ML19312D864

Forwards Response to 800423 Commitment to Submit Nuclear Safety Task Force Priority Items.Addresses Util Corrective Action List,B&W recommendations,NUREG-0667 & NRC 800414 Confirmatory Order

ML19312D864
Person / Time
Site:	Crystal River
Issue date:	05/02/1980
From:	Baynard P FLORIDA POWER CORP.
To:	Office of Nuclear Reactor Regulation
References
RTR-NUREG-0667, RTR-NUREG-667 IEB-79-27, NUDOCS 8005050278
Download: ML19312D864 (52)

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>El Florida Power C O N P O M A T e O ng May 2, 1980 File: 3-0-1-a Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission

• Washington, DC 20555 Subject : Crystal River Unit 3 Docket No. 50-302 Operating License No. DPR-72 Nuclear Safety Task Force Priority Items and Confi rmatory Order for Crystal River Unit 3, dated April 14,1980

Dear Sir:

In our meeting with you on April 24, 1980, Florida Pcuer Corpora-tion conmitted to submit the Nuclear Safety Task Force Priority Items and responses on or about May 2,1980. This letter fulfills that conmitment.

In addition, the Confirmatory Order for Crystal River - Unit 3, dated April 14, 1980, is addressed and is answered in final part.

The first part of the Order was completed upon the submittal of the response to I&E Bulletin 79-27 on April 25, 1980.

This submittal also addresses the FPC Corrective Action List prior to Plant Startup; B&W Recommendations; NUREG-0667, Transient Re-sponse of Babcock & Wilcox-Designated Reactors; NRC Confirmatory Order; and INP0/NSAC Recommendations.

Enclosure A reviews the purpose of the Nuclear Safety Task Force, lists the recommendations from five other organizations that were addressed, and gives a matrix to cross-reference between the Nuclear Safety Task Force Priority Items and the other five lists.

Enclosure B of this submittal gives the fifty-one (51) Priority

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items of the Nuclear Safety Task Force and their responses.

General Office 320i Thirty fourin street soutn . P O Box 14o42. st Petersburg. Florida 33733 e 813-866-5151

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Director Page Two

. Office of Nuclear Reactor Regulation May 2,1980 Enclosure C discussed the items of the other five lists from Enclosure A that are not discussed in Enclosure B.

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If you have any questions about this submittal, please contact

this office.

Sincerely, I FLORIDA POWER CORPORATION j

! Patsy Y. Baynard

Manager l Nuclear Support Services l

RMBekcF01(D1) j Attachment cc: Director i Office of Inspection and Enforcement i U.S. Nuclear Regulatory Commission

! Suite 3100

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Atlanta, GA 30303 l k

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j STATE OF FLORIDA i COUNTY OF.PINELLAS 1  !

'i Dr. P. Y. Baynard states that she is the Manager, Nuclear Support Ser-vices, of Florida Power Corporation ; that she is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission J

the information attached hereto; and that all such statements made and matters set forth therein are true and correct to the best of her knowl-edge, information, and belief.

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Subscribed and sworn to before me, a Notary Public in and for the State j) and County above named, this 2nd day of May,1980.

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Notary Public, State of Florida at Large, My Commission Expires: August 8, 1983 l

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ENCLOSURE A Followi ng the February 26, 1980 transient at Crystal River Unit 3 (CR-3), the Nuclear Safety Review Task Force was established by and to report to Mr. J. A. Hancock, Assistant Vice President of Nuclear Opera-tions, Florida Power Corporation. The purpose of the Task Force was to reexamine the adequacy of various systems, procedures and training at CR-3; identify and develop recommended changes; and recommend appropriate priorities of implementation. The priori ty of impl ementation process assures that needed corrective actions will be accouplished prior to restart of CR-3 after this refueling outage.

The Nuclear Safety Review Task Force has prepared a list of 51 priority items that must be answered prior to startup. In addition, there are five other lists of questions and concerns fran various organizations that have been developed after the formation of the Task Force. These lists are summarized below and restated in Tables A-1, A-2, A-3, A-4 and A-5. A cross reference matrix was developed to show which response to each of the 51 priority items also answers an iten in the 5 additional lists. The cross reference matrix is given in Table A-6 and the responses are discussed in Enclosure B. The items in Tables 2, A-3, and A-5, which are not addressed in Enclosure B are discussed in Enclosure C.

1. FPC Corrective Action List Prior to Plant Startup This list was the result of Request 2, specific to Crystal River Unit 3 from Harold R. Denton on March 6, 1980.

Request 2 stated " Provide a list of proposed actions at CR-3 as a result of this event". These items are listed in Table A-1 and cross-referenced to Enclosure B by Matrix Table A-6.

2. B&W Recommendations Babcock and Wilcox reviewed the past history of NNI/ICS power supply failures and developed a list of recommendations to lessen the vulnerability of the Reactor Plant to these events. These items are listed in Table A-2 and discussed in Enclosure B and C.

3. NUREG-0667, Transient Response of Babcock & Wilcox-Designed Reactors The NRC formed a B&W Reactor Transient Response Task Force to study transients and consequencas of malfunctions and failures of the Integrated Control System (ICS) and Non-Nuclear Instru-mentation (NNI). The report is NUREG-0667, Transient Response i of Babcock & Wilcox-Designed Reactors. The recommendations of l' this report are listed in Table A-3 and discussed in Enclosures B and C.

l Klein(news 280)D55

ENCLOSURE A (Continued)

4. NRC Confirmatory Order The NRC issued a Confirmatory Order for Crystal River Unit 3 on April 14, 1980. The order commitments are listed in Table A-4 and discussed in Enclosure B.

5. INP0/NSAC Recommendations The INP0 and NSAC staff completed a report on the CR-3 i ncident of February 26, 1980. The report was sent to Harold R. Denton by E. P. Wilkinson, and E. L. Zebroski on March 11, 1980. The corrective actions requested are listed

in Table A-5 and discussed in Enclosures B and C.

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TABLE A-1, FPC CORRECTIVE ACTION LIST

__ PRIOR TO PLANT STARTUP (Refer to Matrix Table A-6 for corresponding response in Enclosure B)

ITEM 1 Thorough testing of the NNI(X) system to determine cause of initial f ail ure.

ITEM 2 Review PORV circuitry to assure that credible power failures do not cause the PORV to open when it is not required to open. Review power supply independence between PORV and PORV isolation block valve to as-sure that failure which would affect PORV does not eliminate the possi-bility of PORV isolation block action.

ITEM 3 Modify Pressurizer Spray Valve so that NNI power failure will close valve.

ITEM 4 Provide positive indication of all three (3) Relief Valves.

ITEM 5 Establish procedural controls of selettable snurces for indication and control.

ITEM 6 Train all operators and I&C technicians in response to NNI and ICS fail-ures.

ITEM 7 Move 120 VAC ICS(X) power to Vital Bus.

ITEM 8 Repair Events Recorder System.

ITEM 9 Initiate a more extensive surveillance program on the Events Recorder l System.

ITEM 10 Provide operator with redundant indications of main plant parameters.

Klein(NRC5280)DN90

e TABLE A-1 (Continued)

ITEM 11 Annunciate loss of power to X and Y buses and ICS.

ITEM 12 Develop and institute a test program for changes in designs and modifi-cations.

ITEM 13 Install indicating lights on all. vital bus feeds.

ITEM 14 Modify vital bus panels for quick fuse replacement.

ITEM 15 Modify EF Pump auto start circuit and reactor trip circuit so that any power failure will not prevent activation on low SG level (control grade).

ITEM 16 PORV valve position indicating lights for solenoid will be added.

ITEM 17 Visually inspect the lower portion of the steam generator support skirts and anchor bolts. Remove any corrosive residue observed.

