ML14188B263
| ML14188B263 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 12/30/1986 |
| From: | Requa G Office of Nuclear Reactor Regulation |
| To: | Utley E Carolina Power & Light Co |
| References | |
| GL-85-012, TASK-2.K.3.05, TASK-TM NUDOCS 8701020250 | |
| Download: ML14188B263 (20) | |
Text
Docket No. 50-261 Dneket Fil%)
G. Requa December 30, 1986 NRC POR D. Miller Local PDR ACRS (10)
Mr. E. E. Utley, Senior Executive Vice President PAD#2 Rdg Tech Branch Power Supply and Engineering and Construction T. Novak Gray File Carolina Power & Light Company OGC-Bethesda Post Office Box 1551 E. Jordan Raleigh, North Carolina 27602 B. Grimes J. Partlow
Dear Mr. Utley:
N. Thompson
Subject:
Reactor Coolant Pump Trip (TMI Action Item II.K.3.5, Generic Letter 85-12); Conference Call, H.B. Robinson Steam Electric Plant Unit 2 The staff has reviewed your submittal dated September 30, 1985, in response to Generic Letter 85-12. We have developed preliminary comments and questions which we discussed with your staff during a conference call on October 8, 1986.
During the conference call, we agreed to transmit our understanding of your staff's responses.
Enclosed is our report of the conference call.
Please have your staff confirm or correct the report as necessary and provide us with the confirmation or corrections by letter within 45 days of receipt of this letter.
This information request affects fewer than 10 respondents; therefore OMB clearance is not required under P.L.96-511.
Sincerely, Glode Requa, Project Manager PWR Project Directorate #2 Division of PWR Licensing-A
Enclosure:
As stated cc:
See next page
- 2 PM:PAD PD:PAD#2 GRequf LRubenstein 12/- /86 12/ /86 12/1s/86 8
6701020250 861230 PDR ADOCK 05000261 P
- 9.
Mr. E. E. Utley Carolina Power & Light Company H. R. Robinson 2 cc:
Thomas A. Baxter, Esquire Mr. Dayne H. Brown, Chief Shaw, Pittman, Potts and Trowbridge Radiation Protection Branch 2300 N Street, N.W.
Division of Facility Services Washington, DC 20037 Department of Human Resources 701 Rarbour Drive Raleigh, North Carolina 27603-?008 Mr. McCuen Morrell, Chairman Darlington County Roard of Supervisors Mr. Robert P. Gruber County Courthouse Executive Director Darlington, South Carolina 29535 Public Staff - NCUC P.O. Box ?9520 Mr.-H. A. Cole Raleigh, North Carolina 27626-0520 Soecial Deputy Attorney General State of North Carolina P.O. Box 62Q Raleigh, North Carolina 2760?
Mr. D. E. Hollar Associate General Counsel Carolina Power and Light Company P.O. Box 1551 Raleigh, North Carolina 27602 U.S. Nuclear Regulatory Commission Resident Inspector's Office H. B. Robinson Steam Electric Plant Route 5, Box 413 Hartsville, South Carolina 29550 Regional Administrator, Region II U.S. Nuclear Requlatory Commission Suite 2900 101 Marietta Street Atlanta, Georgia 30303 Mr. R. Morgan General Manager H. B. Robinson Steam Electric Plant Post Office Box 790 Hartsville, South Carolina 29550
FNCLOSURE REPORT OF CONFERENCE CALL CONCERNING REACTOR COOLANT PUMP (RCP) TRIP AT CAROLINA POWER AND LIGHT COMPANY'S ROBINSON UNIT 2 PLANT, OCTOBER 8, 1986
SUMMARY
We reported in Reference 1 that the information provided by the Westinghouse Owners Group (WOG) in support of alternative Reactor Coolant Pump (RCP) trip was acceptable on a generic basis. The review noted that a number of considerations were assigned plant-specific status. Accordingly, we requested that operating reactor licensees and applicants select and implement an appropriate RCP trip criterion based upon the WOG methodology.
