ML19340C812
| ML19340C812 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 12/15/1980 |
| From: | Baynard P FLORIDA POWER CORP. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-1.C.5, TASK-1.D.1, TASK-2.K.2.09, TASK-2.K.2.10, TASK-2.K.2.13, TASK-2.K.2.16, TASK-TM 3-120-4, NUDOCS 8012170505 | |
| Download: ML19340C812 (25) | |
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- co 3E-4 Mr. Darrell G. Eisenhut, Director y- m @
Division of Licensing 4 Office of Nuclear Reactor Regulation Nuclear Regulatory Comission Washington, D.C. 20555
Subject:
Crystal River Unit 3 Docket No. 50-302 Operating License No. DPR-72 NUREG-0737 Post-TMI Requirements
Dear Mr. Eisenhut:
Pursuant to 10 CFR Part 50.54(f), Florida Power Corporation (FPC) hereby provides the information requested in your letter of October 31, 1980.
Your letter of October ' 31, 1980, incorporates into one document, all TMI-related items approved for implementation by the Comission at this time. By this letter, Florida Power Corporation provides confirmation that the implementation date will be met, and, for any date that can not be met, furnishes a revised implementation date, justification for the revision, and any planned compensating measures to be taken during the interim.
Enclosure 1 is a summary listing of all applicable NUPEG-0737 (Action Plan) items, your requested implementation and/or submittal dates, and the dates by which Florida Power Corporation intends to meet the imple-mentation and/or submittal requirement. Enclosure 2 provides a statement of confirmation for each applicable Action Plan item or a request for re-lief and supporting justification for any delay in not meeting the required dates.
i Should any unforeseen circumstances arise that preclude Florida Power Corporat. ion's implementation as identified in this submittal, we will inform you within five (5) working days of the circumstance involved and any changes in comitments.
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- General Othce 3201 ininy founn street soum . P O Box 14042, st Petersburg Florca 33733 613 -866-5151
. 8 012170 foe' , _ _ _ _ _ _
Mr. Darrell G. Eisenhut, Director December 15, 1980 Page 2 As always, we are prepared to meet with you and discuss our implementa-tion and/or submittal of these requirements.
Very truly yours, FLORIDA POWER CORPORATION O .- . .
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Dr. P. Y. Baynard Manager Nuclear Support Services Items (Ltr) i i
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STATE OF FLORIDA COUNTY OF PINELLAS P. Y. Baynard states that she is the Manager, Nuclear Support Services Department of Florida Power Corporation; that she is authorized on the part of said company to sign and file with the Nuclear Regulatory Com-mission the information attached hereto; and that all such statements made and matters set forth therein are true and correct to the best of her knowledge, information and belief.
Q . - y+. re- . 2 0 (#. Y./ Eynard Subscribed and sworn to before me, a Notary Public in and for the State and County above named, this 15th day of December,1980.
N ( Notary Public /
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Notary Public, State of Florida at Large, My Commission Expires: June 8, 1984 PYB/MAHNotary(DN-98)
ENCLOSURE 1 Sult%RY LISTItG 0F APPLICABLE PUREG-0737 ITEMS l NUREG-0737 FPC NUREG-0737 FPC LICENSIfG LICENSIfG ITEM TITLE DESCRIPTION IMPLEENTATION IMPLEMENTATION SUBMITTAL SUBMITTAL I.A.1.1 Shift Technical 1. On duty 1/1/80 1/1/80 1/1/80 1/1/80 Advisor 2. Tech Specs 12/15/80 --
9/1/80 9/15/80
- 3. Trained per 1/1/81 1/1/82 1/1/81 1/1/81 Cat B LL
- 4. Current program 1/1/81 1/1/81 1/1/81 1/31/81 description
- 5. Lon(;-term program 1/1/81 --
1/1/81 --
description I.A.1.2 Shift Supervisor Delegate non-safety 1/1/80 1/1/80 1/1/80 1/1/80 Responsibilities duties I . A.I.3 - Shift Manning 1. Limit overtime 11/1/80 11/1/80 11/1/80 11/1/80
- 2. Minimm shift crew 7/1/82 11/13/80 11/1/80 11/13/80
- 3. Tech Specs No date given --
No date 12/31/80 given I.A.2.1 Upgrading of R0 and 1. SR0 experience 5/1/80 5/1/80 None req'd N/A SR0 Training and 2. SR0s be R0s for 12/1/80 12/1/80 None rug'd N/A Qualifications 1 year
- 3. Three conth training R/1/80 8/1/80 None reg'd N/A on shift
- 4. Modify training 8/1/80 8/1/80 8/1/80 9/15/80
- 5. Facility certifi- 5/1/80 5/1/80 None req'd N/A cation I.A.2.3 Adninistration of Instructors complete SR0 8/1/80 8/1/80 None req'd N/A Training Programs exam -
ENCLOSURE 1 SUM %RY LISTItG OF APPLICABLE NUREG-0737 ITEMS NUREG-0737 FPC NUREG-0737 FPC LICENSIfG LICENSIfG ITEM TITLE DESCRIPTION IMPLEMENTATION IMPLEMENTATION SUBMITTAL SUBMITTAL I.A.3.1 Revised Scope and 1. Increase scope 5/1/80 N/A None req'd N/A Criteria for SR0 2. Increase passing 5/1/80 N/A None mq'd N/A and R0 Exams grade
- 3. Simulator exams 10/1/81 10/1/81 None mq'd N/A I.C.1 Short-Tenn Accident 1. SB LOCA 6/1/80 6/1/80 None mq'd N/A and Procedures 2. Inadequate core Review cooling
- a. Reenalyze and 1/1/81 1/1/81 1/1/81 1/31/81 propose guide-lines .