ITEM 18 Inspect the pressurizer heater bundles for seal leakage. Electrically check the pressurizer heater elements for continuity.

ITEM 19 A fatigue analysis will be performed on the pressurizer heaters to demonstrate that the 40-year design life was not adversely impacted by this transient.

ITEM 20 The relief valve loadings on the pressurizer relief nozzles should be determined and their effect will be assessed.

ITEM 21 The CRDMs will be checked for proper insulation resistance prior to their return to service.

Klein(NRC5260)DN90

t TABLE A-1 (Continued)

ITEM 22

a. Perform a visual inspection of the pressurizer relief system (i.e.,

PORV, both code safety valves, and the discharge piping). Inspec-tion of the discharge piping system, including hangers, should be performed to ensure that no gross distortions have occurred.

b. Confirm by calculation that the structural loads imposed on the valves and pressurizer as a result of the extended period of dis-charge to the quench tanks are acceptable.

c. Disassembly, inspection and refurbishment (as necessary) of the PORV and the code safety valves.

l ITEM 23 -

Provide diverse containment isolation.

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Klein(NRC5280)DN90 l

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1 TABLE A-2 B&W RECOMMENDATIONS

1. Following the loss of an NNI or ICS power supply, the Emergency Feedwater System shall be automatically initiated as needed and controlled to maintain proper heat removal by steam generators.

2. A failure of an NNI power supply shall not cause the PORV to open nor shall it prevent the PORV isolation valve from functioning.

3. Following the loss of an NNI or ICS power supply, the capability of stopping main feedwater to prevent excessive main feedwater addi-tion shall be available.

4. Following the loss of an NNI or ICS power supply, the capability of maintaining or restoring steam pressure in at least one steam gen-erator shall be available.

5. The failure of NNI/ICS power supplies shall not prevent safety or protection systems fron operating or prevent manual override of safety or protection systems.

6. The operator shall be trained and provided with su#ficient infonna-tion to identify a failed power supply and its related instru-ments. The remaining plant instrumentation and controls shall be sufficient to place and maintain the plant in a safe hot shutdown condition.

7. The capability to isolate letdown on loss of NNI or ICS power sup-plies shall be maintained.

8. A failure of an NNI power supply shall not cause the spray valve to open nor shall it prevent the spray block valve from functioning.

9. A loss of an NNI or ICS power supply shall not cause the pressuriz-er heater to fail on, or remain on, when the pressurizer level is low.

10. Seal injection and return (as required to prevent RCP seal damage) to the Reactor Coolant Pump shall be maintained upon an NNI/ICS power failure.

11. Review NNI/ICS systems and associated AC distribution systems fault protection requirements to minimize fault propagation.

12. Field changes to NNI/ICS systems shall be performed in accordance with a formal design control process.

13. NNI/ICS field changes should include reference (s) to identified in-stallation and maintenance precautions.

14. Each utility should perfonn a management review to assure that proper alteration and maintenance practices are in place.

Klein(NRCS280)DN90

i TABLE A-3 NUREG-0667, TRANSIENT RESPONSE OF BABC0CK & WILCOX-DESIGNED REACTORS Auxiliary Feedwater (AFW) System

1. AFW system upgrade to safety grade. ,

2. AFW system automatic initiation and control.

3. Addition of motor-driven AFW pump for Davis-Besse.

4. Modifications to steam line break detection and mitigation system.

Instrumentation and Control

5. Improvements in plant control system (ICS/NNI).

6. Selected data set of principal plant parameters for operator.

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7. Increased usage of incore thermocouples.

8. High radiation signal initiation of containment isolation.

Design and Operational Matters j

9. System response to maintain pressurizer level on scale and i pressure above HPI setpoint.

10. Sensitivity studies of operational modifications.

11. Modifications to eliminate imediate manual actions for emer-gency procedures.

12. Qualified I&C Technician on duty.

13. Operator training on Crystal River 3 event.

14. Guidelines for loss of NNI/ICS.

15. Mandatory one-week simulator training for operators as part of requalification program.

16. Evaluation of RCP restart criteria.

17. Alternative solution te PORV unreliability/ safety system chal-lenge rate concerns.

18. IREP Crystal River Study.

Klein(NRC5280)DN90

I TABLE A-3 (Continued)

General Areas for Improvement

19. Performance criteria for anticipated transients.

i 20. Continued evaluation of need to trip RCPs during small break loss-of-coolant accidents.

! 21. Reevaluate location of AFW injection into 0TSG.

22. Staff study of personnel-related LERs with respect to high j number for B&W plants.

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, TABLE A-4 4

NRC CONFIRMATORY ORDER COMMITMENTS II-1 Action which will allow the operator to cope with various com-binations of loss of instrumentation and control functions.

This includes changes in (A) equipment and control systems to give clear indications of functions which are lost or unreli-i able, (B) procedures and training to assure positive and safe manual response by the operator in the event that competent instruments are unavailable.

II-2 Determination of the effects of various combinations of loss of instrumentation and control functions by design review analysis and verification by test.

II-3 Correction of electrical deficiencies which may allow the power-operated relief valve and pressurizer spray valve to open on non-nuclear instrumentation power failures, such as the event which occurred at Crystal River Unit 3 on February 26, 1980.

In addition, this NRC Confirmatory Order required a written response to IE Bulletin 79-27 to be submitted by April 26, 1980 in accord with Florida Power Corporation's commitment. FPC submitted this response on April 25, 380.

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TABLE A-5 INP0 and NSAC RECOMMENDATIONS I. TRAINING I. A Procedural requirements for declaration of appropriate emergencies should be emphasized in plant training ses-sions.

I. B Review power supply failures and their effects on control systems. Include events such as ICS related malfunctions at Crystal River in plant training sessions and in simu-lator training.

I. C Instrument technician work practices and their potential impact on plant safety should be reviewed in plant train-ing sessions. Attention should be given to events simi-lar to the 3/20/78 and 1/5/79 transients at the Rancho Seco plant where overcooling resulted from maintenance technician actions.

II. PROCEDURES II. A Promulgate written procedures for switching instruments between power suppl ies , in the event of power supply failures and promulgate a procedure designating the pre-ferred bus for each instrument.

II. B Procedures for Steam Generator rupture matrix or its equivalent should be reviewed in conjunction with post-TMI requirements on steam-driven emergency feedwater pumps to determine if aggravating effects exist during loss of heat sink.

II. C Procedures for orderly plant shutdown following loss of power supply should be prepared or reviewed / revised as necessary. Reactor system cooldown limits, and the basis for those limits should be reviewed.

II. D The Industry should further analyze and resolve with the NRC the current reactor coolant pump trip procedures to be followed during a small break LOCA. Mandated proce-dures can be counterproductive to safety if they are not ce/ficiently discriminating to specific circumstances,

.id to specific plant designs.

II. E The Industry should review the current High Pressure In-jection pump requirements and resolve any procedural issues with the NRC. Procedures which avoid or minimize challenges to safety valves, primary system, (and even-tually to the containment building itself) are needed.

Klein(NRC5280)DN90

TABLE A-5 (Continued)

Mandated procedures can be counterproductive to safety if they are not sufficiently discriminating to specific cir-cumstances and to plant designs.

II. F Procedures for declaration of emergencies should be re.

viewed to determine if responsibility for monitoring plant conditions which lead to declaration of a specific emergency category should be assigned to a specific indi-vidual. It is suggested that this individual would also be responsible for immediately informing the senior per-son in charge at the time when these conditions for emer-gencies and emergency notification have been met. -

III. PLANT SYSTEMS AND HARDWARE The following list of problems should be investigated and corrective action taken as required.

III. A Loss of Power Supply III. A. 1 Need for backup or bus transfer capabilities if a fault trips instrumentation and control power supplies.

