ML19211D120
| ML19211D120 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 01/11/1980 |
| From: | Baynard P FLORIDA POWER CORP. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 3--3-A-3, 3-0-3-A-3, NUDOCS 8001160476 | |
| Download: ML19211D120 (150) | |
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Florida Power C O R PO m A F'0 4 January 11, 1980 File: 3-0-3-a-3 Mr. Harold R. Denton 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
Dear Mr. Denton:
This letter supplements Florida Power Corporation's letters of October 17 and Novemaer 17, 1979, concerning implementation of the Short-term Lessons Learned Recommendations contained in NUREG-0578 by February 15, 1980, unless otherwise stated.
Infonnation on system reliability for the Florida Subregion will be provided by January 15, 1980, as requested in your January 4, 1980, letter.
Florida Power Corporation's Suppl _ mental Response, Encl osure 1, details the impl ement3 tion of the Short-Term Lessons Learned Recommendations. The various attachments provide supporting docu-mentation as follows:
Attachment I -- Information Required on the Subcooling Meter.
This provides detailed technical and analytical capabilities of the subcooling meter.
Attachment II -- Babcock & Wilcox Saturation Meter - Implementa-tion of NRC Requirements. The explanation of how the saturation gk meter meets the intent of the NRC reconmendations for single fail.. Cg5 To ure criteria, testability, seismic criteria, and environmental qualification, is identified.
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Florida Power CO R PO R A T IO N January 11, 1980 File: 3-0-3-a-3 Mr. Harold R. Denton Dire. tor 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
Dear Mr. Denton:
This letter supplements Florida Power Corporation's letters of October 17 and November 17, 1979, concerning implementation of the Short-tem Lessons Learned Recommendations contained in NUREG-0578 by February 15, 1980, unless otherwise stated.
Infomation on system reliability for the Florida Subregio.1 will be provided by January 15, 1980, as requested in your January 4, 1980, letter.
Florida Power Corporation's Supplemental Response, Enclosure 1, detail s the implementation of the Short-Term Lessons Learned Recommendations. The various attachments provide supporting docu-mentation as follows:
Attachment I -- Information Required on the Subcooling Meter.
This provides detailed technical and analytical capabilities of the subcooling meter.
Attachment II -- Bab.:ock & Wilcox Saturation Meter - Implementa-tion of NRC Requirements. The explanation of how the saturation meter meets the intent of the NRC recommendations for single fail-ure criteria, testabil i ty , seismic criteria, and environmental qualification, is identified.
1755 005 General Office 3201 Thirty-fourth Street South e P O Box 14042. St Petersburg. Flonda 33733 e 813-866-5151
,I .1 Mr. Harold. R. Denton Pe 'no January 11, 1980 Attachment III -- Design Review of Plant Shielding and Environmen-tal Qualification of Equipment for Spaces / Systems Which May be Used in Post-accident Operations Outside Containment at Crystal River Unit 3 Nuclear Generating Station. lite review of the post-accident radiation fields, based on guidelines provided in Darrell G. Ei senhut's September 13, 1979, letter, and your October 30, 1979, letter is summarized.
Attachment IV -- Venting Design Criteria. This is a summary of the venting design criteria utilized in venting noncondensible gases to aid in refilling the RCS and promoting natural circula-tion flow for core cooling.
Attachment V -- AI-500, Conduct of Operations. This in-plant im-plementing procedure identifies the method of operation for the Operations Staff.
Attachment VI -- Management Responsibility of Nuclear Shift Super-visor. This directive emphasizes the primary management responsi-bility of the Shift Supervisor and clearly establishes his command duties.
Attachment VII -- AI-200, Organization and Responsibility. This in-pl ant implementing procedure identifies the Crystal River Unit 3 organization and responsibilities.
Attachment VIII -- Long-range Plan for Upgrading the On-site Tech-nical Support Center. r'reliminary facility plans are provided.
Based on system reliabiliti considerations, and the necessary equipment for implementation of Recommendation 2.13.a - Response to Direct Indication of Power-operated Relief Vaives and Safety Valve Position for PWRs and BWRs, Florida Power Carporation will implement the PORV and Safety Valve Position Indir ation ~modifica-tion within 30 days of receipt of all equi pment. The mounting cabine+.s are presently scheduled to be shipped from B&W to FPC en February 22, 1980.
As a result, the required compliance and maintenance outages of other FPC generating units will be scheduled in accordance with an anticipated implementation deadline for this modification of late March, 1980.
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Mr. Harold R. Denton Page Three January 11, 1980 Should any unforseen circumstances arise that precl ude Florida Power Corporation's implementation of the Short-term Lessons Learned Recommendations are identified in this submittal, we will inform you within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of the circumstances involved and any changes in commitments.
Very truly yours, FLORIDA POWER CORPORATION f' Y Dr. P. Y. Baynard Manager Nuclear Support Services NUREG-0578(Txmtl Ltrl)DN-94 Attachments 1795.007
STATE V FLORIDA COUNTY OF PINEll.AS Dr. P. Y. Baynard states that she is i.he Manager, Nuclear Support Services, of Florida Power Corporation; that she is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the infonnation attached hereto; and that all such statements made and matters set forth the ein are true and correct to the best of her knowledge, information, and belief.
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[ / (Sr. P/ Y. Baynard Subscribed and sworn to before me, a Notary Public in and for the State and County above named, this 11th day of January,1980.
Notary Public Notary Public, State of Florida at Large, f4/ Commission Expires: August 8, 1983 CameronNotary 3(D12) -
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FLORIDA POWER CORPORATION RESPONSE TO NUREG-0578 SHORT-TERM RECOMMENDATIONS FOR CRYSTAL RIVER UNIT 3 Supplemental Response January 11, 1980 0755 009
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FLORIDA POWER CORPORATI6N'S RESPONS TO NUREG-0578 Recommendation 2.1.1 Emergency Power Supply Pressurizer Heater Emergency Power Supply After further review of the NUREG-0578 recommendations and subsequent clarification for Section 2.1.1, Florida Power Corporation has deter-mined that the existing design satisfies the NUREG recommendations with only one exception. This being that the transfer of the heaters from the normal power source to the emergency power source can not be accomplished completely within the control room. This exception is justified with the fact that after the accident, B&W has determined that the operator has 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to accomplish the connection of the pre-selected pressurizer heaters of sufficier.t capacity (126 kW) to initi-ate and maintair, natural circulation.
A procedure (EP-101, Unit Blackout) is available to the operators which will allow the connection of the preselected heaters to the Engineered Safeguards (Safety-related) Bus during a loss of offsite power. This will be accomplished by utilizing the existing cross-tie breakers and assuring that all nonessential loads are disconnected from the respective buses. This method meets the intent of the NUREt requirements with the exception that the manual transfer of heater breakers is not entirely accomplished in the Control Room. Some of the disconnections of the nonessential loads may have to be accom-plished at the local power center. Load consideration is given in this procedure, to prevent overloading a diesel generator.
Pressurizer Level and Pressurizer Relief and Block Valves Emergency Power Supplies The existing design satisfies the requirements of NUREG-0578 for the power supplies for the Pressurizer Level Indicators and Pressurizer Relief and B1ock Valves, as follows:
- 1. The motive and control components for the Relief Valve are pow-ered from the on-site DC power system.
- 2. The motive and control components for the Block Valve are powered from the AC emerge 1cy power supply (Engineered Safeguards Bus).
- 3. The pressurizer level indication instrument channels are powered from the vital instrument buses (Inverters).
- 4. As noted in 1 and 2 above, the power for the Block Valve is sup-plied from a different bus from that which supplies the Relief Valve.
- 5. The motive and control power connections to the emergency buses are through safety-grade devices.
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- 6. The manual transfer of power from the normal power .o the emer-gency power is not applicable to the design. As ..oted in 1 above, the Relief Valve is normally powered from the on-site DC power system, therefore, no transfer is required. As noted in 2 above, the Block Valve is nomally powered from the Engineered Safeguards Bus, which is normally powered fra an offsite source.
On a Loss of Offsite Power Event, the Emergency Diesel Generators will automatically pick up the Engineered Safeguards Bus and the sa fety-rel ated loads connected to it. This includes the ES 3AB MCC, which feeds the Block Valve, therefore, a manual transfer is not applicable.
Recommendation 2.1.2 PWR Relief and Safety Valve Testing By letter dated December 17, 1979, Mr. William J. Cahill, Jr., Chair-man of the EPRI Safety and Analysis Task Force submitted " Program Safety / Relief Valves and Systems," December 13, 1979.
Florida Power Corporation considers the program to be responsive to the requirements presented in Section 2.1.2 of NUREG-0578. The EPRI Program Plan provides for a completion of the essential portions of the test program by July 1981. Florida Power Corporation will be par-ticipating in the EPRI program to provide technical review and to sup-ply plant specific data as required.
Recommendation 2.1.3.a Response to Direct Indication of Power-Operated Relief Valve and Safety Valve Position for PWRs and BWRs In direct response to NUREG-0578, Item 2.1.3.a. Florida Power Corpora-tion has purchased from Babcock & Wilcox a Valve Monitoring System.
This system incorporates acoustical monitoring techniques to provide the reactor operator with indication of valve open/ closed position.
The equipment is very similar to the existing Loose Parts Monitoring System supplied by Babcock & Wilcox. The engineering design for in-stallation of this equipment is proceeding on an expedited basis to meet the specified inservice date.
This design provides for two transducers mounted on each safety valve and the PORV. Each of these transducers will be wired out of the con-tainment to the PORV/Tsat monitoring cabinet, to be located in the 4160 V ESB SWGR Room. Within thic cabinet will be three channels (one for each valve) of signal conditioning with local indication, alarm (high and low), and selectable aJdio monitor. Only one transducer will be aormally monitored on each valve. The other is manually se-lectable for comparison of perfomance or in the event of transducer fail ure. Each channel will also provide remote analog indication and annunciator events recorder high alarm functions. This analog indica-tor for each channel will be mounted on the PSA section of the main control board. A common annunciator window will be located on the ICS section of the main control board. The events recorder will provic'e CRT and hard copy indication of valves that actuate.
The valve monitoring Tsat cabinet will be powered from a vit I source with all cable routing meeting seismic requirements. Seismic testing ,
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of equipment identical to that used in the Babcock & Wilcox Valve Mon-itoring System has been performed. Environmental qualification of in-containment equipment has been satisfied by similar equipment test and survival of TMI-2 equipment.
The present equipment delivery schedule for the above-described modi-ficatica will not pennit cmplete installation, by February 15, 1980.