Reference 1 required owners of Westinghouse Nuclear Steam Generating Systems to evaluate their plants with respect to RCP trip. The objective was to demonstrate that their proposed RCP trip setpoints assure pump trip for small break LOCAs, and in addition to provide reasonable assurance that RCPs are not tripped unnecessarily during non-LOCA events. A number of plant specific items were identified which were to be considered by applicants and licensees, including the selected RCP trip parameter, instrumentation quality and redundancy, instrumentation uncertainty, possible adverse environments,.
calculational uncertainty, potential RCP and RCP associated problems, operator training, and operating procedures.
The licensee has addressed each of the Reference 1 criteria. We have studied this information and have discussed RCP trip with the licensee's personnel.
The remainder of this Enclosure provides background pertinent to the RCP trip issue, the acceptance criteria, the basis for the discussion with the licensee, and our summary of the results of that discussion.
1
BACKGROUND TMI Action Plan Item II.K.3.5 of NUREG-0737 (Ref. 2) required all licensees to consider solutions pertinent to tripping RCPs under transient and Loss of Coolant Accident (LOCA) conditions. A summary of the industry and NRC programs concerning RCP trip was provided in SECY-82-475 (Ref. 3).
Reference 3 also provided NRC guidance and criteria for resolution of II.K.3.5, and enclosed Generic Letters 83-10 (Ref. 4).
The significant information provided by these references was summarized as follows:
"... appropriate pump trip setpoints can be developed by the industry that would not require RCP trip for those transients and accidents where forced convection circulation and pressurizer pressure control is a major aid to the operators, yet would alert the operators to trip the RCPs for those small LOCAs where continued operation or delayed trip might result in core damage."
"The resolution... is intended to ensure that for whatever mode of pump operation a licensee elects, a) a sound technical basis for that decision exists, b) the plant continues to meet the Commission's rules and regulations, and c) as a minimum, the pumps will remain running for those non-LOCA transients and accidents where forced convection cooling and pressurizer pressure control would enhance plant control.
This would include steam generator tube ruptures (SGTR) up to approximately the design basis event (one tube)."
During a small break accident in certain break size ranges, there exists a window in time during which tripping RCPs will make the accident worse.
Therefore, in a small break situation, one must trip RCPs prior to entering the window. If one wishes to depend upon manual trip, two criteria are applicable:
2
- 1. One must show that at least 2 minutes exist within which to trip RCPs following "receipt of a trip signal" using licensing calculations as a basis.
- 2. One must show that at least 10 minutes exist within which to trip PCPs following "receipt of a trip signal" using best estimate calculations as a basis.
If, for some reason, the RCPs have not been tripped within 10 minutes of the time at which plant conditions indicate trip should be performed, they are to be left running until after the window is closed. Closure can be indicated by parameters such as repaining both adequate subcooling margin and pressurizer level after they have been lost.
Analyses are required to establish timing relative to items 1 and 2, as well as to establish the dimensions of the window.
It is desirable to leave pumps running for control purposes during other transients and accidents, including steam generator tube rupture accidents of sizes up to one tube broken. Therefore, insofar as is practical, procedures and criteria should be developed to attain this goal.
Note that leaving pumps running during "non-break" transients and accidents is not a 100% requirement, as contrasted to the small break, where trip must be accomplished to remain in compliance with the regulations.
(Failure to trip as required could lead to exceeding Appendix K specified temperatures for design basis accidents.)
For "non-break" transients and accidents, RCPs may be tripped when desirable. If in doubt, the small break criteria are to be applied.
New plants coming on line should have dealt with RCP trip prior to power operation.
Note much of the work pertinent to the above criteria has been done on a generic basis, and is applicable to individual plants. Where this is the case, it is sufficient to establish applicability, and the generic work need not be repeated on a plant specific basis.
3
The WOG developed a set of three alternative RCP trip criteria to meet the intent of Reference 4 (Refs. 5 - 7):
- 1. Reactor Coolant System (RCS) pressure with normal instrument uncertainties. This criterion uses RCS pressure with normal instrument uncertainties as the criterion for RCP trip under normal containment conditions. The secondary pressure is assumed to be at the lowest secondary safety valve set pressure to provide conservatism. Instrument uncertainties associated with post-accident containment conditions are used for RCP trip under adverse containment conditions.
- 2. Reactor coolant subcooling. This method provides a direct indication for RCP trip since RCPs can continue to operate as long as sufficient subcooling margin is available. The trip criterion is established as zero subcooling in the RCS hot legs, with allowance for instrumentation uncertainties which exist for normal and abnormal containment conditions.