- b. Revise proced- First mfueling First mfueling Not deter- --
dures outage after outage after mined 1/1/82 1/1/82
- 3. Transients and accidents
- a. Reanalyze and 1/1/81 1/31/81 1/1/81 1/31/81 propose guide-lines
- b. Revise proced- First mfueling First mfueling Not deter- --
ures outage after outage after mined 1/1/82 1/1/82 I.C.2 Shift and Relief Implement shift turnover 1/1/80 1/1/80 1/1/80 1/11/80 Turnover Proced- checklist ures
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ENCLOSURE 1 SuttnRY LISTIfG OF APPLICABLE MJREG-0737 ITEMS NUREG-0737 FPC NUREG-0737 FPC LICENSING LICENSING ITEM TITLE DESCRIPTION IMPLEENTATION IMPLEMENTATION SUBMITTAL SUBMITTAL ,
I.C.3 Shift Supervisor Define msponsibilities 1/1/80 1/1/80 1/1/80 1/11/80 Responsibility I.C.4 Control Room Access Establish authority 1/1/80 1/1/80 1/1/80 1/11/80 limit access I.C.5 Feedback of Oper- FPC to implement proced- 1/1/81 1/1/81 None reg'd N/A ating Experience ures 4
I.C.6 Verify Correct Per- FPC to mvise performance 1/1/81 1/1/81 None req'd N/A fomance of Oper- procedures ating Activities i
I.D.1 Control Room Design Preliminary assessnent TBD --
TBD --
Reviews and schedule for correct-ing deficiencies I.D.2 Plant Safety Param- 1. - Description IBD --
IBD --
eter Display 2. Installed TBD --
TBD --
Console 3. Fully implemented TBD --
TBD --
II.B.1 Reactor Coolant 1. Design vents 7/1/81 7/1/81 7/1/81 7/1/81 System Vents 2. Install vents 7/1/82 7/1/82 7/1/81 7/1/81
- 3. Procedures 1/1/82 1/1/82 1/1/81 1/31/81 II.B.2 Plant Shielding 1. Review designs 1/1/80 1/1/80 1/1/80 1/1/80
- 2. Plant nodifications 1/1/82 1/1/82 1/1/81 1/1/81
-(LLCatB) (exceptWaste Disposal Panel)
- 3. Equipment qualifica- 6/30/82 6/30/82 1/1/82 1/1/82 tion
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ENCLOSURE 1 Sutt1ARY LISTItG 0F APPLICABLE NUREG-0737 ITEMS NUREG-0737 FPC NUREG-0737 FPC LICENSItG LICENSItG ITEM TITLE DESCRIPTION IMPLEMENTATION IMPLEENTATION SUBMITTAL SUBMITTAL II.B.3 Post-Accident 1. Interim system 1/1/80 1/1/80 1/1/80 2/15/80 Sampiing 2. P1 ant modifications 1/1/82 1/1/82 1/1/82 1/1/82 (LL Cat B)
II.B.4 Training for Miti- 1. Develop training 1/1/81 1/1/81 1/1/81 gating Core Damage program
- 2. Implement program
- a. Initial 4/1/81 6/1/81 None mq'd N/A
- b. Complete 10/1/81 (Core damage None req'd N/A mitigation pro-gram not devel-oped II.D.1 Relief and Safety 1. Submit program 1/1/80 1/1/80 1/1/80 1/11/80 Valve Test Require- 2. RV and SV testing ments (LL Cat B)
- b. Plant specific 10/1/81 1/1/82 ules report
- 3. Block valve testing 7/1/82 7/1/82 --
1
ENCLOSURE 1 SUttERY LISTItG OF APPLICABLE NUREG-0737 ITEMS NUREG-0737 FPC NUREG-0737 FPC LICENS!!G LICENSItG ITEM TITLE DESCRIPTION IMPLEMENTATION IMPLEENTATION SUBMITTAL SUBMITTAL i
11.0.3 Valve Position In- 1. Install direct indi- 1/1/80 1980 nefueling 1/1/80 1/11/80 dication cations of valve outage positions
- 2. Tech specs 12/15/80 9/1/80 9/15/80 II.E.1.1 Auxiliary feedwater 1. Short '.erm 7/1/81 See II.E.1.2 Plant See II.E.1.2 System Evaluation Specific
- 2. Long-tenn 1/1/82 See II.E.1.2 Plant See II.E.1.2 Specific II.E.1.2 Auxiliary Feedwater 1. Initiation System Initiation a. Control-grade 1/1/80 1/1/80 1/1/80 11/17/79 and Flow
- b. Safety-grade 7/1/81 Beyond 1st qtr. 1/1/81 12/19/80 1982
- 2. Flow indication
- a. Control-grade 1/1/80 1/1/80 1/1/80 11/17/79 1/11/80
- b. LL A tech specs 12/15/80 9/1/80 12/31/80
& 7/13/79
- c. Safety-grade 7/1/81 Beyond 1st qtr. 1/1/81 12/19/80 1982 II.E.3.1 Emergency Power for 1. Upgrade power supply 1/1/80 1/1/80 1/1/80 11/17/79 Pressurizer Heaters 1/11/80
- 2. Tech specs 12/15/80 9/1/80 12/31/80 II.E.4.1 Dedicated Hydrogen 1. Design 1/1/80 Existing 1/1/80 1/11/80 Penetrations 2. Install- 7/1/81 Existing 7/1/81 1/11/80
4 ENCLOSURE 1 SUNMARY LISTING OF APPLICABLE fUREG-0737 ITEMS NUREG-0737 FPC
+
NUREG-0737 FPC LICENSING LICENSING ITEM TITLE DESCRIPTION IMPLEENTATION IMPLEKNTATION SUBMITTAL SUBMITTAL
!!.E.4.2 Containment Isola- 1-4 Improve diverse 1/1/80 1980 refueling 1/1/80 1/11/80 tion Dependability isolation outage
- 5. Containment pressure setpoint
- a. Specify pressure 1/1/81 N/A 1/1/81 4/12/79
- b. Modifications 7/1/81 N/A 1/1/81 N/A
- 6. Containcent purge 1/1/81 Existing 1/1/81 1/10/79 valves
- 7. Radiation signals on 7/1/81 Existing 7/1/81 FSAR purge valves