The accelerometers, which were manufactured by N. Devco, were received on-site on Januar, . 1980. The visual indicators (edge meters) for the main ccctrol board are being manufactured by International Instru-ments, and are presently scheduled to be shipped from Babcock & Wilcox to Crystal River Unit 3 on January 25, 1980. The mounting cabinets are being manJfactured by Hoffman Engineering Company, and are pres-ently scheduled to be shipped from Babcock & Wilcox to Crystal River Unit 3 on February 22, 1980.
In accordance with Section III of the Commission's Show Cause Order for Crystal River Unit 3, dated January 2,1980, the above modifica-tion to provide direct position indication in the Control Room for the PORV and safety valves at CR-3, will be completely installed within 30 days after receipt of all necessary equipment at CR-3. This in-stallation will be completed no later than June 1,1980. The install-ation of this equipment will require a unit shutdown of approximately 10 days.
Recommendation 2.1.3.b Instrumentation for Inadequate Core Cooling In responce to NUREG-0578, the Babcock & Wilcox Owners' Group has de-veloped an extensive program for inadequate core cooling which has been discussed with the Bulletins and Orders Task Force. In addition, at the request of the Bulletins and Orders Task Force, the program has been expanded beyond the requirements of NUREG-0578. The objectives of this program are as follows:
- 1. Develop operating guidelines that will allow the reactor operator to recognize and respond to conditions of inadequate core cooling under the following conditions:
- a. Power Operation with portions of the core in DNB.
- b. Loss of RCS Inventory without the reactor coolant pumps ope rating.
- c. Loss of RCS Inventory with the reactor coolant pumps operating.
- d. Loss of the Decay Heat Removal System and Loss of RCS Inven-tory During Refueling Operations.
- e. Loss of natural circulation due to loss of heat sink.
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- 2. Provide re.ommendations for any additional instrumentation re-quired i.o indicate inadequate core ccoling under the conditions listed above. Included with the recommendations will be:
- a. A description of the functional design requirements for the additional instrumentation.
- b. A description of the Opera;ing Guidelines to be used with the proposed equipment.
- c. A description of the analyses used in devel oping these guidelines.
- d. Installation schedules for additional instrumentation.
To-date, Operating Guidelines and supportive analyses are complete for the following conditions within the scope of the Inadequate Core Cool-ing Program:
- 1. Loss of RCS Inventory without the reactor coolant mps operating.
- 2. Loss of RCS Inventory with the reactor coolant pumps operating.
- 3. Loss of natural circulation due to a loss of heat sink.
These guidelines and supportive analyses have been submitted to the NRC by Florida Power Corporation in response to IE Bulletin 79-0SC, dated November 14, 1979. Florida Power Corporation has revised Plant Procedures to incorporate these new guidelines and has implemented op-erator training related to the inadequate core cooling. This activity is complete.
Additional guidelines / support analyses for refueling operations have been performed by Babcock & Wilcox and are presently being reviewed by Florida Power Corporation. Copies of the B&W guidelines and suppor-tive analyses will be submitted to the Bulletins and Orders Task Force. Following our review, we will develop the necessary procedures and implement operator training. This effort will be completed by April 1, 1980.
On January 4,1980, Florida Power Corporation received from B&W their analyses and guideline reconnendations for inadequate core cooling due to DNB at power. As a result of these evaluations, no substantive changes to existing operating procedures are necessary. The investi-gations indicated that to obtain inadequate core cooling at power, the operators would need to ignore numerous existing alanns or major non-mechanistic damage to reactor internals needs to occur. Copies of these analyses and guideline recommendations will be submitted to the Bulletins and Orders Task Force.
Babcock & Wilcox is scheduled to submit to Florida Power Corporation a final report containing recommendations for additional instrumentation in late January,1980. FPC will submit this infonnation as soon as 1755 013
possible to you, following our review. Every effort will be made to install this new instrumentation by January 1,1981, subject to equip-ment availablity and NRC review.
Subcooling Meter In direct response to NUREG-0578, Item 2.1.3.b, Florida Power Corpora-tion has purchased from Babcock & Wilcox two saturation meters and a field change package to provide wide-range hT ot input to these devic-es; Tcold wide-range is currently available. This meter was designed by Babcock & Wilcox to monitor plant temperature and pressure and im-plement, with hard wire logic, tha detemination of margin to satura-tien for present plant conditions and indicate this to the plant oper-ator. The engineering design for installa; ion of this equipment is p aceeding on an expedited basis to meet the specified inservice date.
T:is design provides for two T9t meters to be mounted in the PORV/
Tsat monitoring cabinet that will be located in the 4160 V ESB SWGR Room. Each meter will receive the following inputs:
4 hot leg temperature (2 per loop) 120 F to 920 F 2 RC pressure (1 per loop) 0 to 2500 psig These signals will be taken from the Non-Nuclear instrumentation (NNI)
System, with individual buffers to preclude interaction between Tsat meters or NNI/ICS. The temperature inputs are not qualified safety-grade, however, they are reliable in that this NNI provides two vital sources and signal cables are routed in seismic instrument trays.
Each meter will have a remote digital indicator / selector, mounted on the PSA section of the main control board, and a low margin to satura-tion alam to the annunciator events recorder. The low mars.n to sat-uration alarms will light a common window on the PSA section of the control board with CRT and hard copy events recorder identification of loop indicating the condition. The digital indicator on the control board will have a spring return selector switch such that ene meter is normally looking at Loop A, and the other is looking at 1.oop B, with the capability to switch for checking perfomance and, in the event of meter failure, the power to each meter will be from different vital sources.
Attachments I and II provide additional infomation concerning the de-sign of the subcooling meters.
The Construction Work Package has been issued to the site for the Tsat Meter modification, and the equipment is on-site, with the exception of the PORV/Tsat monitoring c^inet (see Item 2.1.3.a).
Recommendation 2.1.4 Diverse Coatj t Isolation 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 1755 014
capabilita and, therefore, should not be isolated on automatic HPI ac-tuation. For the valves listed in our April 12 response, which re-ceive no ES signal and are nonnally closed and remain closed following the accident 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 with the addition of an auto-close isolation signal, based on automatic HPI actuation. These di-verse containment isolation signals will satisfy safety-grade require-ments and resetting of these signals shall not result in the automatic loss of containment isolation.
The Construction Work Package was issued to the site on January 11, 1980. Installation of this modification will require approximately 4-5 weeks, as it will be accaaplished with the unit on-line. This modi-fication will be installed and tested on or before February 15, 1980.
Recommendation 2.1.5.a Dedicated Penetrations for External Recombiners or Post-Accident Purge Systems The present CR-3 design has installed a redundant, dedicated, hydrogen purge system. The CR-3 system uses two penetrations, dedicated to hy-drogen purge only, which are sized consistent with the flow require-ments of the purge system. The CR-3 purge system is single failure proof. Therefore, he conclude that the existing hydrogen purge system satisfies the requirements of Section 2.1.5.a of NUREG-0578.
The hydrogen purge system is described in Section 14B of the CR-3 FSAR and shown on F1ow Diagram FD-302-722.
Recommendation 2.1.5.c Recombiner Procedure CR-3 does not have a requirement for hydrogen recombiners as a design basis for licensing. Therefore, this requirement does not apply to CR-3.
Recommendation 2.1.6.a Integrity of Systems Outside Containment Likely to Contain Radioactive Material for PWRs and BWRs Prior to issuance of the NUREG-0578 requirements, FPC had a leak re-duction program implemented to satisfy the leakage rate requirement identified in the CR-3 Technical Specifications. This program is de-scribed and implemented by SP-317--RC System Water Inventory Baiance.
Since receipt of NUREG-0578, the CR-3 program has been expanded t0 meet these new requirements beyond the CR-3 Technical Specifications.
The present program includes the following systems:
- 1. RC Bleed Line
- 2. Waste Gas Disposal System ,
- 3. Decay Heat
- 4. Building Spray
- 5. Make Up 1755 015
- 6. High Pressure Injection
- 7. RCS Sample Lines.
This leak reduction program is described in SP-716--f 3...ple Line Leak Rate Test, SP-412--ECCS and Pontainment Spray System Leak Rate Test, SP-429--Waste Gas System Leak Rate Test, PT-108--Decay Heat Removal and Reactor Building Spray System Leak Rate Test, and PT-10)--RC Bleed Line Leak Rate Test. Copies of these procedures were submitted on No-vember 17, 1979, for your review, except for SP-716, which is being written to include some additional sample lines in the program. A copy of SP-716 is being written in place of SP-317 for this program, and will be submitted as soon as it is inued. To-date , all of the above systems have been leak tested except for the Decay Heat and Building Spray Systems and the Wase Gas Storage Tanks. We are pres-ently repairing leaks in the Decay Heat System at CR-3. Upon receipt of parts to complete this et fort and the leaks are repaired, the Decay Heat and Building Spray Systems will be leak tested together. The leak test of the Waste Gas Storage Tanks will require a 2-week outage, and will be performed during our April,1980, refueling outage. Re-sults of these additional leak tests will be submitted upon completion of this effort.
Recommendation 2.1.6.b Design Review of Plant Shielding and Environmental Qualifications of Equipment for Spaces / Systems Which May Be Used in Post-accident Operation Enclosed as Attachment III is the GAI Report prepared for Florida Power Corporation, entitl ed " Design Review of Pl ant Shielding and Environmental Qualification of Equipment for Spaces / Systems Which May Be Used in Post-accident Operations Outside Containment at Crystal River Unit 3 Nuclear Generating Station". This Report, which is under review by Florida Power Corporation, is being submitted in response to the September 13, 1979, letter from Darrel G. Eisenhut, requesting implementation of NUREG-0578, Item 2.1.6.b, and clarified in Mr. Harold R. Denton's letter of October 30, 1979. Following our review, Eqy modifications to this report will be submitted as soon as possible.
Recommendation 2.1.7.a Auto-Initiation of Auxiliary Feedwater System AFWS This re:.ommendation has been excluded from NUREG-0578 and will be ad-hsed by the Bulletins and Orders Task Force.
Re:ommendation 2.1.7.b Auxiliary Feedwater Flow Indication to Steam
@1erators
- 1. Short-term Control Grade:
- a. Emergency feedwater flow indication to each steam generator satisfies the single failure criteria because there are ul-tra sonic fiow indications on each steam generator with a backup steam generator level indication on each steam generator.
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- b. The present ultrasonic flow indication channels are testable by electronically verifying the zero and circuit fault con-ditions for each unit.
- c. The present emergency feedwater flow indicating devices are powered from vital buses with a battery-backed inverter.