- 3. Secondary pressure dependent RCS pressure. This method differs from method 1 in that actual secondary side pressure is used in conjunction with RCS pressure.
(The secondary side pressure may be significantly lower than the value obtained via method 1, such as when the atmospheric dump valve is open.)
Instrumentation uncertainties are treated as in the other two methods.
A methodology was provided whereby each licensee or applicant could determine RCP trip setpoints for each of the three criteria.
Each licensee or applicant could then perform a plant-specific evaluation and could select a criterion which is best with respect to prevention of RCP trip for SGTRs and non-LOCA transients.
Overall, the staff found that for most plants, each of the criteria was adequate to provide an indication for RCP trip under small break LOCA conditions, and selection of an RCP trip criterion could be based on the capability to preclude pump trip for SGTPs and non-LOCA transients. However, the criteria may be marginal for some plants under some conditions since the 4
uncertainty analysis provided by the WOG may not be bounding for all plants.
Further, the RCS pressure set point criterion appeared to have the least potential to reduce unnecessary RCP trips. Consequently, the staff determined that each licensee or applicant must consider the instrument and calculational uncertainties when selecting a criterion, and must be prepared to explain how they were considered during future inspections.
The staff further described those plant specific items required from each licensee or applicant in order to complete the response to Generic Letter 83-10. These are reproduced below, in some cases with additional guidance (provided in Ref. 1", and define the basis for the staff review.
Organization of the sections which follow is essentially identical to that of the Generic Letter (Ref. 1) to which the licensee responded. A statement is first presented which describes the Generic Letter request. This is followed by a staff prepared summary.
INTRODUCTION The staff has completed a preliminary review of the Reference 8 submittal from the licensee pertaining to RCP trip, and has discussed the results of the preliminary review during a telephone conference call with personnel representing the licensee. This Enclosure documents both the preliminary review (Ref. 9) and the results of the discussion with the licensee's personnel.
The purpose of the review was to identify any areas where additional information was needed for the staff to complete its evaluation of PCP trip, and to provide guidance pertinent to submittal of additional information where appropriate. The telephone conference call was conducted to obtain clarification and additional information. The call took place on October 8, 1986. The licensee was represented by John Benjamin, Rick Bayton, Sam Crizer, Dale Bates, and Joe Curley. The staff was represented by Bud Requa and Warren Lyon.
5
9 9
Organization of the remainder of this Enclosure is essentially identical to that of the Generic Letter (Ref. 1) to which the licensee responded. A statement is first presented which describes the Generic Letter request. This is followed by a staff summary and critique. The staff understanding of the licensee response is then given, followed by additional staff comments where appropriate.
A.
Determination of RCP Trip Criteria Demonstrate and Justify that proposed RCP-trip setpoints are adequate for small-break LOCAs but will not cause RCP trip for other non-LOCA transients and accidents such as SGTRs.
This is to include performance of safety analyses to prove the adequacy of the setpoints.
Consider using partial or staggered RCP-trip schemes.
Staff Evaluation. Carolina Power and Light (CP&L) has selected Reactor Coolant System (RCS) subcooling as the criterion for tripping the Reactor Coolant Pumps (RCP).
The subcooling option was previously identified by the staff as the second choice of the three options described by the Westinghouse Owners Group (WOG) in their evaluation of RCP trip. The staff requests brief background information pertinent to this selection.
The quantitative value associated with subcooling that is to be used as the trip criterion is identified as 250F for normal containment conditions and 350 F for adverse containment conditions in the response pertaining to operator training. Is the staff correct in believing this is the plant criterion?
Licensee Response. The staff is correct in regard to the trip criteria.
The selection was based on the Westinghouse Owners Group (WOG) executive volume information and recommendations. Plant specific calculations eliminated use of RCS pressure as a criterion due to insufficient separation of LOCA and non-LOCA events. A pressure of less than 1?30 psi was needed, and 1300 psi was obtained. Selection of RCS to steam F
generator secondary pressure differential was rejected because this requires an operator calculation, which was judged to be an unnecessary complication since RCS subcooling met all of the criteria.
Al.