III. A. 2 Coupling of indication, control and computer input sig-nals, e.g., loss of power to ICS, NNI, or RCS results in loss of control board indication of many signals.

III. A. 3 PORV opening and its failure modes due to voltage varia-tion resulting in loss of proper setpoint reference.

I I I . A. 4 Susceptibility of control systems to incorrect informa-tion caused by electrical faults, e.g., choking off feed-water to steam generators, withdrawing rods, and opening the turbine throttle.

III. A. 5 Instrument loops are selected by a switch in the control room. Designs should be reviewed, and wherever practi-cal, field-tested to determine the effects of a loss of power to one of the instrument loops, and to establish the absence of cross-contamination of multiple power sup-plies in the instrument and control functions.

III. A. 6 The coincidence of having a mid-scale operating point and mid-scale instrument failure on loss of power gives un-certain information, e.g., loss of EFW auto start because steam generator level indication appeared to be higher than actual.

III. A. 7 Assignment of instruments to specific busses should in-sute as much redundancy as possible, j Klein(NRC5280)DNf 0

TABLE A-5 (Continued)

III. B Data Handling and Display III. B. 1 The adequacy of data handling and display systems should be reviewed. Examples of specific problems encountered during this event were as follows:

III. B. 1. a Many instances of alann conditions returning to a normal state without any prior indication of having reached an alana state.

III. B. 1. b Computer printout loss due to overload.

III. B. 1. c The system monitoring the in-core temperatures automati-cally orints any temperature which indicates in excess of i 700 F. The basis for selecting 700 F should be reviewed i to determine if this number should be revised, since data i was lost during the transient.

III. B. 2 Plant transient monitoring and recording. Plant trans-ient records independent of process computer, to provide a tape record of main plant parameters, are desirable for all plants. They are desirable on an earliest practic-able schedule.

Steam Generator System III. C The Steam Generator rupture matrix or equivalent should be reviewed and changed as necessary to prevent actuation of isolation and loss of heat sink for events which do not actually involve ruptures in the steam generator sys-tem.

4 Klein(NRC5280)DN90

r TABLE A-6 CR-3 CROSS REFERENCE MATRIX -

Safety Task Force FPC Corrective B&W NUREG-0667 NRC Confirm- INP0 NSAC Priority Items f.ctions for Startup Recommendations Recommendations atory Order Reconsnendations 1 1 2- 5 6 5 II-1 II.A, III.A.2 3 9 4 12 II-2 III.A.5 5 17 6 18 7 21 8 22 9 II.E 10 11 12 13 14 15 II-1 16 6 3 Sa13 11-1 I.B. II.A, II.C

'17 13 11-1 18 8 III.B.1.a 19 III.B.1.b 20 9 21 12 22 23 24 25 26 27 7 II-1 III.A.6 28 II-1 29 30 II-1 31 12 32 13 33 II-2 34 10 1&6 5/26 II-1 III.A.2, III.A.7 35 4 10 II-2 III.A.6

TABLE A-6 (Continued)

CR-3 CROSS REFERENCE MATRIX -

Safety Task Force FPC Corrective B&W NUREG-0667 NRC Confirm- INP0 NSAC Priority Items Actions for Startup Recommendations Recommendations atory Order Recommendations 36 11 III.A.1 37 38 4 39 14 40 15 1 132 III.A.6 41 16 II-3 42 19 43 20 44 23 45 182 46 2 2 II-3 III.A.3 47 3 8 II-3 48 7 5 49 11 6 5 II-1 50 13 51 4 II.B, III.C V,lein(NRCS280)DH59

ENCLOSURE B SAFETY TASK FORCE PRIORITY ITEMS The first priority areas of inquiry for the Nuclear Safety Review Task Force were the subjects of power and power supply failures and the close-coupling of the NSSS with the secondary side of the plant. The objective was to strengthen identified weak points that would otherwise significantly increase the chances of a core damage incident. The con-cern for the " man / machine interface" emphasized the viewpoint of the op-erator as a major factor in review on an equal status consideration with the hardware side. It was not intended for the Task Force to examine all safety considerations identified in the FSAR on a nonprioritized basis; however, the Task Force was not precluded from any area of in-quiry that analysis / strategy lead them.

ITEM 1 Thorough testing of the NNI(X) system to determine cause of initial failure.

RESPONSE

The following five items were checked on the NNI(X) system to determine cause of failure. Procedure PT-454 was developed to examine the system.

1. Verified auctioneering diodes and ;ower supply capability to carry bus voltage.

2. Determined AC voltage level to produce 22 VDC on bus, which trips power supply monitor - 86.8 VAC.

3. Tested to determine if S, and S2 could be closed with load on bus and power supply mom tor in service. Si and S2 time delay 0.5 seconds.

4. Interrupted AC power to St and S2 at V8DP to determine power will restore without tripping Si and S2 -

5. Visually inspected and tested all Type 820 modules in NNI(X) with 24 VDC supplies.

The first four items were completed by March 2,1980, and all results were satisfactory.

The following is the results of Item 5:

1. A complete module-by-module check was made of all NNI(X) modules.

2. On Wednesday, March 5,1980, a buffer module was found in the NNI(X) cabinets with the pin connectors touching between the

+24 VDC and ground. The land on the printed circuit board was burned away.

Klein(NRC5280)DN51

ENCLOSURE B (Continued)

ITEM 1 (Cont'd)

RESPONSE

3. The actual module that failed was installed on February 15, 1980, and was associated with the Psat-Tsay' Monitors. An in-vestigation revealed that the failure in "X power did not im-pair the function of the redundant channel in the "Y" cabi-nets, nor did any non-NNI Tsat equipment contribute to the incident.

4. All bulfer modules in the "Y" cabinets were also inspected for any problems.

ITEM 2 Establish procedural controls of selectable sources for indication and control.

RESPONSE

Our investigation to-date of the February 26, 1980, transient at CR-3 indicates a need for the following procedural changes which will be im-plemented prior to restart:

1. Emergency or Abnormal Procedures (Including recogniti'on and response)

a. Loss of vital bus power to a non-nuclear instrumentation bus,

b. Loss of vital bus power to integrated control system.

2. Surveillance Procedures

a. Events Recorder System.

b. Instrument systems power supply and function switch positions.

3. Functional Test Procedures

a. Subcooling monitor.

b. Redundant instrument availability.

I ITEM 3 Initiate a more extensive surveillance program on the Events Recorder System.

l Klein(NRCS280)DN51  !

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ENCLOSURE B (Continued)

ITEM 3 (Cont'd)

RESPONSE

A surveillance procedure, SP-505, has been developed to perform a peri-odic functional check of the Events Recorder / Annunciator System and will be implemented before startup.

ITEM 4 Develop and institute a test program for changes in designs and modifications.

RESPONSE

The development and institution of a test program for changes in designs and modifications requested in Item 12 of the FPC Corrective Action List Prior to Plant Start is governed by existing Compliance Procedure CP-114, " Procedure for Preparation and Control of Permanent Modifica-tions, Temporary Modifications, Deviations, and MAR Functional Test Pro-cedures." CP-114 will be the controlling document in developing and performing the functional test for redundant instrument modifications.

MAR 80-3-64 is the specific MAR package detailing the modifications to be made on the redundant instruments for major plant parameters.

Specific test procedures will be developed, issued, and approved as part of the construction work package authorizing and detailing the installa-tion once the system modifications / redesigns have been finalized.

ITEM 5 Visually inspect the lower por? .on of the steam generator support skirts and anchor bolts. Remove any corrosive residue observed.

RESPONSE

Visual inspection and cleaning of the lower portion of the steam genera-tor support skirts and anchor bolts was completed on April 12, 1980.