- 2. Long-term Safety-Grade:
- a. We are in the process of evaluating the present equipment for upgrading to safety-grade, as well as evaluating other alternate methods of emergency feedwater flow measurement.
- 3. Other:
- a. The short-term control grade flow indication channels satis-fy the single failure criteria because each steam generator has an ultrasonic flow indicator and steam generator level indication.
- b. Ultrasonic flow indicators were factory-cal ibrated as a matched system (transducers and flow display computers) at 740 gpm and has an accuracy of about 2L Recommendation 2.1.8.a Improved Post-accident Sampling Capability We are currently conducting a design and procedure review regarding post-accident sampling at CR-3. Florida Power Corporation has hired Applied Physics Technology (APT) and GAI to assist us with this ef-fort. These reviews and a report describing the review and corrective actions will be submitted to you as soon as possible, but no later than Feburary 15, 1980.
Recommendation 2.1.8.b Increased Range of Radiation Monitors We are presently developing interim procedures for the estimation of high level accidential radioactive releases, if instrumentation goes of f-scal e. Additional infonnation concerning this effort will be sub-mitted as soon as possible. We are utilizing APT and GAI for this item al so. Completion of this effort is scheduled for on or before February 15, 1980.
Recomendation 2.1.8.c Improved In-Plant Iodine Instrumentation CR-3 presently has six portable air samplers and procedures in place for obtaining and detennini ng airborne iodine concentrations using spectral analyses. Therefore, we currently satisfy the requirements of this section.
Recommendation 2.1.9 Transient and Accident Analyses Task Description Status
- 1. Small Break LOCA analysis and prepara- Complete tion of emergency procedure guidelines 1755 017
Task Description Status
- 2. Implementation of Small Break LOCA Coaplete emergency procedures and retraining of operators
- 3. Analysis of inadequate core cooling See response to Rec-and preparation of emergency procedure ommendation 2.1.3.b guidelines
- 4. Implementation of emergency procedures See response to Rec-and retraining related to inadequate anmendation 2.1.3.b core cooling
- 5. Analysis of accidents and transients late 1980 and preparation of emergency procedure guidelines
- 6. Implementation of emergency procedures 6 months after guide-and retraining related to accidents lines established and transients
- 7. Analysis of LOFT small break tests
- By letter dated December 31, 1979, J. H. Tayl or, Manager, Licensing, Babcock & Wilcox, provided the "B&W LOFT L31 Pretest Prediction Report" to the NRC Staff.
RCS Venting Enclosed as Attachment IV, is the B&W Report, entitled "177 High Point Vent Design Criteria".
This P;p]rt provides the design criteria and other design input which will be used by Florida Power Corporation and our A/E to complete the design. Included are:
- 1. Venting System Design Criteria: This includes the requirement that the vent piping and valving be sized (approximately 1/2")
such that failure of a line does not cause coolant loss in excess of normal makeup system capacity. The actual sizing requirements will be a direct function of the CR-3 actual makeup rate.
- 2. Venting System Schematic.
- 3. Vent Flow Rate Curves, showing mass and volume flow rates as a function of system resistance, since the actual resistance will be determined by the hardware installed.
- 4. Anchor Nbtions, Seismic Response Spectra, and Allowable Nozzle Loads.
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The B&W requirements for installation of high point vents on the Reactor Coolant System are summarized as follows:
- 1. Remotely operable high point vents are required at the top of each hot leg and on tor of the pressurizer.
- 2. The vent line size (approximately 1/2") will be such that the flow rate will be less than the capacity of a makeup pump, thus eliminating a LOCA analysis.
- 3. The discharge will be to the containment.
- 4. Each vent line will have two isolation valves powered from the same IE power supply and operated remotely by separate switches.
The Design Criteria are such that a LOCA analysis is not required.
The criteria limit the size of the vent piping so nonnal makeup capa-bility can accommodate the outflow of a break in the vent 1;nc. This is consistent with the B&W position on breaks in instrument lines.
In addition to the two new high point reactor coolant vents, the gases that accumulate in the pressurizer can be vented by use of the PORV presently installed on the pressurizer. We do not propose to add a remotely operated vent to the reactor head, since any accumulation of gases sufficient to fill the reactor vessel volume will be vented via the hot leg vents, due to the free path available.
Florida Power Corporation is presently review lng this B&W report, and will, with the assistznce of GAI, complete the final design of this modification. This final design work will include the designation of appropriate power supplies to the vent val ves , piping routing, and hydrogen gas concentration limits. Additional infonnation concerning the detailed design and schslule for procurement and installation will be submitted as soon as possible.
Recommendation 2.2.1.a Shift Supervisor Responsibilties The responsibilities of the Shift Supervisor have been defined in Administrative Instruction AI-500, Conduct of Operations and Manage-ment Directives. Copies of AI-500 and our Management Directive con-cerning the Shift Supervisor are enclosed in Attachment V and VI, re-spectively. Florida Power Corporation considers this item completed.
Recommendation 2.2.1.b Shift Technical Advisor The on-shift Technical Ad. isor to the Shift Supervisor will be provid-ed as follows: -
- 1. Short-term Plan:
By J6nuary 1,1980, Shift Technical Advisors (STAS) are on-shift to provide accident assessment. The compliment is provided by current plant personnel who meet the requirements identified .in 1755 019
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paragraphs A.1, 2, and 3 of Enclosure 2 of Darrel G. Eisenhut's letter of September 13, 1979.
The Shift Technical Advisors are assigned for 24-hour periods, are on the plant site, and will remain within 10 minutes of the Control Center. The STAS will be in the Control Center for plan-ned major plant evaluations. They are independent from supervi-sion of the manipuiation of plant controls. In addition to per-fonning the function of the STA, they will be perfonning the functions of their normal positions. To provide the operating experience assessment function, contract personnel are located on-site, dedicated to evaluation of plant operations for poten-tial safety implications. The STAS and contract personnel will provide feedback to one another, on a current basis, on the na-ture and results of their assessments. This plan will be in ef-fect until the necessary personnel can be hired and trained.
- 2. Long-term Plan:
Establish Nuclear Operations Technical Advisor positions to pro-vide both the accident assessment and operating experience func-tions. These Advisors would meet the intent of the requirements iden:ified in Encl osure 2 of Darrel G. Eisenhut's letter of September 13, 1979.
As of October 29, 1979, the Nuclear Operations Technical Advisor positions were approved. The position descriptions are developed and recruitment of personnel is underway. The Nuclear Operations Technical Advisors will undergo training to bring them up to the level of expertise required prior to being placed on-shift to perform the STA function.
Recommendation 2.2.1.c Shift and Relief Turnover Procedures The shift and relief turnover procedure; are defined in Enclosures 9, 10, and 11 of Administrative Instructions AI-500, Conduct of Opera-tions (copy enclosed). Florida Power Corporation considers this item completed.
Recommendation 2.2.2.a Control Room Access Access to the CR-3 Control Room is handled in accordance with AI-500 and AI-200, Organization and Responsibility ( Attachment VII). Florida Power Corporation considers this item completed.
Recomendation 2.2.2.b On-Site Technical Support Center A Technical Support Center (TSC) is established in the office building located on the northwest corner of the turbine building. This is the normal storage and retrieval area for those drawings and records de-scribed in ANSI N45.2.9-1974. The TSC will provide assistance to the operating personnel in evaluating the course of an incident or acci-dent and will also be the designated point of contact with offsi.te agencies (after activation), in providing advice on . the expected 1755020
course of the acident. This area cannot be designated as the penna-nent TSC, due to the requirement for the TSC to be habitable to the same degree as the Control Room for postulated accident conditions.
The designated temporary location is habitable, provided with two con-ference rooms, capable of supporting 15 - 20 assigned personnel. Po r-table monitoring equipment for measuring radiation levels in the TSC is provided. Action level criteria (EM-102) have been developed to define when protective measures (evacuation to the Control Room) should be taken.
Dedicated telephone communications have been provided to allow relia-ble communications between the TSC and Control Room, and the NRC.
Dedicated telephone conmunications will be provided between the TSC and the Emergency Operations Center (EOC), once the E0C is estab-lished, prior to mid-1980. In addition, nondedicated telephone lines and interplant communication systems are avail able for additional communications.
Plant parameters necessary for assessment have been provided by a com-puter printout, located in the TSC and paralleled with the control room printer.
Plans for staffing the TSC during emergency situations, and for per-fonning this accident assessment function from the control room should the TSC become uninhabitable, are developed and will be implemented by January 18, 1980. These plans are covered by EM-102.
Attachment VIII to this letter provides Florida Power Corporation's general long-range plant to upgrade the TSC at CR-3. Additional de-tails will be submitted at a later date.
Recommendation 2.2.2.c On-site Operational Support Center In order to provide an on-site assembly area where assigned support personnel will report in the event of an accident or emergency situa-tion the following actions have been taken by CR-3:
- a. The north end of the shop facilities building, located northeast of the control complex, has been designated as the Operational Support Center (OSC). This choice of locations allows ready ac-cess to the control complex and utilizes the existing control complex personnel radiation shielding to reduce potential radia-tion exposures during accident conditions which require manning of the OSC.
- b. The designated location is habitable and provided with washroom and facilities to support 25 assigned personnel.
- c. The interplant communication system is provided to allow communi-cation between the OSC, Control Room, and Technical Support Center.
NUREG-0578(1/11/80 )DN-94 ,
1755 021
ATTACHMENT I INFORMATION REQl' IRED ON THE SUBC00 LING METER Display No mal-T-Tsat Information Displayed (T-Tsat, Tsat, Press, etc.) Manual-P-Psat Display Type (Analog, Digital, CRT) Digital Continuous or On Demand Continuous 1/ Loop A Single or Redundant Display Redundant 1/ Loop B Location of Display Main Control Board Alarms (include setpoints) T-Tsat (selectable) 4*F, --
Overall uncertaigy (*F, PSI)
Range of Display 4096 Qualifications (seismic, environmental, IEEE323) See Attachment Calculator Type (process computer, dedicated digital, or analog calc.) Dedicated Digital If process conputer is used specify availabilty
(% of time)
Single or redundant calculators Single meter Selection Logic (highest T., lowest press) Manual Qualifications (seismic, environmental, IEEE323) See Attachment Calculational Technique (Steam Tables, Functional Fit, ranges) Steam Tables Input Temperature (RTD's or T/C's) .RTD's Temperature (number of sensors and locations) 8, 2Rx Out/ Loop +2R x In/ Loop Range of temperature sensors Ry Out 120'-190',
Rx In 50* to 650*
1759 n??
