which has inherent drawbacks such as maintainability, accuracy, l and difficulty of calibration. Both systems have their advantages and l drawbacks. Therefore, we see a necessity to help develop this alternative

! - and are willing _to devote resources toward that objective. All the details l of the program and design are not yet finalized and we are working as expeditiously as possible to have the design and ~ test program established i 'and equipment available at the site prior to the October 1980 Unit 1 re-fueling outage. However, APC0 will keep the NRC fully informed of the design and test program as they are finalized, and the test results as they are

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II.D.3 RELIEF AND SAFETY VALVE POSITION INDICATION Position Limit switches and alarms for relief and safety valve position indication must be installed before fuel loading. -

Response

Position indication with main control room alarms has been installed for the pressurizer power operated relief valves (PORV's). Limit switches for the pressurizer safety valves have been received and will be installed upon receipt of the mounting brackets presently scheduled for delivery by September 1,1980.

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'II.K.3 - C.3.10 ANTICIPATORY TRIP Position The reactor protection system must be set to permit by-pass of the antici-patory reactor trip-actuated-by-turbine-trip only below 10% of reactor power, until sufficient data from operating plants is provided to demonstrate that the power operated relief valve on the pressurizer is not actuated when the set point is raised to 50% of reactor power.

Response

Alabama Power Company is in the process of obtaining the required date and will respond to the NRC position by September 1,1980, as agreed to via phone with the NRC Staff on August 5, 1980.

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