ITEM 6 Inspect the pressurizer heater bundles for seal leakage. Electrically check the pressurizer heater elements for continuity.

RESPONSE

Visual inspection of the pressurizer heater bundles for seal leakage was completed April 12, 1980. The heater element continuity checks were completed April 29, 1980. Nu problems were found.

Klein(NRC5280)DN51

ENCLOSURE B (Continued)

ITEM 7 The CRDMs will be checked for proper insulation resistance prior to their return to service.

RESPONSE

The insulation resistance of the CRDMs has been checked and found to be within specifications.

ITEM 8

a. Perform a visual inspection of the pressurizer relief system (i.e.,

PORV, both code safety valves, and the discharge piping). Inspec-tion of the discharge piping system, including hangers, should be perfomed to ensure that no gross distortions have occurred.

b. Confirm by calculation that the structural loads imposed on the valves and pressurizer as a result of the extended period of dis-charge to the querich tanks are acceptable.

c. Disassembly, inspection, and refurbishment (as necessary) of the PORV and the code safety valves.

RESPONSE

a. The pressurizer relief system including hangers for the discharge piping system was visually inspected and no gross distortions were detected. Work was completed on April 29, 1980.

b. Calculations have been performed by GAI which demonstrated that structural loads on the valves snd pressurizer were within accept-able limits.

c. The disassembly. inspection and necessary refurbishment of the PORV and code safety valves is in progress. The PORV is presently being rebuilt and the replacement of the code safety valves is complete and are being checked as of April 30, 1980.

ITEM 9 Review procedures covering when HPI flow can be cut back and secured during a small break or overcoolmg transient.

RESPONSE

Modifications to the small break and non-LOCA overcooling transient procedures will be completed to only require 20 F subcooling. (A Tsat meter exists with digital display for ease of verification.) The guidelines will be submitted to the NRC for approval prior to implementation.

l Klein(NRC5280)DN51 1

ENCLOSURE B (Continued)

ITEM 10 Review procedures covering OTSG tube rupture in accordance with revised B&W Tube Rupture Guidelines and Small Break Guidelines.

RESPONSE

Interim draft guidance on handling a steam generator tube rupture was prepared and provided to Crystal River. A key item in the recommenda-tion section is early recognition and identification of a steam genera-*

tor tube rupture incident.

The final guidelines on handling a steam generator tube rupture will be submitted to the NRC for approval prior to implementation.

ITEM 11 Review procedures concerning proper OTSG tcvel at HPI and manual RCP Trip in accordance with revised B&W Small Break Guidelines.

RESPONSE

For overcooling conditions, the Small Break Guidelines will allow an RCP to be restarted in each loop if the required conditions are met. The operation of the RCP removes the OTSG level requirement for natural cir-culation. The CR-3 Small Break Guidelines have similar requirements as ANO-1.

The Emergency Procedures and Abnormal Procedure will be revised to pro-vide the operator with a means of making a decision on what is causing ESFAS actuation: either a LOCA or a overcooling transient.

The revised Small Break Guide will be submitted to the NRC for approval prior to implementation.

ITEM 12 Review procedures concerning when to initiate HPI cooling upon total loss of secondary heat removal capability.

RESPONSE

Installation of EFW flow indicators was accomplished during the Summer 1979 outage. Should a LOFW include the EFW, as seen on the EFW flow in-dicators, and EFW is not capable of being started, then HPI cooling should be initiated. However, if OTSG level is above the level limit and steam pressure is greater than 600 psi then efforts should be made

! to manually start EFW. However, HPI cooling should only be used if EFW j and MFW are lost and not regainable.

Klein(NRC5280)DN51 l - .

ENCLOSURE B (Continued)

ITEM 12 (Cont'd)

RESPONSE

The operator will be given guidelines as to what OTSG pressure and level can be before HPI cooling is initiated. Prior to this time, he will be instructed to attempt to start EFW.

ITEM 13 Provide procedures and training for recovery from EFW actuation to avoid 0TSG overfill.

RESPONSE

A section entitled " Recovery from Emergency Feedwater Actuation" is be-ing incorporated into EP-108 " Loss of Steam Generator Feed" and will be completed before startup.

Emergency Procedure EP-113 " Plant Shutdown from Outside Control Room" will be revised to include adequate operator verification of proper feedwater water system status prior to leaving the control room.

Emergency Procedure EP-105 " Steam Supply Rupture" will be revised to state that Main Steam Isolation Valves will not automatically open.

Also, new valve closing instructions will be developed for the operator.

Emergency Procedure EP-101 " Unit Blackout" required changes are covered in Item 30.

Abnomal Procedure AP-112 " Loss of Electrical Supplies" will be revised to recognize loss of the startup transformers as unit blackout.

Emergency Procedure EP-108 " Loss of Steam Generator Feed" will be re-vised to verify feedwater valve status earlier in the response.

ITEM 14 Establish mirimum conditions for voluntarily entering degraded modes of operation.

RESPONSE

Administrative Instruction 500 " Conduct of Operations" has been revised j to provide the following guidelines:

l l 1. Plant stable and under control within existing equipment, l procedures and personnel capability.

2. Surveillance of redundant equipment will be determined to be l operable before removing the degraded equipment from service.

Klein(NRC5280)DN51

ENCLOSURE B (Continued)

ITEM 14 (Cont'd)

RESPONSE

3. Additional compensating measures shall be considered (i.e.,

dedicated operator, etc.).

4. Any equipment out of service that causes entry into an action statement of Technical Specifications shall be worked around the clock with all resources necessary to repair the equipment and place it back in service in the shortest possible time.

5. If an emergency diesel is to be taken out of service, no equipment in the opposite Emergency Safeguards train can be out of service.

6. If "A" emergency diesel is taken out of service, the steam-driven EF pump shall be determined to be operable.

ITEM 15 Revise procedures to require ICS Rod withdrawal inhibit reset upon RPS reset.

RESPONSE

Procedures will be revised to require ICS Rod withdrawal inhibit reset upon RPS reset prior to restart.

ITEM 16 Train all operators and I&C technicians in response to NNI and ICS failures.

RESPONSE

The Training Department at CR-3 will work with engineering to develop a course to be presented to all licensed personnel and technicians. The course will include as a minimum:

. A review of NNI/ICS, including all proposed modificaions.

. A review of what indications are available to the operator during system upsets and NNI/ICS power losses.

. A review of control system interactions caused by NNI/ICS pow-er losses.

. A review of Emergency Procedures and Abnormal Procedures ne-cessary to shut down the plant with emphasis on how to regain manual control during NNI/ICS power losses.

Klein(NRC5280)DN51

ENCLOSURE B (Continued)

ITEM 16 (Cont'd)

RESPONSE

. A review of all NNI/ICS event trees developed by the Nuclear Safety Task Force.

ITEM 17 FPC shall train operators on CR-3 sequence of events, concentrating on ICS response to failed NNI and how lessons learned from TMI-2 affected transient and on plant changes that are being made as a result of the CR-3 event.

RESPONSE

The Training Department at CR-3 will work with engineering to develop a course to be presented to all licensed operators. This course will in-clude as a minimum:

. Training on all modifications that result from the 2/26/80 in-cident.

. Training on how modifications will affect plant response and control.

. Training on all procedure changes that occur as a result of these modifications.

. The operators will also be 6 rained ua ine sequence of events, ICS response to failed NNI, and how lessons learned from TMI-2 affected the transient.

ITEM 18 Repair Events Recorder System.

RESPONSE

Analysis of the Events Recorder System following the February 26, 1980 trip of CR-3, revealed that if more than 16 events occurred simultane-ously, everything beyond 16 was lost. Subsequent troubleshooting isola-ted the cause of this to a defective Sequential Memory module connector l

problem. The problem has been corrected and work initiated to install l

equipment necessary for periodically exercising the full capabilities of l the Sequential Memory circuitry. Surveillance Procedure SP-505 refer-l enced in Item 3 will be used to check the operability of the Events Recorder.