Uncertainty
- of temperature sensors (*F at 1)
Qualifications (seismic, environmental, IEEE323) See Attachment Pressure (specify instrument used) RC-3A + 3B - PT3 Pressure (number of sensors and locations) 2, one Ry Outlet / Loop Range of Pressure sensors 0-2500 psig Uncertainty
- of pressure sensors (PSI at 1)
Qualifications (seismic, environmental, IEEE323) PT qualified to Engi-neered safeguards system requirements described in the FSAR Backup Capability t.1 tilability of Temp & Press Two meters, loop A&B, with manual selection opposite loop.
Availability of Steam Tables, etc. } Covered by
{
small break Training of operators operating
( procedures &
Procedures J training
- Uncertainties must address conditions of forced flow and natural circulation.
i755 023 NUREG-0578*(RevAtchl)DN-94
- ATTACHMENT II, ,
. SABC0CK & Mitc0K SATURA) ION METERmm m y-
- IMPl.EMENTA3!DH 0F IlftC REQUIRIMENTS .- ww s S. =
eter, designed by B&W to meet the requirements of NtfREG-De Saturation '
O$78. eccepts existing inputs of plant temperature and pressure, looks up in
- a rtored stem table the values of saturation tagerature and pressure and detemines margin to saturation conditions from present plant conditions. This astgin is displayed to the power plant operator on a digital panel meter.
Gat 4l ration conditions are indicated by the meter displaying zero (00) margin.
~
Dehrees of superheat are displayed as a negative margin should tegerature asceed T sat" ,
Method Of leptementation The values A block diagram of BW's Saturation Meter is shown in figure 1.
of T,,g and Psat im ne 1967 N steam tables have been stored in non-voittile semiennductor read only memory. Incoming analog pressure and temperature signals are converted to digital signals using single diip analog to digital (4/D) converters. The output or digital signal frca the A/D converters addresses
- a read only memory in which the Saturation values fram the steam tables are
- stored. Thus. the incoming signals from the plant *look up" a Saturation Valve frem a stored table. The current temperature is then digitally subtracted from
- the value of Saturation temperature to give T,,g margin and P,,g is subtracted free the current pressure to give margin or Tsat 3 P ,g margin. NPsat margin is disp 1ged on a digital meter depending on operatar selection.
It should be noted that'B&W's taplementation of the Saturation Meter does
' mot use a microporcessor or other pmgramable logic device. It is impleaented with hard wired logic. The only " calculation" done is the subtraction to determine margin. This is also done with hard wired logic.. r Stace the meter uses hard wired logic, its operation is easy to confirm that it is operating properly. In addition, every signal in the meter can be a
tested as discussed further below.
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. Compliance with Requirement _s_ .-
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In the regfonal NRC meeting in Atlanta. Georgia on September.28.1979 (tscussicas were held concerning NRC requirements for implementation of IMtEfr0578, and particularly, the'NRC definition of safety grade as related to.certain contml mon instrimentation such as the Saturatica Meter. The
, 15tc requirements as stated in this meeting and B&W's response to meet these is summarized below: .
A. IIRC Requirement: Single rallure Proof ,
-The design of the T,,g Meter considered the most likely failure to be loss of the internal clock. A circuit was incomorated which will detect loss of the clock'at the remote display and will blank the display.
-The capability was incorporated in the design to switch to alternate input sensors should a sensor fall. In addition. isolation resistors la series with each input prevent a failum in one input from affecting
. ' ' ~ - - ' - "
d r.
' ' ' - ~
-BIN 1.s recommending t5at redundad powe[ sources be provided to the Saturstion Meter. -
r8&W is V1Kosmending that redundant meters be installed, or that operator ansultation of steam tables be used as ' redundancy. -
- 8. ItitC Requirement: System $ hall be Testable Stace the B&W T,,g, Meter does not use a sferocorputer or any other program-mable calculating machine. Input signals simply " loot-un' saturation values
, in a read only merary, and margin is calculated in a hani wired logic adder. A test display m>dule is included in the equfpment rack with each Tsat meter. 'The test module has a digital display which allows an operator or fechniclan to check every signal internal to the meter. These include the input temperatures and pressures in 'F and Psts, the satu 1755 0
om o -
JeoM e ju 1 A _=
l .
,,, .. and the _teimperature that the input , signals ,look up and.the margin to
~
T,,g and margin to:N,,g'as detemined by the meter. Since the meter is implemented with hard wired logic, there are no other intermediate answers calculated internal to the machine. Thus the capability td check every signal in the meter has been provided. ' -
C. Amlification - Seismic A shaker test was performed on the Saturation Meter. EacF. ci kuft board was instrumented with an accelerometer and the entire assewbly was excited at a level of 1 g, sinusoidal. The frequency of excitation was increased from S Hz to 33 liz at a rate of 1 octave per minute to search for resonances. N6ne were found. The excitation uss held at 33 Hz for c.~
5 minutes. The assembly was then excited at 10 g peak acceleration -
at 20 Hz for 5 minutes duration. .
No damage was done to the seter, and testing subsequent to the shaker test confirmed that it functioned properly.
D. Dualification - Envimrmatal_ -
, R+diation - The3 T ,g meter was MgnN to actept W % esting.
~
~
. plant iristrumentation. Qualification o'r'these instruments! to 1ET ~ ~
requirements including radiation has already been performed. The equipment supplied'by BW is for installation in an electronic equipment ' room and the
' control room. No radiation testing has been performed. 1755 027 Temperature - Analog..".ircuitry in the T sat meter has been tested to 60*C
. With results as shown in figure 2. An error analysis performed using worst case errors in the analog circuitry plus quantitation errors in the digital circuitry showed a worst case error of f4'F. That is, the value of T,,g
y --
b D.
{ *
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margin displayed to the operator could be in error 34'T from a value he tould obtain by looking up T,,g for the . indicated control room pressure
. and subtract.idg the indicated temperature. .
For the digital circuitry all components used are rated at 60'C or
- higher.
Hinsitity - fio humidity tests were performed. All components used are hennetically sealed. Specific care was.taken in the design of the meter
~
~o s eliminate components such as carbon resistors whose values are sensitisie to humidity The only resistors used are used for digital logic pull up where exact values are not critical.
SG W eesm enee e see m ese .eoes se este eh t oma es s damentatert e ger . se n . . e d- . . , _ , . _, etgigga ag .. e . > g ,, e p , ep, ye g y g g. . , gg ,,esg o. gge,,m em e este e
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.1755 028
ATTACHMENT III DESIGN REVIEW OF PLANT SHIELDING AND ENVIRONMENTAL QUALIFICATION OF EQUIPMENT FOR SPACES /SYSTDIS WHICH MAY BE USED IN POST ACCIDENT OPERATIONS OUTSIDE CONTAINMENT AT CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING STATION PREPARED FOR FLORIDA POWER CORPORATION
)
PREPARED BY GILBERT /C0te10NWEALTH GILBERT ASSOCIATES, INC. '
READING, PENNSYLVANIA DECEMBER 1979 jfhrj 099 c
1.0 INTP.0 DUCTION This report has been prepared in response to the September 13, 1979, lett7 from Darrel G. Eisenhut of the NRC to all operating nuclear power plants.
This letter presented an implementation schedule for the recommendations presented in "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations" (NUREG-0578) . The requirements defined in the September 13th letter were subsequently clarified in a letter from Harold R. Denton to all operating nuclear power plants dated October 30, 1979.
Among the requiremerts defined in the two NRC letters is a review to determine whether post-accident radiation fields unduly limit personnel access to areas necessary for mitigation of or recovery from an accident 7r unduly degrade the proper operation of safety equipment. Corrective actions for problems identified as a result of the review are also to be determined. This report presents the results of such a review for the Crystal River Unit 3 Nuclear Generating Station.
The review was based on the following guidelines:
a) The post accident dose rate in areas requiring continuous occupancy should not exceed 15 mr/hr.
b) The post accident dose rate in areas which do not require continuous occupancy should be such that the dose to an individual during a required access period is less than 5 Rem whoJ7 body or its equivalent.
c) The integrated dose to safety equipment should be less than the dose that the equipment has been qualified for.
d) The minimum radioactive term used in the evaluation should be equivalent to the source terms recommended in Regulatory Guides 1.3 and 1.4.
1755 030
2.0 SUtHARY Reviews have been performed of: a) the areas requiring access for post-accident operations; and b) the radiation qualification of safety equipment outside containment.
The areas identified as requiring access are:
nuclear sample room hydrogen purge equipment containment air monitor RM-A6 diesel generator room radioactive waste disposal control board control room radiochemistry laboratory count room Considerations associated with the nuclear sample room, radiochemistry laboratory, and count room are addressed in the response to NRC Lessons Learned Task Force Short Term Recommendation 2.1.8.a (Improved Post-Accident Sampling Capability). The evaluations performed for the other areas indicate that with the exception of the radioactive waste disposal control board access is possible for the estimated occupancy. The radioactive waste disposal control board will be moved to a low radiation area to assure access to it.
An additional result of the review is that certain modifications are planned which will make it unnecessary to have post-accident access to certain areas. These modifications include:
o Changing valves DHV-7, 8, 39 and 40 to motor operated valves.
o Adding a motor operated bypass valve for the Makeup and Purification System prefilter.
1755 031
It is not possible to reach conclusions as to the acceptablility of the integrated doses calculated during the review of the radiation qualification of safety equipment. This is due to the lack of available data on:
- 1) qualification doses; and 2) the period of time post accident that safety equipment must be functional. Work is continuing to resolve these concerns.
The period of time post accident that each item of safety equipment must be functional will 'be defined making it possible to calculate an integrated dose for the time period appropriate for the function of the equipment.
These doses will then be comparad to qualification doses where available.
For safety equipmert with no available qualification data, the feasibility of developing the data or prosiding appropriate dose mitigating measures (shielding, equipment relocation, etc.) will be pursued.
1755 032
3.0 METHODS 3.1 SOURCE TERMS The activity assumed for liquid source term calculation is based on 100% of the noble gas inventory, 50% of the halogen core inventory and 1% of all other nuclides in the core inventory. The activity assumed for gaseous source term calculation is based on 100% of the noble gas core inventory and 25% of the halogen core inventory.
Two liquid source terms were used in the evaluation. For systems which contain post accident recirculation fluid, the source term was based on diluting the liquid inventory discussed in the previous paragraph with the expected volume of fluid in the bottom of the reactor building post accident.
For systems which can contain fluid from the reactor coolant system but do not take suction on the recirculation sump, the source term was based on diluting the liquid inventory discussed in the previous paragraph with the volume of fluid in the reactor coolant system.