Identify the instrumentation to be used to determine the RCP trip set point, including the degree of redundance of each parameter signal needed for the criterion chosen. Establish the quality level for the instrumentation, identify the basis for the sensing instruments' design features, and identify the basis for the degree of redundance.
Staff Evaluation. CP&L has elected to use subcooling as determined by a microprocessor which receives inputs from four primary system pressure transmitters, six loop RTDs, and 16 core exit thermocouples, and which outputs a margin to saturation. The subcooling monitor is stated to possess redundancy in that it is comprised of two channels which operate completely independently. Each channel is stated to be powered from a critical instrument bus which receives power from off site or emergency diesels. Testing capability is provided via front panel test switches which test the warning lights, alarms, meter movements, and associated electronics. Redundant control grade temperature inputs are used from each hot and cold leg RTD. These are backed up by multiple core exit thermocouples. Each channel is provided with three pressure signals, one narrow range safety grade pressure and two wide range control grade pressures. All safety grade sensors are isolated from the subcooling monitor by isolation amplifiers. References are provided to further information concerning the reactor coolant subcooling monitor.
Some pressure instrumentation in some plants involves long pipe runs, typically with isolation devices and with transmitters located outside containment. Please provide a brief comment on system response time for H. B. Robinson with respect to pressure in light of the above comment.
Are procedures in place for a back up approach if the subcooling monitor is inoperative? If so, please describe.
7
Licensee Response. Control grade instrumentation is permitted since much of the applicable work was performed prior to issuance of Regulatory Guide 1.97.
Further information can be obtained in the Carolina Power and Light Company response to Supplement 1 of NUREG-0737.
Staff Reviewer Comment. No further licensee action on this item is believed necessary. We will review the referenced information.
Licensee Response (continued). All instrumentation and transmitters are located within containment.
There are two subcooling monitors. Operating procedures are in place which cover response if these are inoperative. These include calculation methods which are based upon in-core thermocouples and pressurizer pressure. Note that narrow range pressurizer pressure involves safety related instrumentation, whereas wide range pressure does not. Narrow range pressure would be used in the range of applicability, which ranges from approximately 1700 to 2500 psi.
The thermocouples will be safety grade after the next refueling.
A2.
Identify the instrumentation uncertainties for both normal and adverse containment conditions. Describe the basis for the selection of the adverse containment parameters. Address, as appropriate, local conditions, such as fluid jets or pipe whip, which might influence instrumentation reliability.
Staff Evaluation. Instrumentation uncertainty for normal and adverse containment environmental conditions is stated to be 25 0F and 35 0 respectively. References are provided for information pertinent to the determination.
A brief discussion of the basis for selection of adverse containment parameters should be provided. This should include definition of the parameters (which is not provided in the CP&L submittal).
Local conditions should be addressed, as identified in the first paragraph of A2, above.
Consideration should also be given to conditions outside containment that 8
may influence uncertainty. For example, has CP&L surveyed the wiring and connections between pressure transmitters and the control room to assure that a steam line break inside or outside of containment will not introduce problems with the pressure information received in the control room?
Licensee Response. An adverse containment condition is assumed to exist if containment pressure is greater than 4 psig.
All applicable instrumentation meets Regulatory Guide 1.97 separation criteria. The monitors are separated consistent with the requirements for safety related equipment. The two wide range pressure transmitters are widely separated, and their relative locations are similar to reactor protection instrumentation.
A3. In addressing criterion selection, provide consideration of uncertainties associated with the WOG supplied analyses values. These uncertainties are to include uncertainties in computer program results and uncertainties resulting from plant specific features not representative of the generic data group.
If a licensee or applicant determines that the WOG alternative criteria are marginal for preventing unneeded RCP trip, it is recommended that a more discriminating plant-specific procedure be developed. Licensees or applicants should take credit for all equipment (instrumentation) available to the operators for which the licensee or applicant has sufficient confidence that it will be operable during the expected conditions.
Staff Evaluation. Calculations of instrument uncertainties are summarized, and comparisons are discussed between plant data and calculations in the Westinghouse Owners Group (WOG) information.
CP9L states that the calculated overall uncertainty for H. B. Robinson is from
+1oF to +50F for the subcooling trip point.