Klein(NRC5280)DN51

ENCLOSURE B (Continued)

ITEM 19 Review the present design of computer alarm printout to help eliminate overload and printout time delay.

RESPONSE

Three deficiencies were considered in the responsiveness of the plant computer system during plant transients:

1. Lack of timeliness with which alarms are printed.

2. Large number of useless alarms.

3. Probability of completely filling the alarm buffer during longer plant transients.

The importance of the computer as an alarming device for situations re-quiring prompt action was weighed against the need for recording as much data as possible for later analysis.

It was decided that the alarm printout is more important as a data log-ger than as an operator tool during transients. Thus, Part 2 above is insignificant and improvements need only be made in the area of Parts 1 and 3. Therefore the system program has been modified so that all alarms will be automatically diverted to the line printer upon plant trip.

ITEM 20 Adjust secondary steam relief valve blowdown settings.

RESPONSE

B&W analysis and changes in the main steam line relief valve settings at the Davis-Besse Plant have shown that a less than 5% blowdown following a reactor trip can significantly increase the probability of the pres-surizer level remaining on scale. Therefore, the main steam line relief valve settings at Crystal River will be modified to obtain less than 5%

blowdown following a reactor trip in an effort to maintain the pressur-l izer level on scale.

i ITEM 21 Provide around-the-clock I&C technician coverage.

RESPONSE

l Around-the-clock I&C coverage will be established as soon as additional l people can be hired and trained. We are looking for I&C technicians for the special maintenance crew. Until the special maintenance crew can be formed, we will use the premium payment clause of the contract and the regular crew. It is planned to have all shifts covered when we start up, if possible.

Klein(NRC5280)DN51

O ENCLOSURE B (Continued)

ITEM 22 Perform corrective action regarding potential shorting in safety system from improperly installed fiber clamp.

RESPONSE

This item will be completed before startup following the current refuel-ing outage.

ITEM 23 Provide temporary backup air system for main feedwater startup control valves.

RESPONSE

A portable diesel-driven air compressor will be utilized as a temporary backup air system for main feedwater startup control valves until an evaluation can be made for a long-term upgrade. This will provide greater instrument air system reliability for controlling of MFW and EFW valves during a loss of offsite power.

ITEM 24 Review and revise procedures for irmediate operator action to trip DH pumps upon spurious closing of the DH dropline valves.

RESPONSE

The Decay Heat pumps will be tripped anytime their suction supply valves are not in the open position. This will minimize pump damage due to cavitation. Operating Procedure OP-404 and Emergency Procedure EP-112 are being revised to provide these procedures.

ITEM 25 FPC management shall perfonn an evaluation of the number of exempt per-sonnel that should hold operators licenses.

RESPONSE

The concern is a walkout of Bargaining Unit licensed operators during an incident, and management having to operate the plant. Since this is a concern whenever union contract negotiations occur, an evaluation has been completed. The results are:

1. To operate the plant without Bargaining Unit personnel re-quired 3 shifts of 8 operators of which one must have an SR0 Klein(NRC5280)DN51

ENCLOSURE B (Continued)

ITEM 25 (Cont'd)

RESPONSE

and 2 must have an R0 lice:.se (i.e., a total of 9 licensed exempt personnel).

2. Exempt licensed personnel available:

7 Shift Supervisors 6 Asst. Shift Supervisors 1 Operations Superintendent 1 Operations Engineer 1 Manager of Training 3 Training Instructors 1 Maintenance Superintendent 1 Maintenance Staff Engineer 1 Maintenance Planner 1 Technical Services Superintendents 21 Total (which is inclusive of the compliment required in 1 above)

3. There are approximately 15 exempt maintenance and technical support personnel that are not licensed but are capable of performing nonlicensed operator duties.

ITEM 26 Establish administrative controls to minimize access to containment in Mode 1.

RESPONSE

The procedure for administrative control of containment entry will be revised prior to startup.

ITEM 27 Include in operator training and plant procedures methods of isolating letdown and makeup in the event of loss of ICS or NNI power supplies.

RESPONSE

The operator can isolate letdown from the control room by closure of makeup system valve MUV-49 or valves MVV-40 and MUV-41. These valves are the system containment isolation valves and are powered by Engi-neered Safeguard power which is independent of NNI and ICS power j supplies.

Operators will be advised by training and procedure that these valves

! are available to isolate letdown in the event of a loss of NNI or ICS power supplies.

Klein(NRC5280)DN90 L.

l l

ENCLOSURE B (Continued)

ITEM 28 Provide capability to facilitate operator action in event of loss of power to the ICS which results in a spurious interlock precluding re-start of reactor coolant pumps.

RESPONSE

Loss of power to the ICS causes the ICS " Reactor Power Greater Than 22%" relay to operate, which is a start permissive interlock for the reactor coolant pumps.

The RCP start logic will be modified to provide bypass of permissive in-terlocks la an emergency condition. This bypass will be under adminis-trative control with a key lock switch for each pump and will be located on the main control board.

ITEM 29 Visually inspect AFW nozzle collars and repair as required.

RESPONSE

The AFW nozzle collars inspection and repair was completed on April 29, 1980.

ITEM 30 Review restart of critical items on loss of offsite power without ESFAS actuation and revise applicable procedures.

RESPONSE

We are presently performing a review to determine which critical items are affected by a loss of offsite power without ESFAS actuation. This review will be completed and all necessary modifications and procedure changes will be implemented prior to startup.

ITEM 31 Field changes to NNI/ICS Systems should be performed in accordance with design control requirements.

RESPONSE

FPC will develop and/or revise procedures prior to restart to implement this item as part of the design control system. FPC presently has a design control system for modifications to safety-related systems. The NNI/ICS are not safety-related systems per our definition. However, due to their importance, we are including these systems in our design control program. This revision to our program will insure adequate design review and control over modifications to this system consistent with safety-related changes that are made at CR-3.

Klein(NRC5280)DN51

ENCLOSURE B (Continued)

ITEM 32 NNI/ICS changes should include specific reference (s) to installation and maintenance precautions identified by the equipment supplier.

RESPONSE

FPC will develop procedures prior to restart to implement this item as part of the design control system.

ITEM 33 Check 820 signal monitor output for seal-in problems.

RESPONSE

A Bailey 820 signal monitor output seal-in caused the PORV to open in the February 26 incident. We are reviewing other NNI/ICS circuits to determine if similar problems exist and implementing corrective action as required prior to startup.

ITEM 34 Provide operator with redundant indications of main plant parameters.

RESPONSE

The main plant parameter NNI system instrumentation consists of signals originated at 120 VAC-powered transmitters, signal conditioning by 24 VDC-powered electronic modules, and display by 120 VAC-powered indi-cators or rcorders. NNI(X) and NNI(Y) supply all thre. types of power.

The signal conditioning modules were mixed in instances between NNI(X) and NNI(Y). These vital instrument loops will be revised to achieve separation from the transmitter through conditioning to new indicators.

These new indicators are self-powered and require no 120 VAC for opera-tion and are an addition to existing main control board (MCB) indicators.

The new indicators with NNI "X" and "Y" power availability lights will be located on the ICS portion of the MCB above the existing annunciator windows. Attachment 1 details location of these new indicators and NNI instrument loop design criteria to assure the availability of vital in-dication in the event of a power failure.

ITEM 35 Provide override closure of atmospheric dump valves upon loss of ICS power.

i Klein(NRC5280)DN51

ENCLOSURE B (Continued)

ITEM 35 (Cont'd)

RESPONSE

Upon loss of ICS power, the atmospheric dump valves fail mid-position.