- 8. Tech specs 12/15/80 9/1/80 9/15/80 II.F.1 Accident-Monitoring 1. Noble 9as monitor 1/1/82 1/1/82 1/1/82 1/1/82
- 2. Iodine / particulate 1/1/82 1/1/82 1/1/82 sampling
- 3. Containcent high- 1/1/82 1/1/82 1/1/82 1/1/82 range monitor
- 4. Containcent pressure 1/1/82 1/1/82 1/1/82 1/1/82
- 5. r an taincent water 1/1/82 1/1/82 1/1/82 1/1/82 level
- 6. Containment hydrogen 1/1/82 1/1/82 1/1/82 1/1/ 82 II.F.2 Instrumentation for 1. Subcool aeter 1/1/S0 1980 refueling 1/1/80 11/17/79 Detection of Inade- outage 1/11/80 qua'a Core Cooling 2. . Tech spec (LL Cat A) 12/15/80 9/1/80 9/15/80
- 3. Install level instru- 1/1/82 Upon NRC 1/1/82 Within 30 ments (LL Cat B) approval days of availability 4
ENCLOSURE 1 SUfMRY LISTING OF APPLICABLE f0 REG-0737 ITEMS NUREG-0737 FPC NUREG-0737 FPC LICENSItG LICENSIfC ITEM TITLE DESCRIPTION IMPLEMENTATION IMPLEMENTATION SUBMITTAL SUBMITfAL II.G.1 Power Supplies for 1. Upgrade to energency 1/1/80 Existing 1/1/80 11/17/79 Pressurizer Relief sources 1/11/80 Valves, Block 2. Tech specs 12/15/80 9/1/80 9/15/80 Valves, and Level Indicators II.K.2 Onfers on B&W 8. Upgrade APd sysem See II.E.1.1 See II.E.1.1 Plants 9. FMEA on ICS 8/17/79 8/17/79 8/17/79 8/17/79
- 10. Safety-grade trip 7/1/81 10/1/81 1/1/81 1/1/81
- 13. Thermal-mechanical 1/1/81 1/1/81 1/1/81 1/1/81 report
- 15. Effects of slug flow Complete 2/7/80 Complete 2/7/80 on OTSGs
- 16. RCP seal damage Complete 12/10/79 Complete 12/10/79
- 17. Voiding in RCS Complete 2/14/80 Conplete 2/14/80
- 19. Benchmark analysis Coaplete 2/8/80 Complete 2/8/80 of seq. AFW flow <
II.K.3 Final RecocTaenda- 1. Auto PORV isolation tions,' B&O Task a. Design l 7/1/81 See II.K.3.2 7/1/81 7/1/81 Force b. Test / install 1st n2 fuel 6 mo. TBD 7/1/81 If reg'd after NRC approval
- 2. Report on PORV fail- 1/1/81 1/31/81 1/1/81 1/31/81 Ures
ENCLOSURE 1 SUtfMRY LISTItG OF APPLICABLE PUREG-0737 ITEMS NUREG-0737 FPC NUREG-0737 FPC LICENSItG LICENSitG ITEM TITLE DESCRIPTION IMPLEMENTATION IMPLEENTATION SUBMITTAL SUBMITTAL ,
II.K.3 Final Recocuenda- 5. Auto trip of RCPs tions, B&O Task a. Propose rodifi- 7/1/81 Contingent on 2/15/81 Contingent Force (Cont'd) cations LOFT L3-6 pre- on LOFT L3-6 diction prediction
- b. Modify 3/1/82 7/1/81 1 7. Eval of PORV opening 1/1/81 1/31/81 1/1/81 1/31/81 probability
- 17. ECC system outages 1/1/81 4/1/81 1/1/81 4/1/81
- a. Schedule outline 1/15/80 11/17/80 11/15/80 11/17/80
- b. Model 1/1/82 1/1/82 .1/1/82 1/1/82
- c. thw analysis 1/1/83 or 1/1/83 or 1/1/83 or 1/1/83 or 1 yr. after 1 yr. after 1 yr. after 1 yr. after approval approval approval approval
- 31. Compliance with 1/1/83 or 1/1/83 or 1/1/83 or 1/1/83 or CFR 50.46 1 yr. after 1 yr. after 1 yr. after 1 yr. after
- 40. RCP seal damage approval approval approval approval See II.K.2.15
- 43. Effects of slug flow See II.K.2.16 III.A.I.2 Upgrade Emergency 1. Interim (TSC, OSC, 1/1/80 1/11/80 1/1/80 1/11/80 Support Facilities EOF)
- - - _ _ _ -_ . _ _ __ _ - - . . . . _ _ . - - _ ._ _ _ - ~ _ . - _. ._- __ _ _ .- .- _,_
ENCLOSURE 1 SUWARY LISTIrG OF
! APPLICABLE f0 REG-0737 ITEMS -
l NUREG-0737 FPC tOREG-0737 FPC LICENSItG LICENSitG ITEM TITLE DESCRIPTION IMPLEENTATION IMPLEfENTATION SUBMITTAL SUBMITTAL III.A.2 Emergency Prepared- 1. Upgrade emergency 4/1/81 4/1/81 1/2/81 1/2/81 ness plan to APP. E, 10 CFR 50
- 2. Meteorological data -6/1/83 6/1/83 1/2/81 1/1/81 III.D.1.1 Primary Coolant 1. Leak reduction 1/1/80 1980 refueling 1/1/80 1/1/81 Outside Containment 2. Tech spe.cs 12/15/80 9/1/80 12/31/80 1
I III.D.3.3 In-plant Radiation 1. Provides means to 1/1/80 1/1/80 1/1/80 1/11/80 Monitoring detemine presence of radioiodine
- 2. Modifications to 1/1/81 1/1/81 1/1/81 7/7/80 ,
accurately measure I2 III.D.3.4 Control Room 1. Review 1/1/81 1/31/81 1/1/81 1/31/81 Habitability 2. Modification 1/1/83 1/1/83 1/1/81 1/31/81 4
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9 ENCLOSURE 2 NUREG-0737 ITEMS ITEM I.A.I.1 SHIFT TECHNICAL ADVISOR In compliance with the short-term requirements of NUREG-0578 and the sub-sequent clarification dated October 30, 1979, FPC is presently utilizing a group of iterim STAS. The qualifications, training, duties and shift rotation of the interim STAS have been accepted by the NRC Staff (see your May 5,1980, " Evaluation of NUREG-0578 Category A Implementation";
Reid to Hancock).