Gaseous source terms were determined for containment and for the waste gas system. The containment airborne source term was based on diluting the gaseous inventory discussed previously with the air contained in the containment free volume. The vaste gas system source term was determined by taking 100% of the noble gases and 50% of the halogens in the makeup tank, assuming it to be full of reactor coolant as discussed above, and diluting that inventory in one-h,21f the makeup tank volume.
Table 3-1 presents the inventories and source terms discussed above for the time period immediately after the postulated accident (T=0). For other time periods, the decay paramotors given in Refs. I and 2 were used to adjust the source terms for radioactive decay.
1755 033 G
3.2 CALCULATION OF DOSE RATES Dose rates for the areas of interest in this review were calculated by determining the potential contributing sources at a representative
, location. The appropriate source term data for these sources was selected from Table 3-1 and adjusted for decaf as required. The dose rate at the representative location was used as the general area dose rate. Both the SDC Code (Ref. 3) and the Gilbert / Commonwealth developed SPOTl Code were used in performing the dose rate calculations. The SPOTl Code uses the methodology originally presented by Ono and Tsuruo (Ref 4) and developed by Shure and Wallace (Ref. 5). Energy groups required as input to the codes were determined using, the gamma ray energy and intensity data in Refs. I and 2 for the nuclides in Table 3-1.
3.3 CALCULATION OF DOSES TO PERSONNEL DURING POST ACCIDENT ACCESS TO VITAL AREAS Personnel doses received in performing a given operation in a given vital area are calculated as the sum of the doses received during travel to and f rom the vital area and the dose received while performing the given operation in the vital area.
The doses received during travel are determined by calculating dose rates at selected locations (or at a single location if the dose rate along the travel route is relatively uniform) along the travel route using the methodology discussed in Section 3.2. and multiplying the dose rates by The appropriate travel time for each selected location along the travel routt.
Doses received while performing a given operation are determined by multiplying the dose rate for the given area by the time required to perform the operation.
Dose rates for the given vital area are determined using the methodology discussed in Section 3.2.
. 1755'034
3.4 CALCULATION OF INTEGRATED DOSES TO SAFETY EQUIPMENT The in'.agrated dose to a given item of safety equipment is determined by integrating the dose rate appropriate for the given item over the time period that it is required to be available to perform its safety function.
Dose rates are calculated using the methodology disucssed in Section 3.2.
1755 035
b Table 3-1 t' i
Shielding Source Terms (T=0)
Liquid ( Caseous ( Containment Containment Source Source Sump Reactor Coolant Airborne Waste Gas Activity Activity Concentration Concentration Concentration Concentration Isotope (Ci) (Ci) (uci/cc) (uci/cc) (UCi/cc) (uci/cc) ;
Be-84 7.85 + 6 3.93 + 6 4.24 + 3 2.43 + 4 6.94 + 1 2.44 + 4 Kr-83m 9.25 + 6 9.25 + 6 4.99 + 3 2.87 + 4 1.63 + 2 5.74 + 4
~
Kr-85m 2.19 + 7 2.19 + 7 1.18 + 4 6.79 + 4 3.87 + 2 1.35 + 5 Kr-85 5.30 + 5 5.30 + 5 2.86 + 2 1.64 + 3 9.36 + 0 3.28 + 3 Kr-87 4.00 + 7 4.00 + 7 2.16 + 4 1.24 + 5 7.06 + 2 2.48 + 5 Kr-88 5.60 + 7 5.60 + 7 3.02 + 4 1.74 + 5 9.89 + 2 3.48 + 5 Rb-88 5.64 + 5 -
3.04 + 2 1.75 + 3 - -
Sr-89 7.42 + 5 -
4.00 + 2 2.30 + 3 - -
Sr-90 3.99 + 4 - 2.15 + 1 1.24 + 2 - -
St-91 9.72 + 5 -
5.25 + 2 3.01 + 3 - -
Sr-92 9.50 + 5 -
5.13 + 2 2.94 + 3 - -
Y-90 3.96 + 4 -
2.14 + 1 1.23 + 2 - -
Y-91 9.85 + 5 -
5.32 + 2 3.05 + 3 - -
Mo-99 1.28 + 6 -
6.91 + 2 3.97 + 3 - -
Ru-106 2.29 + 5 -
1.24 + 2 7.10 + 2 - -
Xe-131m 4.38 + 5 4.38 + 5 2.36 + 2 1.36 + 3 7.73 + 0 2.72 + 3 Xe-133m 3.07 + 6 3.07 + 6 L.66 + 3 9.51 + 3 5.42 + 1 1.91 + 4 Xe-133 1.27 + 8 1.27 + 8 6.35 + 4 3.93 + 5 2.24 + 3 7.87 + 5 2[1'Xe-135m 3.26 + 7 3.26 + 7 1.76 + 4 1.01 + 5 5.76 + 2 2.02 + 5 j Xe-135 2.09 + 7 2.09 + 7 1.13 + 4 6.48 + 4 3.69 + 2 1.29 + 5 Xe-138 1.17 + 8 1.17 + 8 6.31 + 4 3.63 + 5 2.07 + 3 7.26 + 5 u
Ch
Liquid II Gaseous I' Containment Containment ,
Source Sump Reactor Coolant Airborne Waste Gas Source Activity Activity Concentration Concentration Concentration Concentration Isotope (Ci) (CL) (uci/cc) (uCi/cc) (DCi/cc) (uci/cc)
I-131 3s 68 + 7 1.84 + 7 1.99 + 4 1.44 + 5 3.25 + 2 1.14 + 5 1-132 4.31 + 7 2.16 + 7 2.33 + 4 1.34 + 5 3.81 + 2 1.34 + 5 6.40 + 7 3.20 + 7 3.45 + 4 1.98 + 5 5.65 + 2 1.98 + 5 I-133 I-134 8.00 + 7 4.00 + 7 4.32 + 4 2.48 + 5 7.06 + 2 2.48 + 5 I-135 6.35 + 7 3.18 + 7 3.4 3 + 4 1.97 + 5 5.62 + 2 ,
1.98 + 5 Cs-134 1.27 + 4 - 6.85 + 0 3.93 + 1 - -
Cs-136 8.02 + 3 -
4.33 + 0 2.48 + 1 Cs-137 4.99 + 4 - 2.69 + 1 1.55 + 2 - -
Cs-138 1.23 + 6 -
6.64 + 2 3.81 + 3 Ba-137m 4.67 + 4 - 2.52 + 1 1.45 + 2 - -
Ba-140 1.25 + 6 - 6.75 + 2 3.87 + 3 La-140 1.27 + 6 - 6.85 + 2 3.93 + 3 - -
Ce-144 7.50 + 5 - 4.05 + 2 2.32 + 3 C r-5'1 - - 9.06 - 4 5.20 - 3 Hn-54 - - 1.01 - 4 5.80 - 4 Ha-56 - -
2.96 - 3 1.70 - 2 Fe-59 - - 1.01 - 4 5.80 - 4 Co-58 - - 5.23 - 3 3.00 - 2 Co-60 -- - 6.97 - 4 4.00 - 3 Zr-95 - - 8.71 - 5 5.00 - 4
.. (1) Based on 100% noble gas core inventory, 50% halogen core inventory, and 1% of all others core inventory.
N LJ7 '(2) Based on 100% noble gas core inventory and 25% halogen core inventory.
U1 CD tra N
4.0 REVIEW OF AREAS REQUIRING ACCESS FOD , vSTML IDENT OPERATIONS A review has been performed to ' ' atify areas requiring acci.ss for post-accident operat: ane. Dose ev..uations for the required access to these areas were performed, and the results are documented below.
An additional result of the review is that certain modifications are planned which will make it unnecessary to have post-accident access to certain areas. These modifications include:
a) Changing valves DHV-7, 8, 39, and 40 to motor operated valves, b) Adding a motor operated bypass valve for the Makeup and Purification System prefilter.
4.1 AREA A Area A is the designation for the nuclear sample room located on the 95' elevation of the auxiliary building (see Figure 4-1). For the evaluations associated with this area, refer to the responses for NRC Lessons Learned Task Force Short Term Recommendation 2.1.8.a (Improved Post-Accident Sampling Capability).
4.2 AREAS B AND C Areas B and C are the general location of the hydrogen purge equipment which could be used for controlling hydrogen buildup in containment. These areas are located on the 119' elevation of the Intermediate Building (see Figure 4-2). Access could be required 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> post-accident to line up and monitor hydrogen purge equipment.
phe estimated occupancy requirements to perform the necessary operations is 10 minutes. The activity inside containment is the predominant source of direct radiation for these areas. The dose rates at T=48 hours are estimated to be 350 rem /hr and 10 mrem /hr for Areas B and C, respectively. The 1755 038
resulting doses are 60 and 2 mrem, respectively. Access to Areas 5 and C is via Travel Route 1 shown on table 4-1. The dose for this travel route is negligible. The total doses for the required access to these areas is within the guidelines discussed in Section 1.0.
4.3 AREA D Area D is the general location of containment air monitor RM-A6. For the evaluations associated with this area refer to the response for NRC Lessons Learned Task Force Short Term Recommendation 2.1.8.a (Improved Post-Accident Sampling Capability).
4.4 AREA E Area E as identified on Figure 4-2 is the diesel generator room. Continuous access to this area following an accident would be required if of fsite power is lost. Access to Area E is via Travel Route 2 presented on Table 4-1. The dose rate in the diesel generator room from direct radiqtion following an accident is negligible. The dose for Travel Route 2 is also negligible. Area E could thus be continuously occupied if required.
4.5 RADIOACTIVE WASTE DISPOSAL CONTROL BOARD The radioactive waste disposal control board is presently located on the 95' elevation of the auxiliary building. Access to operate waste systems post-accident is required. This access will be provided by relocating the control board to a low radiation area.
4.6 CONTROL ROOM The control room is located on the 145' elevation of the control complex.
Continuous and immediate access is required post accident. Access to the control complex is via Travel Route 3 shown on Table 4-1. The dose rate in the control room from direct radiation from containment is estimated to be 11 mr/hr at *i=0. The dose for Travel Route 2 is negligible. These values would thus allow continuous occupancy of the control room.
1755 039
4.7 RADIOCHEMISTRY LABORATORY AND COUNT ROOM The radiochemistry laboratory and count room are located on the 95' elevation of the control comple. . Access is required for analysis of nuclear samples.