C)
The licensed Westinghouse LOFTRAN computer code is referenced for performance of the non-LOCA analyses. The computer program result uncertainties evaluation is based on the assumption of no changes in initial plant conditions (such as full power, pressurizer level, all Safety Injection (SI) pumps running, and all Auxiliary Feed Water (AFW) pumps running). The major contributors to uncertainty are stated to be break flow rate, SI flow rate, decay heat generation rate, and AFW flow rate. Parametric studies are summarized in which the major uncertainties are stated to be due to the break flow model and SI flow inputs.
The uncertainty appears low, and is lower than submitted for any other plant the staff has reviewed. Please confirm these values and provide brief backup information to further substantiate that they are correct.
Also discuss how the values associated with the information contained in this section of the response are factored into the previously identified uncertainties in determining the RCP trip point.
The licensee has not directly addressed such topics as the accuracy of the numerical solution scheme or of nodalization. Further, there is no determination of the influence of equipment or operational failures.
Information pertinent to the former result from comparisons of the LOFTRAN code to operational and experimental data, and as a result should have been factored into the calculational basis and included in the uncertainty value. Determination of equipment or operational failures is not a necessity as long as the expected configuration of the plant is addressed since the objective of RCP trip is to provide reasonable assurance of not tripping for transients for which a trip is undesirable.
It is not necessary to establish that one will never trip unnecessarily since the plant is capable of being safely controlled if an unnecessary trip does occur. Thus, the licensee submittal is adequate with respect to these items.
Licensee Response. The uncertainty range is a Westinghouse provided value which covers the variation between the generic calculations and the Robinson Unit 2 plant.
10
The WOG performed boundina calculations and studied plant information to be certain Robinson Unit 2 was within the bounds of the generic calculations. We have reviewed the WOG work, and have determined that the information is directly applicable to Robinson Unit 2. No changes or modifications to the calculations are necessary.
B.
Potential Reactor Coolant Pump Problems B1.
Assure that containment isolation, including inadvertent isolation, will not cause problems if it occurs for non-LOCA transients and accidents.
Demonstrate that, if water services needed for RCP operations are terminated, they can be restored fast enough once a non-LOCA situation is confirmed to prevent seal damage or failure. Confirm that containment isolation with continued pump operation will not lead to seal or pump damage or failure.
Staff Evaluation. The CP&L response is that no essential services to the RCPs are lost in a Phase A containment isolation, but that some essential services are lost as a result of a Phase B containment isolation. CP&L policy is to trip RCPs if essential services are lost unless RCP operation is required to prevent core damage. RCP restart requires that essential services be available.
It is not clear from this response whether RCP trip is required in response to a Phase 8 isolation signal because it is not clear whether RCPs are tripped upon loss of some essential services or upon loss of all essential services. Please clarify.
The staff requests additional information pertinent to RCP cooling and RCP restart. For example, do the conditions which lead to containment isolation result in termination of any portion of the Chemical and Volume Control System (CVCS) and is there an indirect effect upon RCP seal injection? If CCW pump operation is terminated, how long does it take for the pump to be restarted and flow restored to the thermal barrier heat exchangers and other RCP associated components? What are the implications?
11
Information should be provided pertinent to restart of PCPs followina restoration of services leading to a trip.
Items such as trip parameters, operator response and timing of operations should be identified. The staff concerns pertain to (1) protection of the RCPs under containment isolation conditions, particularly Phase P, (2) assurance that satisfactory Reactor Coolant System (RCS) and RCP conditions exist prior to RCP restart, and (3) that restart does not damage the RCPs. The staff does not need a large volume of material on these topics, but does need a brief discussion with perhaps reference to procedures in a list provided with the submittal.
Licensee Response. No actions occur which affect RCP operation in response to a containment phase A isolation. A phase B containment isolation signal causes rerouting of the seal water return to the pressurizer relief tank as well as isolation of Component Cooling Water (CCW).
Seal injection is maintained. RCPs are tripped in response to loss of CCW, which provides cooling to both the thermal barrier heat exchangers and indirectly to other pump components. RCP operation is allowed if seal injection is lost, but CCW continues to be available.
Restart of CCW and/or seal injection after its termination is covered by plant procedures. The major concern is avoidance of thermal shock, which could damage pump components as well as could lead to seal leakage.