A circuit modification will be made prior to startup to provide automat-ic closure of these valves upon loss of ICS power to preclude uncontrol-led secondary side blowdown. This new circuitry will, upon loss of ICS power, automatically vent control air to these pneumatic-operated valves causing valve closure.

ITEM 36 Provide automatic bus transfer switches for NNI (X) and ICS AC power, normally on a vital bus with auto transfer to regulated instrument bus.

Dual 24 VDC supplies will be powered from a vital bus and regulated in-strument bus.

RESPONSE

Loss of 120 VAC power causes both indication and control losses in the NNI/ICS systems. We will install automatic transfer switches for the NNI (X) and ICS AC power with the vital bus as the preferred supply and transfer to the regulated instrument bus upon loss of vital bus fail-ure. See Attachment 2 for further clarification.

ITEM 37 Provide subcooling monitors with reliable backup. One monitor shall be operable on loss of either inverter A, B, C, or D or offsite power.

RESPONSE

Dual subcooling monitors at CR-3 provide redundant indication of satura-tion margin. Loss of power due to either an inverter or offsite failure would not cause loss of both subcooling meters. Our subcooling monitor power and signal sources have been verified to provide this reliability.

Details concerning the design of the subcooling nonitors at CR-3 were provided in our January 11, 1980 response regarding NUREG-0578.

ITEM 38 Provide positive indication of all three pressurizer relief valves.

j RESPONSE l

I In direct response to NUREG-0578, Item 2.1.3.a, FPC has purchased from i

Babcock & Wilcox a Valve Monitoring System. This system incorporates acoustical nonitoring techniques to provide the reactor operator with indication of flow through the valve. The equipment is very similar to the existing Loose Parts Monitoring System supplied by Babcock & Wilcox.

Klein(NRC5280)DN51

ENCLOSURE B (Continued)

ITEM 38 (Cont'd)

RESPONSE

This design provides for two transducers mounted on each safety valve and the PORV. Each of these transducers will be wired from the contain-ment to the PORV/Tsat monitoring cabinet, located ir the 4160 V SWGR Room. Within this cabinet will be three channels (one for each valve) of signal conditioning with local indication, alarm, and selectable au-dio monitor. Only one transducer will be normally monitored on each valve. The other is manually selectable for comparison of performance or in the event of transducer failure. Each channel will also provide remote analog indication and annunciator events recorder high alann functions. This analog indicator for each channel will be mounted on the ICS section of the main control board. A connon annunciator window will also be located on this section. The events recorder will provide CRT and hard copy indication of valves that actuate.

The valve monitoring /Tsat cabinet will be powered from a vital source with all cable routing meeting seismic requirements. Florida Power Cor-por' tion is participating with Babcock & Wilcox in a generic program to environmentally qualify the valve nonitoring equipment. This program is presently scheduled to be completed by October,1980.

ITEM 39 Modify vital bus panels for quick fuse replacement.

RESPONSE

The replacement of any vital bus fuse requires the removal of a panel front including the access door to the breaker. This will be revised such that this entire panel front is hinged for quick access for replacement.

ITEM 40 Modify EF pump auto start circuit and reactor trip circuit so that any power failure will not prevent actuation on low SG level (control grade).

RESPONSE

The control grade emergency feedwater pump auto start and anticipatory reactor trip on low-low steam generator level will be modified to pro-vide Auto Start and reactor trip during any single power failure.

ITEM 41 l PORY valve position indicating lights for solenoid will be added.

l Klein(NRC5280)DN51 i

ENCLOSURE B (Continued)

ITEM 41 (Cont'd)

RESPONSE

PORV position indicating lights indicate selector switch position, high pressure setpoint select, and low pressure setpoint select. These indi-cating lights will be modified to provide indication that opening has been commanded. This, combined with the new acoustical monitors, and the existing pressure setpoint indication will provide the operator with the capability to cross check indications for verification of PORV position.

ITEM 42 A fatigue analysis will be performed on the pressurizer heaters to demonstrate that the 40-year design life was not adversely impacted by this treatment.

RESPONSE

During the February 26, 1980 transient experienced at Crystal River 3, it was postulated that some of the pressurizer heaters may have been ex-posed to a saturated steam environment and concurrently energized (Ref.

Transient Assessment Report). B&W reconnended that the heaters be elec-trically checked for continuity, the bundles inspected for possible leakage and a 40 year fatigue analysis performed. Althougth it was not the intent of B&W, FPC committed to the NRC to perform the fatigue anal-ysis prior to restart. B&W now believes this fatigue analysis is not necessary because it has been confirmed that the heater bundles were not uncovered. Therefore, Florida Power Corporation will not pursue this analysis any further.

ITEM 43 The relief valve loadings on the pressurizer relief nozzles should be determined and their effect will be assessed.

RESPONSE

This has been accomplished by the GAI calculations (see Item 8) which demonstrated that piping loads were within design considerations and as such require no further analysis of pressurizer relief nozzles.

ITEM 44 Provide diverse containment isolation.

Klein(NRC5280)DN51

ENCLOSURE B (Continued)

ITEM 44 (Cont'd)

RESPONSE

Florida Power Corporation, in its April 12, 1979 response to Item 6 of IE Bulletin 79-05A, identified essential and nonessential systems with regard to containment isolation and core cooling. Essential systems were defined as those systems which are required for core cooling capa-bility and, therefore, should not be automatically isolated on HPI actu-ation. For the valves listed in our April 12 response, which receive no ES signal and are normally closed and remain closed following the acci-dent conditions, no further action is required.

The nonessential valves, listed in our response, which receive a con-tainment isolation signal (4 psig RB pressure) will be provided with a diverse containment isolation parameter by the addition of an auto-close isolation signal, based on automatic HPI actuation. This diverse con-tainment isolation signal will satisfy safety-grade requirements and re-setting of HPI will not result in the automatic loss of containment isolation.

ITEM 45 Electrically interlock motor-driven EFW pump to start on loss of MFW.

RESPONSE

The motor-driven EFW pump is presently interlocked to start on loss of main feedwater provided that offsite power is available. The emergency diesel generator block loading sequence will be modified to include the auto start of the motcr-driven emergency feedwater pump when offsite power it not available.

Details of this design change are included in our May 2,1980 submittal to Mr. Robert Reid of the NRC staff.

ITEM 46 Review PORV circuitry to assure that credible power failures do not cause the PORV to open when it is not required to open. Review power supply independence between PORV and PORV isolation block valve to assure that failure which would affect PORV does not eliminate the pos-sibility of PORV isolation block action.

RESPONSE

The PORV circuitry will be modified to include a loss of NNI power interlock to interrupt 125 VDC control power to the PORV solenoid. Loss l

Klein(NRC5280)DN51 l

ENCLOSURE B (Continued)

ITEM 46 (Cont'd)

RESPONSE

of either NNI(X) 120 VAC or 24 VDC will cause the auxiliary relay to de-energize and block the automatic opening circuit of the PORV.

The power source to the PORV and block valve are independent. The power supply to the PORV solenoid is 125 VDC, while the block valve is a motor-operated valve, fed from a 480 VAC Engineered Safeguards motor control center.

The control switch for the PORV will be relocated to the Main Control Board. This switch will have auto (NNI control), open and close posi-tions. Positions override NNI commands and do not rely on NNI power.

ITEM 47 Modify Pressurizer Spray valve so that NNI power failure will close the valve.

RESPONSE

The Pressurizer Spray Valve will be modified to include an interlock from the auxiliary relay described above for the PORV. Loss of NNI(X)

AC or DC power will cause the spray valve to close. The operator will be able to manually operate the valve from the main control board to ei-ther open or close the valve. Failure of the NNI(X) power will not pre-vent manual control.

