ML20216H870

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Forwards Response to RAI on Proposed Amend Re Natural Circulation Testing
ML20216H870
Person / Time
Site: Catawba Duke Energy icon.png
Issue date: 09/10/1997
From: Gordon Peterson
DUKE POWER CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TAC-M98728, NUDOCS 9709170090
Download: ML20216H870 (12)
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Text

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  • Duke Power Company A tu two Gmper Casauha Nudeer Station Atu We 4800 Corwood Road i

York. SC 29745 Gary R. l%reon (803) 831-4251 omCE Yur h nidens (803) 83).3426 fat September 10, 1997 U.S. Nuclear Regulatory. Commission Attention: Document Control Desk Washington, D.C. 20555

Subject:

Catawba Nuclear Station, Unit 1 Docket Number 50-413 Reply to Request for Additional Information on the Proposed Amendment Regarding Natural Circulation Testing (TAC No. M98728)

Reference:

Letter-from Peter S. Tam, NRC, to Gary R.

Peterson, Duke, Request for Additional Information on the Proposed Amendment Regarding Natural Circulation Testing, Septenber 2, 1997 Gentlemen:

Please find attached Catawba Nuclear Station'_s reply to the request for additional information contained in the reference letter. Each NRC question in the reference letter is restated, followed by our response.

If you have any questions concerning this information, please call L.J. Rudy at (803) 831-3084.

Very truly yo rs, t k G.R. Peterson LJR/s CD I ' ,

Attachment 9709170090 970910 -

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Document Control Desk

.Page'2 September 10, 1997 xc (with' attachment):

L.A. Reyes, Regional Administrator Region II D.J. Roberts, Senior Resident Inspector Catawba Nuclear Station P.S. Tam,. Senior Project Manager ONRR

_ _ _ _ _ _ l

ATTACHMENT REPLY TO NRC REQUEST FOR ADDITIONAL INFORMATION CATAWBA UNIT 1 NATURAL CIRCULATION TEST FOR REPLACEMENT STEAM GENERATORS

Background

The purpose of this test is to demonstrate the ability of the Nuclear Steam Supply System (NSSS), and more specifically, the BWI Replacement Steam Generators (RSGs),

to remove heat via natural circulation of the primary coolant and to gather information useful in fine-tuning simulators for natural circulation conditions. This test will be performed in Mode 3, during the normal planned shutdown for the Unit 1 End of Cycle 10 (EOC10) refueling outage.

Calculations performed by the Steam Generator Replacement project team showed that natural circulation capabilities of the RSGs should be slightly superior to original SGs. This e~ is due to their greater heat transfer surface area and higher thermal center due to a 1,nger tube bundle. The RSGs are already modeled in approved safety analyses. This test is therefore not being conducted as a requirement, but rather at the direction of the Duke Nuclear Safety Review Board (NSRB) and management to gather data to benchmark the training simulator and to thereby improve operator knowledge of RSG characteristics. As such, the test procedure utilizes equipment most likely to be used in such a scenario, and uses it consistent with existing procedures.

NRC Reactor Systems Branch Position RSB 5-1, which requires the capability to shutdown to cold shutdown from the control room using redundant safety grade equipment, is not adversely impacted, and all such equipment will be verified operable prior to this test.

The positive impact of the RSGs on natural circulation has been confirmed by a RETRAN thermal-hydraulic simulation of a loss of offsite power scenario. The results of this analysis show that, within 5 minutes of the loss of offsite power, a stable reactor coolant system (RCS) flow rate of approximately 5% of the full power value is established with a core delta T of less than 300F.

It should be noted that operators at Catawba Nuclear Station have had recent experience with natural circulation not only on the training simulator, but also during the Unit 2 Loss 1

of Offsite Power (LOOP) Event of February 6, 1996. During

,this event, operators controlled the unit using natural circulation to achieve Mode 4 (hot shutdown). There were no significant problems as a result of this event, and natural circulation proceeded to cool the RCS until the residual heat removal (RHR) system was placed in operation 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> later.