FPC has developed and is presently implementing 9 program for permanent STA training based upon the document included in tne Action Plan (INP0 Guidelines Rev. O, April 18,1980). The program is being conducted utfl-izing the University of Floriaa Nuclear Engineering Department, NUS, B&W and FPC. This program will be completed by December 31, 1981. The Jan-uary 1,1981, schedule for the completion of this program is unrealistic based upon the time frame required to recruit qualified personnel to be trained and tN vast scope of NRC and INP0 guidelines used to develop this program. Until the training program for permanent STAS is complete, the interin STAS will remain on duty. The current FPC " Nuclear Opera-tions Technical Advisor Training Program", STA job description, STA qual-ifications and STA requalification program till be submitted to the NRC Staff for review by January 31, 1981.
Until the NRC Staff provides additional guidance on the long-term STA program (i.e., how the long-term program should differ from the present program), FPC cannot provide definite plans for this program. Several options are being considered and will be addressed in the January 31, 1981 submittal to assist the NRC in establishing long-term improvements in the STA program.
A technical specification change request to include the STA into the min-
- imum shif t complement was filed on September 15, 1980.
ITEM I.A.1.2 SHIFT SUPERVISOR RESPONSIBILITIES FPC has implemented the requirements of this Item and has received NRC's concurrence (see your May 5, 1980, letter " Evaluation of NUREG-0578 Category A Implementation"; Reid to Hancock).
ITEM I.A.1.3 SHIFT MANNING As delineated by this Action Plan Item, the limitation of overtime for i nuclear power plant operators (ref. IE Circular 80-02, dated February 1, 1980) has been implemented and incorporated into Section 2.26 of AI-500 of the Crystal River Unit 3 Plant Operations Quality Assurance Manual.
The minimum requirements for shift manning as delineated by Darrell Eisenhut's letter to all licen.>ees (July 31,1980) and superceded by NUREG-0737 have been implemented and incorporated into Section 2.1 of
Al-500 of the Crystal River Unit 3 Plant Operations Quality Assurance
] Manual. See our letter (Bright to Eisenhut) dated November 13, 1980.
A technical specification change request will be submitted to you on the minimum shif t manning requirements by December 31, 1980.
ITEM I.A.2.1 IMMEDIATE UPGRADING 0F REACTOR OPERATOR (RO) AND SENIOR REACTOR OPERATOR (SRO) TRAINING AND QUALIFICATIONS All FPC applicants for SR0 or R0 examinations will meet the experience criteria delineated in this Action Plan Item and Harold R. Denton's letter dated March 28, 1980.
The SRO and R0 training programs have been modified vis-a-vis Harold R. Denton's letter dated March 28, 1980. The FPC response to the above letter was dated September 15, 1980 (Items 2C.1.,2., 3. address the modified training).
Certifications completed pursuant to Sections 55.10a(6) and 55.33a(4) and (5) of 10 CFR Part 55 will be signed by a designated officer of the Com-pany within the Nuclear Operations Department.
ITEM 1.A.2.3 ADMINISTRATION OF TRAINING PROGRAMS FPC Nuclear Operations Training instructors and/or contractor personnel who teach plant systems, plant integrated responses, and plant transient response have successfully completed the SR0 examination. This qualifi-cation will be maintained.
ITEM 1.A.3.1 REVISE SCOPE AND CRITERIA FOR LICENSING EXAMS FPC recognized the revised scope and criteria for SR0 and R0 examinations vis-a-vis Harold R. Denton's letter of March 28, 1980, and has incorpora-ted additional heat transfer, thenaodynamics and fluid mechanics study into the SR0 and R0 training program.
FPC will coordinate new SRO & R0 examinations with the NRC and the NSSS vendor after October 1,1981.
ITEM 1.C.1 SHORT-TERM ACCIDENT AND PROCEDURES REVIEW FPC meets the intent of the requirements of this Item through participa-tion in the B&W Owners Group Abnormal Transient Operating Guidelines (AT0G) Program. Crystal River Unit 3 specific guidelines will be devel-oped for all postulated single and multiple failure events.
The Crystal River Unit 3 guidelines are scheduled to be completed by June 30, 1981. These guidelines will be implemented no later than the first refueling outage af ter January 1, 1982.
An in-depth program description will be submitted by January 31, 1981.
I 1
ITEM I.C.2 SHIFT AND RELIEF TURNOVER PROCEDURES The requirements of this Item are complete. Plant procedures provide guidance for a complete and systematic turnover between the off-going and on-coming shift at Crystal River Unit 3 to provide greater assurance that critical plant paramaters are within limits and that the availability and alignment of safety systems are made known to the on-coming shif t (see your May 5, 1980, " Evaluation of NUREG-0578 Category A Implementation";
Reid to Hancock).
ITEM I.C.3 SHIFT SUPERVISOR RESPONSIBILITY The requirements of this Item are complete. FPC has revised the respon-sibilities of the Shift Supervisor. This revised responsibility has been set forth in plant procedures (see your May 5, 1980, " Evaluation of NUREG-0578 Category A Implementation"; Reid to Hancock).