ITEM 48 Move 120 VAC ICS(X) power to Vital Bus.

RESPONSE

The ICS(X) 120 VAC power supply is presently fed from a regulated power l source. A change to incorporate an Automatic Transfer Switch (ATS) be-l tween a vital bus power source and a regulated source is being providad.

This scheme provides reliable power to the ICS (see Item 36 for further clarification of this modification).

ITEM 49 Annunciate loss of power to NNI(X) and NNI(Y) buses and ICS.

Klein(NRC5280)DN51 m _

ENCLOSURE B (Continued)

ITEM 49 (Cont'd)

RESPONSE

There were no provisions to independently annunciate the loss of ICS, NNI(X) or NNI(Y) 120 VAC and 24 VDC power sources. Separate annunciator windows will be provided to inform the operator of a loss of the NNI(X),

NNI(Y), or ICS power sources. In addition, power available indicating lights are provided on the main control board adiacent to the added in-dicators such that the absence of a light indicates the adjacent indica-tors are without power.

ITEM 50 Install indicating lights on all vital bus feeds.

RESPONSE

An indicating light for each branch circuit supplied from safety-related vital distribution panels will be provided. The indicating lamps and isolating fuses will be installed in a box which will be mounted near the distribution panels and easily visible to control room personnel .

There is a separate box for each distribution panel.

ITEM 51 Evaluate the OTSG Rupture Matrix with the intent to remove the signal from FWV-161 and FWV-162 to assure a 7assive EF flow path to both OTSGs on initiation of EFW.

RESPONSE

The CR-3 s tea,a line rupture matrix presently isolates FWV-161 and FWV-162 upon actuation. Concern is that this isolation of emergency feed-water mai t be in the best interest of overall plant safety. This concern 1. ased upon the higher industry operating experiences of actu-ation due to generator dry-out or RC system cooldown rather than a steam line rupture.

The protection afforded by isolation of EFW should be evaluated against the increased reliability of EFW by not isolating EFW. This evaluation is being performed by GAI and B&W and will be submitted for NRC review upon completion. If the evaluation supports not isolating EFW via the rupture matrix, a minor control change will remove these valves from the steam line rupture matrix actuation logic.

Klein(NRC5280)DN51

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ENCLOSURE C The response to items in the B&W Recommendations, NUREG-0667 Recommenda-tions, and INP0/NSAC Recommendations that were not addressed by the Nuclear Safety Task Force Priority Items (Enclosure B) are discussed in this enclosure.

Klein(NRC5280)DN90

B&W RECOMMENDATIONS NOT COVERED ON TASK FORCE LIST ITEM 5 The failure of NNI/ICS power supplies shall not prevent safety or pro-tection systems from operating or prevent manual override of safety or protection systems.

RESPONSE

The NNI/ICS supplies no signals to the RPS or ES systems, nor can ICS or NNI failures affect ES through signals supplied to the NNI. Therefore, this concern does not apply to CR-3. Additionally, there is no RPS override. The ESF channel override has no connection with the NN!/ICS systems.

ITEM 9 A loss of an NNI or ICS power supply shall not cause the pressurizer heater to fail on, or remain on, when the pressurizer level is low.

RESPONSE

Loss of NNI power could cause heater groups 7 through 13 to come on.

This is not considered to be a safety concern and as such not a startup restraint. The problem of heater burnout on low pressurizer level has been evaluated, and it was determined that heater well failure does not occur before the heating element fails and terminates the temperature increase. Therefore RC system integrity will not be unacceptably degraded during this incident.

ITEM 10 5eal injection and return (as required to prevent RCP seal damage) to the Reactor Coolant Pump shall be maintained upon an NNI/ICS power failure.

RESPONSE

This item requires additional B&W study before implementation and is not considered to be an immediate safety concern. Therefore, this is a long term consideration and not a startup restraint.

ITEM 14 Each utility should perform a management review to assure that proper alteration and maintenance practices are in place.

RESPONSE

The Nuclear Safety Review Task Force has recommended the Nuclear General Review Committee of FPC direct a review / audit of existing procedures concerning change and maintenance practices.

Klein(NRC5280)DN90

1 B&W RECOMMENDATIONS NOT COVERED ON TASK FORCE LIST ITEM 14 (Cont'd)

RESPONSE

Since the Quality Programs Department (QPD) personnel are familiar with the use of the existing procedures, they will perform this audit for the NGRC. The Quality Programs Department will submit a final report to the NGRC for committee action. From this QPD Report, the NGRC will escer-tain key areas that could be improved or corrected.

Klein(NRC5280)DN90

l NUREG-0667 RECOMMENDATIONS NOT COVERED ON THE TASK FORCE LIST l ITEM 3 i Addition of a motor-driven AFW pump for Davis-Besse.

RESPONSE

This recommendation does not apply to CR-3. t l ITEM 7 Increased usage of incore thermocouples:

RESPONSE

l Each Tsat meter, as presently installed, has two hot leg and two cold leg temperature inputs and a pressure input from each loop.

In addition to the above temperature inputs, FPC is also providing, in-dependently, the hottest of 5 core exit T/Cs to each Tsat meter. These T/Cs are selected from each core quadrant and the central region. This addition will give the operator the capability to selectively observe saturation margin of the hottest T/C, Th or Tc against pressure in eith-er loop. The alarm on reduced saturation margin will be from the not-test temperature input and selected loop pressure. Installation of the modification is scheduled to be completed prior to restart.

CR-3 also has automatic printout of all incore thermocouples on activa-tion of high temperature setpoint. FPC is working on a modification to automatically initiate this printout on reactor trip.

ITEM 8 l High radiation signal initiation of containment isolation.

RESPONSE

1 Containment purge isolation at CR-3 is presently actuated upon high rad- l iation, 4 psig RB pressure or HPI actuation. The RB pressure and HPI actuation systems are redundant and safety-grade. FPC is presently per-forming an evaluation of the purge systen at CR-3 in response to NRC I questions contained in Mr. Reid's letter of February 29, 1980. Upon completion of our evaluation, FPC will identify any modifications neces-sary to permit continuous purging at CR-3.

l l,

Klein(NRC5280)DN55

NUREG-0667 RECOMMENDATIONS NOT C0VERED ON THE TASK FORCE LIST (Continued)

ITEM 11 Modifications to eliminate immediate manual actions for emergency procedures.

_ RESPONSE The design philosophy of CR-3 relies on well-trained, intelligent opera-tors as the most effective means of responding to unforeseen situa-tions. Where the time requirement for operator action is too short, automatic activation modifications have been made. A long term study is required in order to establish additional modification requirements.

ITEM 14 Guidelines for loss of NNI/ICS.

RESPONSE

This item requires B&W evaluation. However, FPC is developing plant procedures for the operators to follow in the event of NNI/ICS losses.

Training in these procedures will be completed prior to startup after this refueling outage.

ITEM 15 Mandatory one-week simulator training for operators as part of requali-fication program.

RESPONSE

This recomendation was made a part of the requalification program at Crystal River Unit 3 in 1979.

ITEM 16 Evaluation of RCP restart criteria.

RESPONSE

This item was referred to the NRC staff for evaluation.

Klein(NRC5280)D55

NUREG-0667 RECOMMENDATIONS NOT COVERED ON THE TASK FORCE LIST i (Continued)

ITEM 17 Alternative solution to PORY unreliability/ safety system challenge rate Concerns.

RESPONSE

This item was referred to the NRC staff for evaluation.

ITEM 18 IREP Crystal River study.

RESPONSE

This item was referred to the NRC staff ror evaluation.

ITEM 19 Performance criteria for anticipated transients.

RESPONSE

This item was referred to the NRC staff for evaluation.

ITEM 20 Continued evaluation of need to trip RCPs during small break loss-of-coolant accidents.