This Natural Circulation Verification Test is expected to take only a few hours while in Mode 3 (hot standby) to gather data, meet acceptance criteria, and return reactor coolant pumps to service for normal cooldown.

Answers to the specific NRC questions are provided in the following responses.

NRC Question #1 Describe the test procedure and the major recovery steps in the procedures that will be used for recovery from the test. Include a list of prerequisites for the test including those related to auxiliary feedwater system status, steam generator relief valve status, charging /high head safety injection system status, etc.

Response

Description of the Test Procedure:

  • The procedure initially begins with the unit in a stable condition in Mode 3, Hot Standby:
  • While the unit is in Mode 3 (557 i 2 F, 2235 i 25 psig), all four reactor coolant pumps (RCPs) will be simultaneously tripped.

Establishment of natural circulation will be verified by observation of wide range loop temperatures as well as core exit thermocouples.

  • Pressurizer and SG pressure and level response will be monitored throughout this test.
  • Pressurizer auxiliary spray from the chemical volume control system (CVCS) will be actuated if additional pressure control is needed. The procedure specifies that letdown be in service to maintain an acceptable 2

temperature of the auxiliary spray so that the use

, of auxiliary spray is acceptable.

Stable natural circulation will be maintained for approximately 30 minutes while data are gathered to verify that the acceptance criteria have been met.

  • The Operator Aid Computer (OAC) transient monitor will be frozen to obtain data of plant parameters during the transition period to natural circulation and during the natural circulation period itself.
  • The plant will be recovwred by realigning normal charging and restarting the RCPs per normal operating procedure.
  • A detailed evaluation of the data will be performed against certain review criteria.

Major Recovery Steps:

Recovery from this test either from normal or early termination is performed by exiting the test procedure and using preexisting operating procedures:

OP/1/A/6200/01 " Chemical and Volume Control System" OP/1/A/6150/02A " Reactor Coolant Pump Operation" OP/1/A/6100/02 " Controlling Procedure for Unit Shutdown" The following emergency procedures are available to the operator in case of a design basis accident that would make all the RCPs unavailable for continuation of the shutdown:

EP/1/A/5000/ECA-0.1 " Loss of AC Power Recovery without SE Required", and EP/1/A/5000/ES-0.2 " Natural Circulation Cooldown" Test Prerequisites:

1) All RCPs are in operation.
2) The CVCS is aligned for power operation with letdown and charging in service.
3) The following control systems are operable and have been placed in automatic control:

3

{

  • pressurizer pressure control
  • pressurizer power operated relief valve (PORV) control
  • pressurizer spray control e secondary steam dump control (pressure mode) to maintain approximately 1092 psig main steam pressure and RCS T-cold at approximately 557*F
  • 1NV-294(centrifugal charging pumps A & B flow control)
4) Verify the following operable:
  • All pressurizer PORVs (with associated block valves open)
  • All pressurizer safety valves All main steam PORVs (with associated block valves open)
5) OAC RCS system saturation graphics program is operable and outputs are valid.
6) SG levels are being maintained using main feedwater pump speed control.
7) Conduct shift briefing.
8) Establish equilibrium conditions as follows:
  • RCS T-avg is stable at 557 i 2 F
  • RCS pressure is stable at 2235 i 25 psig
  • SG levels are at no-load level 1 3%
9) Restrict work in the switchyard area.
10) Verify one time change to Technical Specification 3.4.1.2 to allow all four RCPs to be turned off for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> in Mode 3 has been approved by the NRC.
11) In addition to the above, the Outage Manager shall ensure the following key safety functions needed to allow natural circulation testing:

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  • Two Technical Specification boration flow paths

. operable

  • Source range or boron dilutio'n mitigation system (BDMS) operable per Technical Specifications
  • Boron concentration 2 cold shutdown concentration
  • 1A & 1B centrifugal charging / safety injection trains operable
  • Containment integrity established per Technical Specifications
  • 4160 volt emergency busses 1 ETA and 1ETB energized from offsite source
  • Standby auxiliary transformer SATA or SATB powered from Unit 2
  • Unit 1 is not in any Technical Specification action statement of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or less duration NRC Question #2: During the initial test, you experienced a problem with the reactor coolant pump that you described as follows: " a reactor coolant pump could not be restarted for a period of time at the completion of the test due to a high standpipe level." Discuss how this problem will be prevented in this test. Discuss this not only fram the perspective of violating technical specifications but from the perspective of being able to start the pump.