ITEM I.C.4 CONTROL ROOM ACCESS The requirements of this Item are complete. FPC has implemented plant procedures which will limit control room access during an emergency (see your May 5,1980, " Evaluation of NUREG-0578 Category A Implementation";
Reid to Hancock).
ITEM I.C.5 FEEDBACK OF OPERATING EXPERIENCE FPC has implemented procedures to provide operating information pertinent to plant safety originating within and outside FPC. This information is provided to operators and other personnel and be incorporated into train-ing and retraining programs. These procedures will meet the NUREG-0737 position for this Item. ,
ITEM I.C.6 VERIFY CORRECT PERFORMANCE OF OPERATING ACTIVITIES FPC reviews and revises plant procedures, as part of an on-going pro-gram. To assure that an effective system of verifying the correct per-formance of operating activities is provided as a neans of reducing human error and improving the quality of normal operations. Procedure improve-ments will also be made during the human factors Control Room Design Reviews detailed in Action Plan Item I.D.1.
ITEM I.D.1 CONTROL ROOM DESIGN REVIEWS Since no implementation date is requested, FPC will await NRC guidelines to be issued in 1981 prior to making any commitments on this Item.
ITEM I.D.2 PLANT SAFETY PARAMETER DISPLAY CONSOLE Since no implementation date is requested, FPC will await further guidance prior to making any comnitments on this Item.
ITEM II.B.1 REACTOR COOLANT SYSTEM VENTS By the Action Plan required submittal date of July 1,1981, FPC will sub-mit information on the reactor coolant vent system for staff review. Our submittal will include a description of the design, results of LOCA analyses, procedures and supporting analysis for use, and other sup-porting information, as appropriate, for vents on the reactor coolant system hot legs and the pressurizer.
No vent for the reactor vessel head is proposed due to the current B&W position that there 'is no need to vent in this location to maintain natural circulation conditions, even during cooldown. A position docu-ment on venting the reactor vessel head will be provided to you by Jan-uary 31, 1981. This position supercedes our previous commitment to install a reactor vessel head vent. See our letter date April 11, 1980 (Moore to Denton).
FPC will install hot leg and pressurizer vents during the Fall 1981 refueling outage and will therefore meet the ~ quired Action Plan imple-mentation date of July 1,1982 for this Iter FPC will submit a technical specification change request defining opera-bility requi rements of the vent system within 90 days of the actual implementation date.
ITEM II.B.2 DESIGN REVIEW OF PLANT SHIELDING AND ENVIRONMENTAL QUALI-FICATION OF EQUIPMENT FOR SPACES / SYSTEMS WHICH MAY BE USED IN POST-ACCIDENT OPERATIONS All plant modifications identified as a result of FPC's shielding review have been initiated and completion is scheduled prior to the Action Plan implementation date of January 1, 1982, except for relocation of the Waste Disposal Panel. Evaluation of the relocation of this panel is cur-rently underway and is estimated to be achievable by January 1,1982.
However, a comprehensive review of the entire Waste Disposal System has been undertaken. The outcome of this review could impact the ultimate location for the Waste Disposal Panel.
Because of the preliminary nature of the Waste Disposal System upgrade, a commitment to the Action Plan required date of January 1,1982, could re-sult in a request for extension to consider integration of the relocation and system upgrade in the overall radwaste system modifications. Should this extension become necessary, FPC will submit its request, including justification of the extension, within 30 days of our acknowledgement.
In the interim, FPC proposes to continue use of the short-term modifica-tions.you previously found acceptable (see your May 5,1980, " Evaluation of NUREG-0578 Category A Implementation"; Reid to Hancock).
ITEM II.B.3 POST-ACCIDENT SAMPLING CAPABILITY By the Action Plan required date of January 1,1982, FPC will complete the sampling and analysis modifications as described to you in our letter dated February 15,1980, (Baynard to Denton). A description of the de-tails will be provided for your review prior to January 1,1982.
l
1 ITEM II.B.4 TRAINIflG F_0R MITIGAllijG CORE DAMAGE FPC is developing a program along INPO guidelines. The first part of the program up to Recognition of Core Damage is to be developed by January 1, 1981, and implemented by June 1,1981.
Part 2 of the program will address Mitigation of Core Damage. It is FPC's position that present information is insufficient to develop this program (see letter Hancock to Collins, dated September 15, 1980, item 2C,2.).
ITEM !!.D.1 RELIEF AND SAFETY VALVE TEST REQUIREMENTS FPC is a sponsor of the EPRI PWR Safety and Relief Valve Test Program and intends to comply with the requirements of NUREG-05/8, Item 2.1.2. By
! letter dated December 15, 1980, R.C. Youngdahl of Consumers Power Company has provided the current PWR Utilities' positions on NUREG-0737, Item 11.D.1 clarifications. FPC endorses the following positions:
A. Safety and Relief Valves and Piping - The EPRI " Program Plan for Performance Testing of PWR Safety and Relief Valves", Re-vision 1, dated July 1, 1980, ducs provide a program that satisfies the NRC requirements. Discussion with the NRC Staff and their consultants are resolving specific detailed issues.
B. Block Valves - The EPRI Program has not formally included the testing of block valves. However, a small number of block valves has been tested at the Marshall Steam Station Test Facility. The PWR Utilities and EPRI cannot provide a detailed block valve test program until results of the Wyle Labs and CE relief valve tests are available. Therefore, a block valve test program will not be provided before July,1981. The PWR Utilities and EPRI believe that the proper operation of the TMI-2 and Crystal River block valves and other operational experi ence, plus knowledge of the Marshall tests, support a less hurried and more rational approach to block valve testing.
- C. ATWS Testing - PWR Utilities do not plan to support additional efforts for ATWS valve testing until regulatory issues are re-solved. The major safety and relief valve test facility (CE) is nearing completion and some measures were taken to provide
- additional test capability beyond the current program require-ments. The NRC should recognize that results from the current program are likely to provide most of the information necessary to address ATWS events (i.e., relief capability at high pres-
! sures).