Staff Reviewer Comment. Discussion of RCP operation and operator actions clearly established a depth of knowledge on the part of plant personnel with respect to this topic.
B2.
Identify the components required to trip the RCPs, including relays, power supplies and breakers. Assure that RCP trip, when necessary, will occur. Exclude extended RCP operation in a voided system where pump head is more than 10% degraded unless analyses or tests can justify pump and pump-seal integrity when operating in voided systems. If necessary, as a 12
result of the location of any critical component, include the effects of adverse containment conditions on RCP trip reliability. Describe the basis for the adverse containment parameters selected.
Staff Evaluation. The major components associated with RCP trip are identified, as is their location.
A brief consideration should be given to the potential for adverse conditions outside containment and the implications, if any.
For example, can a steam line break outside of containment introduce difficulties with respect to the equipment of interest here?
The timing of operations associated with alternate operator actions required to trip the RCPs should be mentioned. For example, if the operator attempts a trip from the control room and fails, how long will it take to trip from an alternate location, including travel time?
RCP operation in a voided system is not mentioned.
Licensee Response. Separation and interaction with high energy lines is consistent with the prior discussion on these topics.
The RCP trip breakers are located near the control room, and can be readily tripped. Two of the RCPs could additionally be tripped by de-energizing buses. The third RCP could be tripped by de-energizing its respective bus, but critical equipment is also supplied from that bus.
Loss of bus voltage would result in start of diesel generators, which in turn would re-energize the bus so that critical equipment would be supplied with electrical power. The last path would not normally be used to trip RCPs.
RCP trip criteria are designed to avoid RCP operation under voided conditions. (Note there is an exception under response to inadequate core cooling conditions, where RCPs could be operated regardless of the RCS inventory condition.)
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C. Operator Training and Procedures (RCP Trip)
C1.
Describe the operator training program for RCP trip.
Include the general philosophy regarding the need to trip pumps versus the desire to keep pumps running. Also cover priorities for actions after engineered safety features actuation.
Assure that training and procedures provide direction for use of individual steam generators with and without operating RCPs.
Assume manual RCP trip does not occur earlier than two minutes after the RCP-trip set point is reached.
Determine the time available to the operator to trip the RCPs for the limiting cases if manual RCP trip is proposed. Best Estimate calculational procedures should be used. Most probable plant conditions should be identified and justified by the licensee, although NRC will accept conservative estimates in the absence of justifiable most probable conditions.
Justify that the time available to trip the RCPs is acceptable if it is less than the Draft ANSI Standard N660. If this is the case, then address the consequences if RCP trip is delayed. Also develop contingency procedures and make them available for the operator to use in case the RCPs are not tripped in the preferred time frame.
Staff Evaluation. The total CP&L response is:
"The HBR2 operators have been trained to trip the RCPs as soon as RCS subcooling reaches 250F (35oF for adverse containment conditions) during a depressurization event unless it is a planned and controlled depressurization during the longer term recovery actions. This training was accomplished during the implementation training for the new EOPs conducted in 1984. This training is reinforced during regularly scheduled simulator retraining."
14
This does not address the points requested in Item C1.
For example, the background philosophy is not described. What is the general need to trip RCPs as contrasted to keeping them running?
What is the RCP restart philosophy? If one has tripped RCPs, and desires to restart them for control purposes and to aid in plant cooldown, what requirements must be met?
Licensee Response. The background information pertinent to the need for RCP trip as contrasted to RCP operation is covered in both training and simulator operation.
If RCP trip were required in response to an event, and the operators were to miss the trip, then the RCPs would ordinarily be tripped when the mistake was identified. Plant response and necessary operator actions from that point are covered in the plant emergency operating procedures.
An additional qualifier on RCP trip is that safety injection must be available. RCPs are left running if there is no safety injection capability.
Staff Reviewer Comment. Discussion of the training and use of plant simulators clearly established an understanding of plant behavior and interaction with RCP operation.
C2.
Identify those procedures which include RCP trip related operation:
(a) RCP trip using WOG alternate criteria (b) PCP restart (c) Decay heat removal by natural circulation (d) Primary system void removal (e) Use of steam generators with and without RCPs operating (f) RCP trip for other reasons Ensure that emergency operating procedures exist for the timely restart of the RCPs when conditions warrant.