Access is via Travel Route 3 shown on Table 4-1. The dose rate in these areas f rom sources other than those associated with sampling is negligible. The dose for Travel Route 3 is also negligible. For the evaluation of the sources associated with sampling refer to the response for NRC Lessons Learned Task Force Short Term Recommendation 2.1.8.a (Improved Post-Accident Sampling Capability).
s 1755 040
TABLE 4-1 Travel Routes Travel Route Description 1 From the control room in the control complex at elevation 145 down to elevation 124 and north into the turbine building, down to elevation 119 and west in the turbine buildint to access the intermediate building at elevation 119.
2 Access from the southeast corner of the auxiliary building.
3 Access from office building to turbine building to control complex.
1755 041 O
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5.0 REVIEW OF RADIATION QUALIFICATION OF SAFETY EQUIPMENT Table 5-1 identifies the approximate location of the safety equipment considered in the review. The table also presents the radiation sensitive materials in each item and the integrated dose calculated for the six month period following the accident at the given location. Table 5-2 lists the radiation sensitive materials identified in Table 5-1 and radiation damage information for the stated reference materials of similar composition. It is not possible to reach conclusions as to the acceptability of the integrated doses calculated during the review of the radiation qualification of safety equipment. This is due to the lack of available data on:
- 1) qualification doses; 2) the period of time post accident that safety equipment must be functional.
1755 044
TABLE 5-1 Calculated I Integrated . Identification Location _ Dose (6 mon) r i 1 ulyment Name 3 Number Buildinj Elevation Coluno RADS Radiation Sensitive Material MUV 23/24 MCC Aux 95 304/J 5. 3 + 5 MUV 25/26 MCC Aus 95 304/L 5.3 + 5 Sample Boom Inters 95 304/g 5. 3 + 3 MUP Pump Filter (2) MU- FL 3A , B gthyg.ppgo 3, gg, gpony gepg,g Cellulose Fib. Filt. Seal Betura Cooler (2) MU-HE2A,5 Aum 119 302/N 7.528 + 4 Makeup Pump (3) MU-PI A, B,C Aun 95 303/J,E I.475 + 5 MU-V49F Aux 95 304/g 3.687 + 6 316+ Asbestos Casket, JC1873CR Letdoun Cooler Isolation Valve Packing Letdoun Cooler Shutoff valve MU-V50F Aus !!9 304/L 1.835 + 6 Evalpak 187-II Packlas + Fist Bypass Control Valve to MUD 111-IA MU-Vl94 Ai.s 119 305/D 1.144 + 6 Asbestos MU-F125 lalet Isolation Valve MU-V243 Aum 119 305/M 1.373
- 6 Asbestos + Pist MU-n2B Discharge Isolation Valve MU-V244 Aum 119 305/M 1.173 . 6 Asbestos + Pist Cation Denia Cross T$e Isolation Valve MU-V144 Aum 119 304/N 8.128 + 5 Asbestos + Fist Bypass leolation Valve to MU-F MU-V126 Aum 119 304/M 8.728 + 5 Asbestos + Past MU-Vil6F Aum 119 304/L 1. 8 35 + 6 JC-187-I-CR Asbestos Packing, Isolation Valve to MU-F S/S 316 + Ash Cash, Rubber Pist Casam Demineraliser to Steed Holdup MU-Vil2P .A us 119 304/M 8.728 + 3 Ethylene Propylene Asbestos, TFE, SAE P-3 Felt h W9)F Aus 119 303/L 1.144 + 6 JC-187-1-CR Asbestos Pech ing, Purification MU Inlet to MU-F2B S/S 316 + Asb Cash, Rubb.+ Fist.
Gramm MU Filter Bypass to MU-T1 MU-VIDO Aus 119 303/L 1.144 + 6 JC-187-3-CR Asbestos Packing. O S/S 316 + Asb. Cash, Rubb.*Pist. MU-V357 Aum 119 302/L 7.528 + 5 JC-187-1-CR Asbestos Pec king, Seal Return Isolation Valve S/S 316 + Asb. Cash, Rubb.+ Pist MU-Tl Isolation Valve MU-V64 Aus 95 302/L 5.299 + 5 Bralded Asbestos Packing + Oper. MCP Section Isolation Valve (4) MU-V68,69,62,6 3 Aus 95 304/J.L 3.687 + 6 Asbestos Packing MUP Discharge Stop Check Valve (3) MU-V2,6,Il Aua 95 30' / L. E.J 3.687 + 6 Asbestos Packing MUP Discharge Isolation Valve (4) MU-V3,4,8,9 Aus 95 304/L,R J 3.687 + 6 Braided Asbestos Packing JG-187 MU-V23F 24F.25F.26F Aun 95 304/K.L 3.68 7 + 6 316 + Asbestos Casket , JC1871CR HP1 Control Valve (4) Pecking RC Pissp Seal Control valve MU-V16F Aum 95 304/K 3.687 + 6 TFE Packing, Witrile Diaph. MUP Bypass Control valve MU-V17 Aun 95 304/K 3.687 + 6 Bralded Asbestos Packing JC-187 O 5S316 + Asbestos Cash RC Pump Seal Isolation valve MU-VIBF Aus !!9 305/N 1. 373 + 6 316 + Asbestes Casket. JClb7CB Packing Seal lajection Bypass Valve N V452 Aus 95 305/N 1.144 + 5 Braided Asbes tos Pecking JC-187, N 5S316 + Asbestos R3 Spray Pump (2) $4-PIA,B Aun 75 305/N.0 1. 3 71 + 6 Carlock # 7022 Casket. A-108 + Rubber Seal b- . 305/P,0 1.173 + 6 JC-38 7-I-CR Asbest is Pack, S/5316 Ra Spray Pump Isolation valve (2) BS-v16F.17F Aun 75
+A$$ Cash, Rot . Gr amm g RB Spray header Isolation Valve (2) BS-VJ,4F Aux 95 305/N 1.144 + 5 Asbestos Pachang Jg 187-1-CR 5
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TABLE 5-1 (Cont.) Calculated ; lategrated tocatton Dose (6 mon) Identification Column RADS Dediation Sensitive Material Building Elevation Number hulpment Name 1. 37 3 + 6 JC-187-1-CR Asbestos Fack, Aux 75 305/F S/S 316 + Asb Cask, Rubber R$-V 36F.3?F SS-T2 1 solation Valve (2) Croem Bed Rubber or Neopsene Caskets Aus 95 306/T (1) AH-HE30A,5 1/16" Tack Auz Bldg. Air Handitsg Cooling Coil (2) 1. 37 3 + 6 Red Rubber on Neoprene Caskets Aux 75 30's/N ,0 BG-FIA,3 1/16" Tack , 13 Spray Pump Cooling coil (2) (1) 1/8" Tack Neoprese caskets Aus 95 304/T l DH Closed Cycle Heat Eachanger (2) DC-HEIA B Aus 95 306/S,7 (1) DC-FI A.B 75 304,306/F 1.373 + 6 DH CC Cooling Pump (2) Aus DH-MEIA.R Aus 75 305/Q l.373 + 6 DH Heat Enchanger (2) DH-FIA,B 3.475 + 5 DH Pump (2) Aus 95 303/J.K Carlock 7021 Caskets, Carbon push-HU-FIA.C 95 307/P.R 5.299 + 4 MUP Cooling Coll Fump (2) RW-F 3A,8 Aun ings (Cutless Rubber) 0-Ring DH Semuater Cooling Coil Pump (2) BunaN. Crane superseat il (1) TFE Packing, patrile Diaphree Aum 95 306/S DCV-10,12 Dante Water to DHCC Surge (2) 306/S,7 (1) Buna M (Hycar) Seat and Seals Aum 95 JC-187-8 Asbestos Facking DHCC Tank Isolation Valve DCV-1,2,3,4 306/7 (1) Aun 95 Fune isolation valve (4) DC-V31,32 DH Air Unit Inlet isolation valve (2) 95 306/T (1) Asbestos DH Air Unit Discharge Isolation Valve Aus Buna u (Hycar) Seal sad Seals DC -V37,38 Aus 95 305/T (1) Ashestos Packing (2) DC-V13,14,15,16 95 304/J.E 1.373 1 DH CC Beat Exchanger (4) Aus Bralded Asbestos Facking - JC-187 DC-V21,22 Aus 95 307/F 5.299 + = JC-187-1 Asbeston Focking MUF Punip Motor (2) DC-v43,44 305/N,0 1.373 + 6 Aun 75 Breided Asbestos Fecking JC-187, DH Shutoff Valve (2) DC-V27,28 75 305/N,0 1373 + 6 kB Spray Pump Inlet Isolation Valve (2) DC-Vll5,Il7 Aux S/S 316 + Asbestos EB Spray Fump Stuffing Sun Bearing (2) 1.373 + 6 Aatestos
- Aun 75 305/N JC-187-1 Asbestos Facking DC-V31 75 305/q 1.37 3 + 6 RB Spray Pump Discharge Isolation valve DC-V29,30 Aun DCH Pump Inlet Isolation Valee (2) 1.373 + 6 Buna N (Hycar) seat and Seals DCM Heat Enchanger inlet Isolation Valve DC Aun 75 304/R Carbon Filled TFE Bushings, TFE d DC-VS,6 75 304,306/R 1.373 + 6 Facking, Nitrile Diaph.
(2) DC-v17,lt Aus S DCH Cooler Control valve (2) Buna N (pycar) Seat and Seels DCJ Jest I.acjanger Discharge Isolatiot 75 304,305/3 1.373 + 6 Carbon Filled TFE Bushings, IFE Aux DC-V7,8 75 304,305/R 1.373 + 6 Packing, Nitrile Diaph. Valve (2) DC-V177,178 Aus d DCH Cooler Control Valve (2) a 305,306/N 1.144 + 5 316 + Asbestos Casket. JC1871CR lou Pressure Injection Isolation valve Aum 95 racking N (2) DH-V5F 6F @ 2.709 + 5 JC-1871 Asbestos Packing 95 30 5/0 JC-187-1-CR Asbestos Pack, S/S 316 Aum DH-V 7,8 95 306/F 5.299 + 4 + Asb Cask, Rubber Crown Motor DH Discharge Crosstle (2) Aus DH-V12 DH Fump Discharge g
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TABLE 5-1 (Cont.) Calculated lategrated , _ Location _ Duse (6 mon) Identificattoo _ RADS Rad _tation Sensittwe Material Number Building Elevation Co g hulgeent Name Asbestos Facking Aus 75 306/F 1.373 + 6 DH-V144 5.299 + 4 Ersided Asbestos Packing JC-187 Du Beat Enchanger Aum 95 306/Q Du Fump Discharge Isolatjoe Valve DM-V106 5/5 316 + Asb. Cas + Motor 305/F 1.373 + 6 JC-187-1-CA Asbeston Fath. l Aun 75 - DH Fmp Section to DH-T1 1 solation (2) DM-V34F.35F S/S 316 + Aab Cas. Rubber Cromm I Du Fump Isolation valve (2) DH-V 39F.40F Aux 95 305/O 305/m 2.709 + 5 1.144 + 5
+ Ntos Asbestos Faching JC 187-1-CR Asbestos Fack. 5/5 316
'f DH-V41F Aun 95 + Asb Cask. Rubber Gramm + Ntor f Du Costalement Isolation 1.373 + 6 316 + Asb Cash. Asb Packlag Aun 75 305/M.N Once opened must DH-V42F,43F JC-187-1-CR. ,
Ra sump to DH Fump Isolation (2) remala operator recirt. 304/F 1.373 + 6 JC-187-1-CR Asbestos Pack. 5/5 316 DH-V110,111 Aus. 75
+ Ash Cask. Rubber Cream + Motor DH Exchanger Discharge laolation (2) Braided Aabeston Facking JC-187 Aum 119 305/N 1.373 + 6 DH Fressuriser Spray isolation DH-V91 5/5 316 + Amb. Cask + Motor !