ITEM II.D.3 DIRECT INDICATION OF RELIEF AND SAFETY VALVE POSITION Relief and safety valve position indication equipment has been installed at Crystal River Unit 3 per NUREG-0578, Item 2.1.3.a requirements.
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i A testing program is currently underway to document seismic and environ-mental qualification of the installed acoustical monitoring system. This testing program is scheduled for completion by June 30, 1981.
A technical specification change request was submitted to you on Sep-tember 15, 1980.
ITEM II.E.1.1 AUXILIARY FEEDWATER SYSTEM EVALUATION On December 27, 1979, FPC submitted to you a simplified auxiliary (emergency) feedwater system reliability analysis, as required in Item (1). As stated in your clarification, FPC will await a letter from you requesting specific modifications based on your review of Item (1).
However, we believe the EFW upgrade proposed in response to Action Plan Item II.K.1.2 will eliminate the necessity of Staff recommendations for this Item.
ITEM II.E.1.2 AUXILIARY FEEDWATER SYSTEM AUTOMATIC INITIATION AND FLOW
. INDICATION The short-term requirements for automatic initiation and flow indication of emergency feedwater have been completed. FPC has installed at Crystal River Unit 3 a reliable, control-grade, redundant system which meets the single failure criteria to automatically initiate emergency feedwater and to indicate the emergency feedwater delivery to -each steam generator.
The methods by which we have implemented these requirements are discussed in our letters to you dated November 17, 1979, and January 11, 1980.
Since the Three Mile Island Unit 2 incident, Florida Power Corporation has reviewed the many recommendations regarding emergency feedwater (i.e., NRC Commission Orders for B&W designed plants, NUREG-0578, the CR-3 IREP, NUREG-0667, NUREG-0660, NUREG-0737 and FPC's Nuclear Safety Task Force). In an effort to consolidate the many recommendations re-garding the emergency feedwater system into an overall upgrade effort, FPC, in conjunction with two other utilities, has taken the initiative to develop and design an EFW system that incorporates the following major features:
. Safety grade automatic initiation and control of EFW
. Independent of ICS, NNI and other non-safety systems
. Redundant and testable
- . Meets single failure criteria
. Manual initiation and control provisions
. Qualified hardware (seismic and environmental)
. Provides EFW operational status, flowrate, and 0TSG level information
. Minimizes overcooling following loss of MFW or RCP's
. Minimizes OTSG dryout
. Terminates MFW or EFW overfill
. EFW supplied to " good" steam generator only following steamline break event (F0GG)
9 This design philosophy and FPC's intention to proceed with the overall upgrade of the EFW system were presented to members of the NRC Staff in a meeting on September 4, 1980. A submittal that describes our EFW upgrade design is scheduled to be provided to you no later than December 19, 1980. Your timely review and concurrence with our approach will be re-quested.
l In surrinary, the benefits to be realized by the overall EFW upgrade far outweigh the incremental benefits gained from upgrading to a control-grade system for automatic initiation and flow indication as required by s this Item. Specifically, the EFW upgrade will incorporate features to
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minimize overcooling transients and steam generator dryout through use of a safety-grade control system, preclude steam generator overfill, and I. assure feedwater is supplied to only the " good" steam generator following a steamline break event. Incorporation of the EFW upgrade package will provide the greater enhancement toward assuring heat removal from the
- primary system under all conditions.
Our efforts to upgrade the EFW system beyond the requirements of this Item do not allow for incorporation of safety-grade initiation and flow indication for EFW by the required implementation date of July 1,1981.
The overall EFW upgrade will require major equipment rnodifications.
Delivery dates of approximately one year upon receipt of the order are expected for this equipment. Therefore, as discussed with your Staff on September 4,1980, equipment delivery is not expected until early 1982.
This EFW upgrade will be incorporated at the first available outage of sufficient duration following completion of engineering and procurement.
A proposed change to the technical specification was submitted on July 13,1979 (Baynard to Denton).
A supplemental proposed change to tSe technical specifications will be submitted by December 31, 1980, for your review and approval. This change request will address the automatic initiation and flow indication aspects of the currently installed EFW system.
In conclusion, FPC believes that the presently installed redundant con-trol grade systems used to initiate and provide flow indication for the EFW system at Crystal River Unit 3 substantially meets the intent of the July 1,1981 requirements and provide adequate margins of safety during the period of operation until the overall upgrade identified herein can be installed.
ITEM II.E.3.1 EMERGENCY POWER SUPPLY FOR PRESSURIZER HEATERS This item is complete for Crystal River Unit 3 as we assured that 126 KW of pressurizer heaters is available to establish and maintain natural circulation at hot standby corJitions. A technical specification change request for assuring the availability of emergency power to pressurizer heaters will be submitted to you by December 31, 1980.
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- ITEM II.E.4.1 DEDICATED HYDROGEN PENETRATIONS Crystal River Unit 3 has satisfied the requirement for dedicated hydrogen control penetrations for post-accident hydrogen purge. However, a system modification to supplement the existing purge filter and piping with a dedicated H2 purge unit is underway. The added purge unit will be de-signed to allow contaminated filter changeout and to facilitate condensa-
! tion drainage back to the containment.
Installation will be accomplished during our Fall 1981 refueling.
Although this Item is not addressed in NUREG-0737, it is included here for documentation of a NUREG-0578 Category B Item completion schedule.
ITEM II.E.4.2 CONTAINMENT ISOLATION DEPENDABILITY ltems 1 through 4 were completed under the Short-Term Lessons Learned Requirements of NUREG-0578.