Response

During the initial Catawba Unit 1 natural circulation verification performed on January 19-20, 1985, a high standpipe level indication resulted in the inability to restart RCP A while recovering from the test. This test is documented in the Catawba Unit 1 Startup Report, RCP standpipe level problems have been resolved by drain line rerouting some years ago. Some level problems are not 5

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unusual during plant startup following seal maintenance or

, replacement and usually indicate that the seal faces are not in proper contact. Such a problem is unlikely during this test because seal leakoffs will be verified to be within normal range and the RCS will remain hot and pressurized throughout the test. Standpipe levels and other RCP operating parameters during the last two startups were examined and were found to be within acceptable limits.

Plans include having the RCP equipment engineer onsite for the test to assist in monitoring key parameters and resolving any problems if they occur. If a RCP cannot be started for any reason, operators will enter the appropriate procedure for loss of a RCP and proceed as directed by plant procedures and Technical Specifications. Mode 3 Technical Specifications requires 3 of 4 RCPs to be operable and in operation. The RHR system may be placed in service below 3 5 0'F .

In addition to the plant emergency procedures, the safety parameter display system (SPDS) directs the operators along the path for a loss'of heat sink, loss of inventory, etc.

The critical safety functions are then followed along the heat sink " red path", " yellow path", etc. to address the specific function. The test procedure contains termination criteria as well as acceptance criteria. Also, see response

  1. 4 for a discussion of the training provided to the operators.

NRC Question #3: The initial test was required to be performed at core burnups to ensure that no significant core decay heat levels are present. In this case, a reactor trip would effectively terminate the event. In your proposed test, you will use decay heat to conduct the test and therefore, there will be significant decay heat. Discuss any contingencies that you will have in place to cool the core in case of unexpected system behavior.

Response

This type test was required for initial Unit 1 startup.

Whereas the initial test was performed with the reactor critical (Mode 2) and overtemperature delta temperature (OTAT) and overpower delta temperature (OPAT) trip setpoints bypassed, this procedure is designed to utilize natural decay heat in Mode 3. The initial test was performed during the initial startup program and decay heat was not available, therefore, nuclear heat was used instead.

Performance of this test in Mode 3 instead of Mode 2 has 6

been judged by Engineering and management to have lower

,overall plant risk. No reactor trip setpoint will be bypassed during this test and the possibility of violating the minimum temperature for criticality ( 551*F ) will not be an issue while in Mode 3.

In addition to providing forced cooling, the RCS pumps during Mode 3 are required operable for boron mixing in the event of an inadvertent dilution. The RCS boron concentration will be increased to greater than or equal to the cold shutdown boron concentration prior to shutting down the RCPs. Performing the natural circulation test in Mode 3 with all rods in and cold shutdown boron concentration is considered more conservative from a reactivity management perspective. Temperature control with rods would not provide as good or realistic data to benchmark the training simulators, which is the main purpose of this test. The RSGs have been shown by analysis to perform better than the original SGs, and startup testing and turbine online response has been closely analyzed to assure that expectations were met. The RSGs show good level stability, smooth transition to turbine online, and less shrink and swell during transients. The test implementation plan contains a checklist that specifies that all safety equipment is operable, in addition to the required status of all applicable automatic control systems. Main feedwater and steam dump to the condenser will be employed. Auxiliary feedwater and the SG PORVs will be fully operable as a backup. Redundant trains of emergency core cooling systems (ECCS) and normal RHR cooling will also be required operable.

NRC Question #4 Discuss any special training (class room, briefings, simulator, etc.) that the operators were or will be provided prior to conducting this test.