RESPONSE

This item was referred to the NRC staff for evaluation.

ITEM 21 Reevaluate location of AFW injection into OTSG.

RESPONSE

This item requires B&W evaluation and is a long term study.

Klein(NRC5280)DN55

. e t

i NUREG-0667 RECOMMENDATIONS NOT C0VERED ON THE TASK FORCE LIST (Continued)

ITEM 22 Staff study of personnel-related LERs with respect to high number for j B&W plants.

RESPONSE

This item was referred to the NRC staff for evaluation.

An assessment of licensed operator errors was completed in support of the Crystal River - Unit 3 Nuclear Safety Task Force by P. E. Dietz, Institute for Nuclear Power Operations, in his May 1,1980, letter to Dr. P. Y. Baynard.

Recommendation 22 of the DRAFT NUREG-0667, April 2, 1980, Transient Response of Babcock and Wilcox -

Designed Reactors, urges the performance of an analysis of the number of licensee event reports attributed to licensed personnel error to determine the significance and cause of the higher number associated with the operation of B&W 2 facilities. As noted in the report, LERs have only been categorized by licensed personnel error since January 1978.

The following table is from the NRC report and covers LERs between January 1978 and January 1980.

TABLE 1 LER OUTPUT ON LICENSED OPERATOR EVENTS FOR 1978 and 1979 I

VENDOR TOTAL PLANTS LERs AVERAGE / PLANT i

B&W 9 58 6.44 CE 8 45 5.63 GE 25 142 5.68 W 25 131 5.24 1

Klein(NRC5280)DN55 i

h ^

i NUREG-0667 RECOMMENDATIONS NOT COVERED ON THE TASK FORCE LIST (Continued) 1 RESPONSE (Cont'd) j To address the expressed concern, INP0 performed an independent analysis

of the LER data. A search of the LER Data Base provided by NRC yielded 1 the following information on licensed operator errors between January 1978 and March 1980.

TABLE 2 LER OUTPUT ON LICENSED OPERATOR

EVENTS BETWEEN JANUARY 1978 AND MARCH 1980 VENDOR TOTAL PLANTS LERs AVERAGE / PLANT B&W 9 70 7. 8 CE 8 53 6.6 GE 25 151 6.0 W 25 146 5.8 i

l Klein(NRC5280)DN55

.. o NUREG-0067 RECOMMENDATIONS NOT COVERED ON THE TASK FORCE LIST (Continued)

A cursory review of the data might indicate that licensed operators gen-erate more LERs at B & W plants than at the other vendor-designed facil-ities. As noted in NUREG-0667 the reported LERs dacreased significantly i with age of the plant, those having already undergone the first several i years break-in period generally submitting the fewest LERs. According-ly, INP0 attempted to analyze the data in such a way as to remove the i bias caused by plant maturity. One approach was to eliminate from con-sideration plants more mature than any B&W plant. The starting date for I each vendor's first plant is shown in Table 3.

TABLE 3

, YEAR FIRST PLANT TAKEN CRITICAL VENDOR YEAR B&W 1973 CE 1971 GE 1959 W 1960 i

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Klein(NRC5280)DN55)

, . . . o NUREG-0667 RECOMMENDATIONS NOT C0VERED ON THE TASK FORCE LIST (Continued)

If only the plants that have been taken critical for the first time since 1973 are considered the B & W plants no longer generate the most licensed operator errors.

TABLE 4 LICENSED OPERATOR ERRORS AT PLANTS CRITICAL FOR THE FIRST TIME SINCE JANUARY 1973 VEND 0R TOTAL PLANTS LERs AVERAGE / PLANT B&W 9 70 7.8 CE 6 30 5.0 GE 12 100 8.3 W 16 119 7.4

. Analyzing only LERs since 1978, is like taking a time exposure picture l of operating history. By looking at a single time period, one can see all plant data affected by the same industry wide influences such as the TMI-2 accident. The number of LERs varies with plant age. No one

vendor-designed plant has a tendency to generate more licensed operator errors (LERs) than the others. In fact the data fluctuations for both GE and Westinghouse plants are larger than for B & W plants, and each bounds the B & W fluctuation.

INPO concludes that the LER data does not support a concern that the er-ror rate for licensed operators at B & W plants is greater than for other nuclear plants, once the bias due to plant maturity has been removed.

The INP0 letter will be forwarded to the NRC upon receipt. It will in-clude graphic representaions for clarity of conclusions.

Klein(New)DN55

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, ,. o INP0/NSAC CONCERNS NOT C0VERED ON THE TASK FORCE LIST

_I_ TEM I.A Procedural requirement for declaration of appropriate emergencies should te emphasized in plant training sessions.

RESPONSE

This is covered during :ocemse Training and Requalification Training at CR-3.

ITEM I.C Instrument technician work practices and their potential impact on plant safety should be reviewed in plant training sessions. Attention should be given to events similar to the 3/20/78 and 1/5/79 transients at the Rancho Seco plant where overcooling resulted from maintenance technician actions.

RESPONSE

This is covered during Craft Systems Training at CR-3 by implementing a revised program that includes the above transients.

ITEM II.D The industry should further analyze and resolve with the NRC the current reactor coolant pump trip procedures to be followed during a small break LOCA. Mandated procedures can be counterproductive to safety if they are not sufficiently discriminating to specific circumstances, and to specific plant designs.

RESPONSE

This item, which is also covered by NUREG-0667 Recommendation No. 20 requires further analysis to resolve the concern.

ITEM II.F Procedures for declaration of emergencies should be reviewed to deter-mine if responsibility for monitoring plant conditions which lead to declaration of a specific emergency category should be assigned to a specific individual. It is suggested that this individual would also be responsible for immediately infonning the senior person in charge at the time when these conditions for emergencies and emergency notification have been met.

RESPONSE

The Emergency Coordinator is responsible for the declaration of a specific emergency category based on plant conditions. The Shift Super-visor acts as the temporary Emergency Coordinator until the Nuclear Klein(NRC5280)DN90.

,.w o INP0/NSAC RECOMMENDATIONS NOT C0VERED ON THE TASK FORCE LIST (Continued)

ITEM II.F (Cont'd)

RESPONSE

Plant Manager or his designee assumes the role as the Emergency Coordi-nator. The temporary Emergency Coordinator is responsible for immedi-ately notifying the Nuclear Plant Manager or his designee when an emer-gency has been declared.

ITEM III.A.4 Susceptibility of control systems to incorrect information caused by electrical faults, e.g., choki ng off feedwater to steam generators, withdrawing rods, and Opening the turbine throttle.

RESPONSE

This is a long tem item which requires additional analysis.

ITEM III.B.1.C The system monitoring the incore temperatures automatically prints any temperature which indicates in excess of 700 F. The basis for selecting 700 F should be reviewed to determine if this number should be revisec, since data was lost during the transient.

RESPONSE

The automatic thermocouple printout is for information only and is not safety related. A review of the printer start logic and setpoints is underway. FPC's action to resolve this concern is described in our response to NUREG-0667 Recommendation No. 7 of this submittal.

ITEM III.B.2 Plant transient monitoring and recording. Plant transient records inde-pendent of process computer, to provide a tape record of main plant parameters, are desirable for all plants. They are desirable on an earliest practicable schedule.

RESPONSE

FPC intends to install the B3W RECALL System equipment as part of the Technical Support Center upgrade per NUREG-0578. The RECALL System records 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of 160 analog and 65 digital inputs and writes over the previous record continuously. Upon an incident signal, such as Reactor j Trip, RECALL will record the next 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> of data but will retain the hour prior to the incident signal. Additional details of this system will be provided per our discussions with the NRC staff concerning our l implementation of NUREG-0578.

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Klein(NRCS280)DN55 I