Item 5 requires justification of the containment pressure which initiates isolation of nonessential penetrations. Our April 12, 1979, let'ar Stewart to 0'Reilley) and February 11, 1980, letter (Baynard to Denton)
. identified those penetrations which were classed as nonessential and which are diversely isolated on High Pressure Injection (1500 psig RCS pressure). These nonessential penetrations also receive isolation signals on high containment pressure along with those of essential pene-t ration's. Therefore, reduction of the high containment pressure isola-tion setpoint for nonessential penetrations only would not apply to Crystal River Unit 3.
Item 6 has been satisfied by. compliance with the Staff Interim Position i
of October 23, 1979. See our letter ddted January 10, 1979; Stewart to l Reid.
item 7 is satisfied with the existing logic for isolation on high venti-l lation duct activity.
l ITEM II.F.1 ADDITIONAL ACCIDENT-MONITORING INSTRUMENTATION FPC will install the following by January 1,1982:
l . Noble gas effluent monitors t
. lodine gaseous effluent monitors
. Containment high-range radiation monitors
. Containment pressure monitors
.< Containment water level nonitors l . Containment hydrogen monitors 1
To meet the intent of the Action Plan position and clarification dated October 31, 1980. No deviation from your requirements are expected;
~i therefore, as requested by your letter of October 31, 1980, a final de-sign description of the as-built system will be available for staff re-view by January 1, 1982. Our design details for these modifications are 8
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not complete at this time, but a subnittal to describe our proposed modi-fications will be made within a re.isonabic period following completion of these design details and, if necessary, justification for implementation schedules beyond the January 1,1982, date.
A technical specification change request, as appropriate, will be sub-mitted within 90 days of implementation of this Item.
ITEM II.F.2 INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE COOLING FPC's position on additional instrumentation to detect inadequate core cooling was provided to you in our letter dated October 31, 1980. The existing incore thermocouple system, in conjunction with the redundant saturation meters, provides an advance warning of (and an unambiguous in-dication of) inadequate core cooling.
At the same time, FPC is interested in the development of other instru-mentation systems to accomplish the early and unambiguous detection of inadequate core cooling conditions. We have entered into a design devel-opment program for a hot leg level instrument. The present concept employed will be a differential pressure sensor which utilizes the hot leg vent and the RCS pressure tap as sensing locations. This system would be used in conjunction with the saturation meters and the incore thermocouples to meet the intent of your requirements on this Item.
The design development for the hot leg differential pressure system has not progressed to a point where a report can be provided to you by Jan-uary 1, 1981. Within 30 days of the availability of a system des-
- cription, a report will be provided for your review and conceptual approval. Consistent with our previous position (see our October 31, 1980 letter), we feel prudent engineering judgement dictates the hot leg differential pressure system be thoroughly developed and tested (and re-ceive NRC's concurrence) prior to its installation and use. Therefore, an installation date of January 1,1982 may not be achievable.
l ITEM II.G.1 EMERGENCY POWER FOR PRESSURIZER EQUIPMENT As noted in your May 5,- 1980, letter (" Evaluation of NUREG-0578 Cate-
! gory A Implementation", Reid to Hancock), power supplies for the pres-surizer relief valve, lock valve, and the pressurizer level indicators
- are crable of being powered from both off-site power and the on-site i emergency power system and meet the intent of this requirement.
A technical specification change request for operability requirements of these power supplies was submitted to you by our letter dated Septem-ber 15, 1980, (Baynard to Reid).
ITEM II.K 2.8 AUXILIARY FEEDWATER SYSTEM UPGRADING No separate implementation or submittal is required for this Item. See our responses to Items II.E.1.1 and II.E.1.2 as stated previously.
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ITEM II .K.2.9 FAILURE MODE EFFECTS ANALYSIS ON THE INTEGRATED CONTROL SYSTEM A generic failure mode and effects analysis of the ICS was submitted on August 17, 1979. NRC Staff recommendations are pending completion of i
Staff review.
FPC is awaiting NRC staff recon..endations. i t l ITEM II.K.2.10 SAFETY-GRADE ANTICIPATORf REACTOR TRIP Implementation of a control-grade anticipatory reactor trip (ART) was completed during our 1979 refueling and revised during our 1980 refueling to provide full redundancy of sensors and actuators in addition to pro-viding fail-safe logic to enhance reliability. This control-grade ART i system provides a reactor trip upon main turbine trip, trip of both main feedwater pumps and/or low-low level in both steam generators.
In addition, FPC requested B&W to develop a modification to provide a safety-grade anticipatory reactor trip upon main turbine trip and/or trip of both main feedwater pumps. This modification and the Bailey equipment have been received; however, analytical verification of seismic and heat load impact upon the existing reactor protection system will not be com-pleted until mid-February,1981. Issuance of a construction work package is scheduled for March 1,1981. The present cable and pressure switch delivery schedules associated with this modification may permit instal-lation of the entire modification by July 1,1981. However, a special, unscheduled outage of approximately two weeks for the final connections and testing would be necessary. Therefore, FPC is rec uesting relief from the July 1, 1981, implementation date to permit it..piementation of this modification during our scheduled refueling outage beginning in mid-to-late September, 1981.
1 FPC considers this request justified as we presently have installed a reliable, redundant, control-grade anticipatory reactor trip system which meets the single active failure criteria. This system is backed up by a fully qualified, safety-grade reactor protection trip system. We con-clude that these presently installed systems provide an adequate margin of safety to permit operation of Crystal River Unit 3 until our Fall 1981 refueling outage and these systems insure that the reactor is tripped before operating limits bounded by the safety analyses for CR-3 are exceeded, b
In addition, should a two week outage be required for this modification prior to the Fall 1981 refueling outage, the potential impact on over-all system reliability due to the high'sumner loads would need to be evaluated and discussed with your Staff.
ITEM II.K.2.13 THERMAL-MECHANICAL REPORT By January 1,1981, FPC will submit a detailed analysis on the thermal-mechanical conditions in the reactor vessel during recovery from small breaks with an extended loss of all feedwater. This report will meet the requirements of this Item.