3.373 + 6 Bratded Asbestos Facking JC-187 Aus 119 305/N WD-9405 5/5 316 + Asb. Cask + Motor RS Vent BeeJer teolation valve Vinyl Coated Mylar label Grade M Aum 95 303/F 3.703 + 6 WD-V 39 3.394. 395 EFT Nordel Diaph.. Neopreme cask ' Waste ces Decay Tank to secycle valves Elastomer Diaph. Dash #114 (3) EPROttag 303/F 3.703 + 6 Vinyt Coated Myler Label Grade M Aus 95 EFT Nordel Diaph. . Neops eme Gash WG DT Outlet valve to vent System (3) WV436.437.438 Elastomer Diaph. #112 0-Ring i 3.70) + 6 Crede M EFT hordet Diaph. Aum 95 301/F O-Ring - EPR Dash f1124222 WD-V 382 WG Surge Tank Bypase Isolactoa Flestic Shtm W..her 3.703 + 6 Crade M 1.F7 pordel Diaph. Aux 95 303/F WG Serge Tank Inlet Isolstica WD-V 380 O-R!as - LFR Dash #112&222. Plastic Shtm Washer 3.703 + 6 Crede M EFT Nordel Diaph. Aux 95 303/F O-Fing - EPR Dash #1126222, 4 WD-V341 WG Surge Tank Outlet isolation F astic Shia Washer 3.703 + 6 TFE Packlag Aus 95 303/F Crade M EFT hordel Diaph.. C WG Compressor Control Valve WFV465F WD-V384,385 Aus 95 303/F 3. 70 3 + 6 Neoprene Cask. Vinyl Coated WGC Suction Isolation valve (2) Mylar Label. Elastomer Diaph. O king Dash $112 EFR Tard 119 307/F (1) N 7.37-LT 119 302/L 3.7 + 7 Borated Water Tank Level Trans MU-le-LT Ava s Makeup Tank Level Trans Aus 3.7 + 6 RM-L-01 305/u 1.4 + 6 ,, BC Letdoun Nattor RM-A-06 Aus 119 RS Air Sample Tard 119 307/F (1)
- f SS-5-LT 95 302/F 3.7 + 6 NaOS Task Levol Trans. WD-36.17.18-LT Aum g cas Decay Tank Level Trans.
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k1 ~ *** - - " - TABLE 5-1 (Cont.) Calculated Integra.ed IdentificatJoe Location Dose (6 mon) Number Building Elevatton ' Column RADS Radiation Sensitive Material l Egup ment Name SW- 7 8,80,81,82 Aux 119 305,306/M 1.144 + 6 & Seale (Sulf ur RS Fump Inlet Isolation (4) LPD,M r.e,Nordel Seat . A ux !!9 305/N 1. 3 7 3 + 6 Br aided Asbestos Faching + Fist. I MS ? solation to Rs SW-IID SW-12 Aus 119 304/F 1.732 + 5 LPDM No. 4el Seat & Seals (Sulfur i EvapoJ'.ar Fackage Isolation Free) Fist. Oper. -f SW-57 Aux 95 302/L !.715 + 3 1. PDM Nors 1 Seat & Seals (Sulf ur Misc. Evap. Outlet Isolation Free) + 11st. Oper. . Aus 95 302/L 1.715 + 3 EFDM Nordel Seat & Seals (Sulfur I p.: Evap. Outlet Isolation SW-58 Free) + Fist. Oper. ( Nuclear Service Heat Enchanger (4) SWhiE-1A,B,C.D Aum 95 305/5 (1) RWV-17,18 Aum 95 306/R.F 5.299 + 4 Buna N (Mycar) Seat & Seals Du Seavster Pump Discbarge (2) 5.299 + 4 Buoa N (uycar) Seat & Seals RWV-21 Aum 95 306/Q Energency NS SWF Discharge Isolation Aux 95 304/Q,P 5.299 + 4 Buna N (uycar) Seat & Seals Normal MS SWF Discharge Iso. (2) kWV-22,24 RWV-5,6,7,8 Aus 95 305/3.5 (1) Buna N (Nycar) Seat & Seals MS Heat Exchanger Inlet Isolation (4) Bona N (Rycar) Seat & Seals NS Heat Each Discharge Isolation (4) RWV-13,14,15,16 Aux 95 305/5 (1) RWY-32,33 Aum 95 307/S R 5.299 + 4 Buna N (Mycar) Seat & Seals MS Meet Each Isolation (2) RWV-40,41 Aus 95 s 307/5.R 5.299 + 4 Bana N (Bycar) Seat & Seats Du CC Beat Each Isolation (2) Aum 95 308,309/G (1) Asbestos Faching + Motor Botwell lealetion to EFF (2) EFV-1,2 . Aus 95 309,308/C (1) JM 3s8 Facking + Mr. tor
, EFF to Steam Cea Isolation (2) EFV-7,8 Aus 95 308/C (1) Asbestos Packing + Motor
. Condensate Storage Tank Isolation - .. EFV-4 ' -e*
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a t . h id. p,dn - - Aus 95 304/N 5.659 + 4 , j Radunate Disposal Control Board Auv-1A Aum 143 303/N (1) '
- * , RA Furge Eahaust valve 1st 119 309/J2 3.359 + 2 RB Purge supply Valve AHV-1D Ana . 95_ . 306/7 5.299 + 4 Neopreme 8. Fiberglass laeul.
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119 yJ41 304/J (1) . , %& ,, L , . . ,,; , , f . g* [gBand11ag, Unit'(2)d'43 OL , CC 164 301,304/a (1) Control Comples Emergency Supply Fan (2) AHF-18A,5 303/C (1) ANF-19A,5 CC 164 Control Comptes Retura Faa (2) 302/9 (1) Class-Teflon lasnl. AIDIE-4&,5 CC 764 Control Comples Heattag coil (2) Red Rubber or peopreme Cesket d Control Comples Coo 11eg coil (2) AHHE-5A,3 CC 2 6e 301/C (1) 1/16 Thck [ Emergency Diesel Can Supply Fan (4) AMF-22A,3 C D AMFL-5A,3 DC DC 119 119 301/Q 301/Q (1) (1) Neoprene Claes Mat Media, peopreme Base [ Energ DG Filter (2) asu 95 301/E 5.299 + 5 Adhesive i ES Motor Control Center 3Al (5) MTMC-3 M1MC-4 Aus 119 301/L 1. * *.2 + 2 ES MCC 3A2 Aux 119 301/M 1.622 + 4
"""' ES MCC 351 MTMC-5 Aus 95 303/N 5.659 + 4
== J [ g ES MCC 382 ES MCC 3As MTMC-6 HTMC-7 Aum 119 304/0 1.732 + 5 (2) Silfcom Bubber. Tefles, Aabestos.
Feuer & Control Cable Bypelon, Eerste (Strad),
-- -- -- --- (2)