15
Staff Evaluation. CP&L has listed three procedures and references the reader to the Robinson Plant Specific Emergency Operating Procedures for further information. The staff does not consider this responsive to Item C2.
Licensee Response.. The Robinson procedures are available to the NRC staff at the plant site.
Staff Reviewer Comment. This is an acceptable response since the intent was to assure that adequate procedures exist. This was established during discussion of Item C1.
Availability of the procedures to the staff assures that further investigation can be performed at any time this is desired.
9 REFERENCES
- 1. Thompson, Hugh L. Jr., "Implementation of TMI Action Item II.K.3.5,
'Automatic Trip of Reactor Coolant Pumps' (Generic Letter No. 85-12),"
Letter from Director, Division of Licensing, NRC, to all applicants and licensees with Westinghouse (W) designed nuclear steam supply systems (NSSSs), June 28, 1985.
- 2.
"Clarification of TMI Action Plan Requirements," NUREG-0737, US NRC, November 1980.
- 3. Dircks, William J., "Staff Resolution of the Reactor Coolant Pump Trip Issue," Policy Issue for the Commissioners from the Executive Director for Operations, NRC, SECY-82-475, November 30, 1982.
- 4.
Eisenhut, Darrell G., "Resolution of TMI Action Item II.K.3.5, 'Automatic Trip of Reactor Coolant Pumps,' (GENERIC LETTER No. 83-10)," Letter to all applicants with (PWR vendor) designed nuclear steam supply systems from Director, Division of Licensing, NRC.
- 5. Westinghouse Owners Group, Letter OG-110, "Evaluation of Alternate RCP Trip Criteria," December 1, 1983.
- 6. Westinghouse Owners Group, Letter OG-117, "Justification of Manual RCP Trip for SBLOCA Events," March 9, 1984.
- 7. Westinghouse Owners Group, Letter OG-137, "Response to NRC Ouestions on RCP Trip," October 25, 1984.
- 8. Zimmerman, S. R., "H. B. Robinson Steam Electric Plant, Unit No. 2, Docket No. 50-261/License No. DPR-23, Response to TMI Action Item II.K.3.5, Generic Letter 85-12," Letter to Hugh L. Thompson, Jr., NRC, from Carolina Power & Light Company, Serial: NLS-85-335, NRC TAC #
49690, Sep 30, 1985.
- 9. Berlinger, Carl H., "RCP Trip, Robinson Unit 2," NRC Memorandum for Bud Reaua, April 3, 1986.
17
)ISTRIBUTION accket w/o end.
PAD2 Rdg w/o encl.
DMiller w/encl.
BRequa w/encl.
December 29, 1986 DOCKET NO(S).
50-261 Mr. E. E. Utley, Senior Executive Vice Presidebt Power Supply and Engineering & Construction Coaelina Power & Light Company P.O. Box 1551 Raleigh, North Carolina 28602
SUBJECT:
H. B. ROBINSON STEAM ELECTRIC PLANT,sUNIT 2 The following documents concerning our review of the subject facility are transmitted for your information.
LjNotice of Receipt of Application, dated
[ ] Draft/Final Environmental Statement, dated
[j] Notice of Availability of Draft/Final Environmental Statement, dated F-] Safety Evaluation Report, or Supplement No.
dated
[
Environmental Assessment and Finding of No Significant Impact, dated
[
Notice of Consideration of Issuance of Facility Operating License or Amendment to Facility Operating License, dated j Bi-Weekly Notice; Applications and Amendments to Operating Licenses Involving No Significant Hazards Considerations, dated 12-17-86
[see page(s)]
[ ] Exemption, dated FL1 Construction Permit No. CPPR-
, Amendment No.
dated
[ ] Facility Operating License No.
, Amendment No.
dated
[l] Order Extending Construction Completion Date, dated
[L] Monthly Operating Report for transmitted by letter dated
[L] Annual/Semi-Annual Report transmitted by letter dated Division of PWR Licensing-A Office of Nuclear Reactor Regulation
Enclosures:
As.stated cc:
See next page OFFICEO SURNAMEr DATEo OF I
RC D
P NRC FORM 318 (10/80) NRCM 0240 OFFICIAL RECORD COPY