Response

Training personnel have been involved in the procedure validation using the Catawba simulator. Operators on the shift performing this test will receive "Just in Time" training in the form of a pre-job briefing, simulator exercise, and a second pre-job briefing in the control room immediately prior to the test. It should be noted that

  • Catawba Nuclear Station has recent experience with natural circulation not only on the training simulator, but also during the Unit 2 Loss of Offsite Power Event of February 6, 1996. During this event, operators controlled the unit using natural circulation to achieve Mode 4. There were no 7

significant problems as a result of this event, and

, procedures have been optimized to provide better guidance based on lessons learned.

NRC Question #5: In your submittal, you provided a list of example test termination criteria. Provide a complete list of the test termination criteria and a list of expected values for the same parameters. Also, discuss and justify your expectation of system pressure response and primary power-operated relief valve operation.

Response

Specific test termination criteria are provided:

  • Lowest RCS subcooling margin s 15'F
  • SG 1evel Inarrow range) on any 2 loops < 12%
  • Pressurizer level < 17% or > 5% unexplained decrease
  • RCS T-hot (wide range) or any valid incore thermocouple reading > 590 F
  • RCS AT in any loop > 45*F
  • Steam line pressure s 900 psig Dilution of RCS (controlled or uncontrolled)

(Technical Specification 3.4.1.2)

  • RCPs have been deenergized for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (Technical Specification 3.4.1.2 when approved by NRC)

During the performance of this test, the following parameters are expected to be maintained within the specified limits:

  • Lowest RCS subcooling margin 2 25 F
  • SG level (narrow range) > 25%

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  • Pressurizer pressure > 1955 psig
  • Pressurizer level with RCPs running > 25%
  • Pressurizer level with RCPs off 2 level at time pumps tripped off ,
  • RCS T-hot (vide range) and incore thermocouple readings s 590'F .

i

  • RCS T-cold (wide range) 557 i 5*F
  • RCS AT in every loop 5 45'F Loops are filled and pressurized to normal operating pressure (2235 psig) in Mode.3, and will remain so throughout the test. SG pressure control is provided by the condenser dumr valves to ensure stable test conditions.

Pressurizer heaters will be operable and will cycle as needed. This will also ensure that the primary side pressure is stable and'that level remains in the pressurizer within the expected range. Letdown is in service and primary pressure control is by pressurizer auxiliary spray since the RCPs-will be secured. During simulator validation of the test procedure, auxiliary spray was capable of preventing a RCS pressure increase up to the PORV setpoint.

Additionally, the PORV controller has anti-windup protection to limit PORV response due to prolonged operation close to,.

but below its 2335 psig setpoint. Pressurizer-PORVs are not expected to open, but plant procedures assure using these as a backup means of pressure control. Chemistry control assures there is no non-condensible gas present. Separation of non-condensible gas does not occur in this mode as it can in Mode 5 when the RCS is depressurized. It is not part of this test to cool the plant down to Mode 5, but rather to restart the RCPs for normal forced flow during the cooldown from Mode 3 to Mode 5.

NRC Question #6 In your submittal, you state that the operators will verify natural circulation by observation of the wide range loop temperatures and core exit thermocouple.

Explain how this verification is accomplished and the type of training operators received in this regard (i.e., how will operators know if there is natural circulation).

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Response

'Oper'ators are trained in the diagnosis of natural circulation as a function of training on emergency i procedures (e.g., Ep/1(2)/A/5000/ECA-0.1 " Loss of AC Power Recovery without SI Required" and EP/1(21/A/5000/ES-0.2

" Natural Circulation Cooldown"). After RCP trip, core exit temperatures will increase and core AT between hot leg and cold leg is expected to increase to approximately 10'F -

40'F depending on the time since shutdown and the number of loops that go into natural circulation. Given the test prerequisites, all 4 loops should go into natural circulation. Operations personnel participating in the simulator validation of this natural circulation test procedure had no problem monitoring core AT, subcooling margin, SG pressure control, and associated parameters using OAC graphics, the SPDS, and control board gauges. Example OAC displays showing core AT and pressurizer pressure and level, taken during the simulator validation of this procedure, will be provided to the shif t as part of the pre- o job briefing.

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