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! ITEM II.K.2.14 LIFT FREQUENCY OF PORVs and SVs
ITEM II.K.2.15 EFFECTS OF SLUG FLOW ON STEAM GENERATOR TUBES l The results of this ar.alysis have been submitted by FPC and are under-going NRC Staff review. (See FPC letter dated February 7,1980; Baynard to Reid)
- ITEM II.K.2.16 REACTOR COOLANT PUMP SEAL DAMAGE 1 ,
The results of this analysis have been submitted by FPC and are under-
! going NRC Staff review. (See FPC letter dated December 10, 1979; Baynard
, to Reid)
ITEM II.K.2.17 POTENTIAL FOR V0IDING IN THE REACTOR COOLANT SYSTEM 1
DURING TRt.NSIENTS The results of this analysis have been submitted by FPC and are under-going NRC Staff review. (See FPC letter dated February 14, 1980; Baynard toReid)
ITEM II.K.P.19 SEQUENTIAL AUXILARY FEE 0 WATER FLOW ANALYSIS i
The results of this analysis have been submitted by FPC and are under-going NRC Staff review. (See FPC letter dated February 8, 1980; Baynard
- to Ross)
ITEM li.K.2.20 SMALL BREAK LOSS-0F-COOLANT ACCIDENT WHICH REPRESSURIZES THE REACTOR COOLANT SYSTEM TO THE POWER-0PERATED RELIEF VALVE SETPOINT
- The results of this analysis have been submitted by FPC and are under-going NRC Staff review. (See FPC letter dated February 28, 1980; Monre toRoss)
ITEM II.K.3.1 INSTALLATION AND TESTING OF AUTOMATIC POWER-0PERATED l
RELIEF VALVE ISOLATION SYSTLM l
This Item will be evaluated if required by Action Plan Item !!.K.3.2.
ITEM II.K.3.2 REPORT ON OVERALL SAFETY EFFECTS OF POWER-OPERATED RELIEF VALVE INDICATION SYSTEM This report will be submitted for NRC Staff review by January 31, 1981.
l ITEM II.K.3.5 AUTOMATIC TRIP'0F REACTOR COOLANT PUMPS DURING LOSS-0F-COOLANT ACCIDENTS l
By letter dated December 4,1980 (James H. Taylor of B&W to Paul Check of NRC), B&W submitted, on behalf of the B&W Owners Group, a description of i
I the analytical model B&W intends to use for the blind post-test predic-tion.of LOFT (L3-6). Approximately five (5) weeks after B&W has received the initial conditions of the LOFT L3-6 test, prediction results will be submitted to you.
. ITEM II.K.3.7 EVALUATION OF POWER-OPERATED RELIEF VALVE OPENING PROBABILITY DURING OVERPRESSURE TRANSIENT FPC will provide a report by January 31, 1981 which evaluates the PORV opening probability.
l ITEM II.K.3.17 REPORT ON OUTAGES OF EMERGENCY CORE C0OLING SYSTEMS (ECCS) LICENSE REPORT AND PROPOSED TECHNICAL SPECIFICATION CHANGES FPC is compiling this data for use in the Crystal River Unit 3 Reliabil-ity Model. A report deta'iling outage dates and lengths for all ECC system outages since commercial operation will be provided to you by
- April 1,1981. The nonavailability of this data by January 1,1981, will not impact the safe operation of the nuclear plant.
ITEM II.K.3.30 REVISED SMALL BREAK LOSS-0F-COOLANT ACCIDENT METHODS TO 1 SHOW COMPLIANCE WITH 10 CFR PART 50, APPEhDIX K FPC is participating with the B&W Owners Group on this item and intends to revise models and perform reanalysis, as necessary, vis-a-vis LOFT L3-6.
ITEM II.K.3.31 PLANT SPECIFIC CALCULATIONS TO SHOW COMPLIANCE WITH 10 CFR PART 50.46 Based upon the results of Item II.K.3.30, FPC wil'1 submit calculations, as necessary, by January 1,1983, or one year after Staff approval of revised LOCA models.
ITEM III.A.1.2 UPGRADE EMERGENCY SUPPORT FACILITIES FPC is awaiting the additional clarification on this Item and will sub-sequently provide descriptions of the TSC and EOF as required.
ITEM III.A.2 IMPROVED LICENSEE EMERGENCY PREPAREDNESS - LONG-TERM FPC emergency plans and procedures will be submitted as required in NUREG-0737. Complete updated emergency plans will be provided by January 2, 1981, and complete implementation procedures will be submitted by April 1,1981.
ITEM III.D.1.1 INTEGRITY OF SYSTEMS OUTSIDE CONTAINMENT LIKELY TO L CONTAIN RADI0 ACTIVE MATERIAL FOR PRESSURIZED-WATER REACTOR AND BOILING-WATER REACTORS This Item is complete except for submittal of the test leakage data and a f minor-ven'ilation system change identified by a review of North Anna type
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release potentials. Submittal of the above test leakage data will be prior to January 1,1931 and an auxiliary building cubicle door for posi-tive ventilation control will be conpleted in the first quarter of 1981.
A technical specification change request submittal will be made by Decem-ber 31,1980.
ITEM 111.0.3.3 IMPROVED IN-PLANT IODINE INSTRUMENTATION UNDER ACCIDENT CONDITIONS l This iten has been complete.1 with the single channel analyzer identified in our July 7,1980, letter, (Bright to Reid).
l ITEM 111.D.3.4 CONTROL ROOM HABITABILITY REQUIREMENTS I
- An extensive re-evaluation of habitability studies performed for the Crystal River Unit 3 control room are underway. This re-evaluation will provide the basis for our submittal to comply with identified criteria of the referenced Standard Review Plan (SRP) sections.
A formal submittal of the results of the above review, including NUREG-0737 Item 111.0.3.4, Attachment 1 information, will be provided on l January 31, 1981.
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