Instruneet & Control Cable -- --- --- (2) Cnosellok FolyethyPane
'(f) Control & Therecouple Cable ---
1.732 + 2 f9# l' " " ' Wall Mounted contactors - MIMC-4
-- Aus 119 301/L 301/M 1.732 + 2
--- Aux 119
# C Wall Mounted Contactor - MTMC-5 Wall Mounted Contactor - MTMC-7 -- Aus 119 304/0 1.7 32 + 5 A 8 reds.
FOOTNOTE 2: Dose not calculated. Cable has been qualified to 10
@ f FOOTNOTE la Negligible <100 R E
i TABLE 5-2 . Radiation Sensitive Exposure (Rads) Component Material Reference Material Ethyl - Ppro seals EPM ethylene propylene monomer 5.0 x 106 (3) Epoxy impregnated cellulose fiber filter Epoxy cellulose pulp filled 108 (g) Asbestos Asbestos fabric filled phenolic 1010 (g) Evalpak 187-1X packing --- JC-187-I-CR Asbestos packing Asbestos fabric filled phenolic 10 '(1) . Rubber gromm Natural rubber 4.5 x 10 (2) Ethylene propylene EPH ethylene propylene monomer 5.0 x 106 (3) TFE Teflon < 1.0 x 106 Cg) SAE P-3 Felt Felt 5.4 x 106 (4) Braided asbestos packing JG-187 Asbestos fabricfilled phenolic 10 (1) 0 Braided asbestos packing JC-187 Asbestos fabric filled phenolic 10 Cg) Carlock #7022 gasket A-108 --- Rubber seal Natural rubber 4.5 x 107 (2) Neoprene gaskets Neoprene 5.0 x 106 (3) Garlock 7021 gaskets --- Crane super seal #1 Buna N 0-ring IlYCAR OR-15 NBR, llYCAR PA-21 1.5 x 107 (2), 1.6 x 107 (2) ( W Carbon bushings Buna N (ilYCAR) Seat & Seals Carbon HYCAR OR-15 NBR, HYCAR PA-21 negligible damage 1.5 x 107 (2), 1.6 x 107 (2) W Vinyl coated mylar label Mylar A 100 (l') Grade H EPT Nordel diaphragm EPD'l e.thylene peopylene diene monomer 5.0 x 106 (3) o
TABLE 5-2 (Cont.) Radiation Sensitive Reference Material Exposure (Rads) Cpmponent Material 2.1 x 107 (2), 1.2 x 107 (2) Elastomer diaphragm Butyl rubber GR-I 50, Styrene - butadiene rubber (SBR) (GRS-50) Neoprene 5.0 x 106 (3) Neoprene JM-398 packing Viton o-ring Natural rubber 4.5 x 107 (2) Red rubber Nitrile rubber HYCAR OR-15 IIBR 1.5 x 107 (2) Nitrile diaphragm EPM ethylene propylene monomer 5.0 x 106 (3) EPR o-ring Plastic shim washer Armco-I JAF gasket Plastic insulation ASB polyacrylate silicon Sponge 2.1 x 107 (2), 1.2 x 107 (2) Butyl rubber GR-1 50, Styrene-Synthetic rubber butadiene rubber (SBR)(GRS-50) INVAR Carbon negligible damage Carbon bushings EPDM 5.0 x 100 (3) EPDM Nordel (sulphur free) Teflon < 1.0 x 106 (1) Teflan seat and seal 5.0 x 109 (4)
'%J Fiberglass (y, Fiberglass insulation < 1.0 x 106 gg) ll' Class - teflon insulation Teflon Glass 5.0 x 109 (4) c:3 Class mat media t_n
TABLE 5-2 (Cont.) . Radiation Sensitive Exposure (Rads) Component Material Reference Material Neoprene 5.0 x 106 (3) Neoprenebase adhesive Silicone rubber Silastic 7-170 6.8 x 106 (2) Silicon rubber Teflon < 1.0 x 100 (1) Teflon 6 HYPALON 5.0 x,10 C3) Hypalon Kerite (Brand) Black KLPE crosslinked 5.0 x 106 (3) Crosslink polyethylene polyethylene (1) Exposure which the given material received and upon subsequent testing retained 80% or more of its pre-irradiation mechanical properties of alongation and tensile strength. Engineering Compendium on Radiation Shielding, Vol. II, Springer-Verlag, Berlin /Heidelberg, pp. 311-314, (1975). Engineering (2) The 25% damage dose for the mechanical properties of tensile strength or elongation. Compendium on Radiation Shielding, Vol. II, Springer-Verlag, Berlin /Heidelberg, p. 316 (1975). (3) Exposure which the material received and upon subsequent testing retained 75% or more of its pre-irradiation cachanical properties of tensile strength and elongation. " Insulations and Jackets for Control and Power Cable in Thermal Reactor Nuclear Generating Stations," R. B. Blodgett and R. G. Fisher, Okonite Co., Passaic, NJ, 07055. Institute of Electric and Electronic Engineers - June (1969). Remote Har.dling of Mobile Nucl. ear Systems, (4) Dose at which a physical or mechanical property changes by 25%. D. C. Layman and G. Thornton; U.S. Atomic Energy Commission /Div. of Tech. Infoimation, Oak Ridge, Tennessee, pp. 506-515, Jan. (1966).
%d O'1 LJ7 CD*
t 71 N
REFERENCES
- 1. C. M. Lederer, J. M. Hollander, and J. Perlman, " Table of 1sotopes,"
Sixth Edition, John Wiley and Sons, Inc., 1967.
- 2. A. Tobias, " Data for t'.e Calculation of Gamma Radiation Spectra and Beta Heating from Fission Products (Revision 3), RD/B/M2669, June 1973.
- 3. E. D. Arnold and B. F. Maskewitz, "SDC - A Shield Design Calculation Code for Fuel Handling Facilities," ORNL -3041, March 1966.
- 4. H. Ono and A. Tsuruo, "An Approximate Calculation Method of Flux for Spherical and Cylindrical Source with a Slab Shield," Journal of Nuclear Science and Technology, 2 (6), pp. 229-235, June 1965.
- 5. K. Shure and O. J. Wallace, " Compact Tables of Functions for Use in Shielding Calculations," Nuclear Science and Engineering, 56, pp. 84-94, January 1975.
s 1755 053
ATTACHENT IV VENTING DESIGN CRITERIA A. Summary Description Various postulated small breaks can lead to eccident scenarios in which steam and/or non-condensible gases accumulate in the reactor vessel head, the upper portion of the hot legs and in the pressurizer. Following repressurization of the RCS by HPI, which will tend to col-lapse steam bubbles, remotely controlled vents on the upper hot legs and pressurizer can be used to vent non-condensible gases to aid in refilling the RCS and promoting natural circulation flow for core cooling. B. Piping and Y31 ving Considerations
- 1. Wnts shall be provided at the following reactor coolant system high points:
- a. Top of each hot leg (2 vents total, one per hot leg)
- b. Top of pressurizer (1 vent)
- 2. Vent piping and valving shall be designed and sized s;ch that the failure to close off any one (1) cf the vent points listed above could not cause a loss of coolant at a rate in excess of the nor-mal makeup system capability at full design RCS pressure.
- 3. The effluent flow from all vent points shall be routed directly to the containment atmosphere. The region into which the dis-charge is diverted shall enhance mixing and dilution so as to minimize the potential for local regions from reaching flammable concentrations of gases. Discharge needs to be routed and directed so that liquid effluent can not discharge on or fall on electrical equipment or mechanical operating equipment.
- 4. The piping and valving for the venting system shall be routed, oriented and protected so that damage from pipe whip, jet impingement and missiles will not occur.
- 5. Pipe routing, orientation and elevation shall assure that all remotely operable valves are (a) located well above the maximum level of water in the containment expected for the worst case DBA and (b) protected from the containment spry and relief dis-charges. Each vent shall be designed to remain functional after all design basis events except large LOCA's, evacuation of the Main Control Room and loss of all AC power.
- 6. Vent piping and valving shall be designed for 2500 psig and 670'F and any gaskets or seals shall be compatible with all anticipated effluent fluids. This includes water, saturated steam, steam water mixture, superheated steam, fission product gases, helium, nitrogen and hydrogen. Provisions for venting hydrogen shall in-clude the necessity for spark-free valves.
1755 054
B. Piping and Valving Considerations (Cont'd)
- 7. Each venting point shall be individually operable independent of any other point vent. Operator guidelines shall be provided to reduce the possibility of venting from more than one vent point at a time and to minimize the possibility of excessive RC system depressurization. Ref. Section B2.
- 8. All piping and valving shall be connected to the RCS and support-ed in such a manner so that any stress due to weight, thermal transients, internal piping conditions and external environment
- will be within the maximum allowable stresses at the existing vent nozzles. Piping shall be designed to prevent the fonnation of traps and minimize the possibility of water and/or steam ham-mer.
- 9. Existing nozzles in the RC System shall be used for the venting system. No new nozzles shall be added exclusively for venting.
- 10. All remote operable "two position" (on-off) valves shall be of the fail closed type, with power required to open and power re-quired to maintain open. Valves shall have proven fail " closed" action on loss of power. No packings are pennissible.
C. Control and Instrumentation Considerations
- 1. All valves for any one vent nozzle shall be powered from a supply separate from that which powers the valves for any other vent nozzle so that any single power supply failure cannot cause a failure to vent at more than one vent nozzle. Vital power sup-plies shall be provided for all of the vent isolation valves.
Complete vent shutoff of any one vent nozzle shall be assured on loss of all power to its venting system.
- 2. Control of vent valves shall be remote manual and operable from the control room only. There is no requirement for operation fran any auxiliary location. Direct indication of valve posi-tions shall be provided in the control room.
- 3. Control of valves for any one vent point shall be independent of the control for valves for any other vent point.
- 4. Both unt valves at a vent point shall be powered by the same power source, but controlled by two (2) independent switches. An alann shall indicate both valves are energized.
- 5. The vent valve operating switches shall be such that the vent valve will not open when power is applied to the switch. It must take an independent action to operate the switch.
D. Operating Guidelines and Modes
- 1. The venting system may be used to vent the RCS during RCS filling operations if no venting system functional requirements are violated.
F755 055
D. Operating Guidelines and Modes (Cont'd)
- 2. Priorities of system design shall be as follows:
- a. RC system integrity
- b. Capability to vent to containment atmosphere
- 3. Special precautions shall be taken to prevent any unauthorized venti ng. Such precautions shall be in addition to nomal admini-strative controls.
E. Testability
- 1. Provisions shall be made for testing all portions of the venting system at any time during startup of the plant. Testing shall consist of the following:
- a. Confinnation of free flow passage from each vent point.
This will include exercising of all valves and checking the flow. Flow indication need only indicate that flow is present, and quantification is not required. Such testing shall nomally be done durirg initial fill.
- b. Confimation of vent shutoff capability shall be established by initially filling the vent lines curing pre-service hydro and establishing effluent flow into the containment with the vent valves open. The vent valves in e/4ch vent system will then be closed until effluent flow is observed to cease. If it does not cease after a few minutes and continues at some observatle rate, this should be quantified by timing flow into a vessel of known volume. A leakage rate in excess of 10cc per hour fran a vent system at pre-service hydro pras-sures shall be considered unacceptable. No external leakage is permissible.
- c. Flow indications which have been previously tested shall be monitored to assure that gross inadvertent venting is not occurring during normal itactor opera:. ion. Ref. Sec-tion E.1.a. above.
F. Th_ermal Stress and Insulation Considerations
- 1. When practical, effluent piping from the high point vent nozzles to the vent valves shall be themally insulated to:
- a. Provida piping protection
- b. Reduce heat losses
- c. Minimize piping stresses 1755 056
- d. Provide personnel protection F. Thermal Stress and Insulation Considerations (Cont'd)
- 2. Provisions shall be made to minimize thennal stresses due to venting, so that intennittent venting can be tolerated.
G. Environmental Qualifications
- 1. The high point vent system shall be designed to maintain its in-tegrity and ftstetion for the lifetime of the plant, assuming periodic replacement of consummables.
- 2. All canponents located inside primary containment shall be quali-fied to the maximum LOCA or main steam line break (MSLB) environ-mental conditions and to the process conditions stated in Section B.6.
H. Safety Classification
- 1. All fluid portions of the venting systems from the vent nozzles, up to and including all vent valves, are part of the reactor coolant pressure boundary and as such are seismically qualified and classified per Reg. Guide 1.26 as safety class A or B, depending upon piping size.
- 2. The valves shall be classed as active, subject to the require-ments of Reg. Guide 1.48.
- 3. As a minimum, all electrical portions of the system for hot leg "two position" (on-off) valves shall be Class IE.
1755 057 S 4 VENTING SYSTEM SCHEMATIC 1755 058
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e m ee ** VENT FLOWRATE CURVES
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1755 062
co-11UDDdU A study was performed to determine the flow rate of water, saturated steam, super heat steam and non-condensible gases through High Point Vent Valves (HPV). The venting system envisioned is shown below: L A.g K^b i
* ~^,
r-p, 5 3 .. f P) = System Pressure at Vent , P2 = Pressure at End of Vent Pipe P3 = Receiver Pressure . K) = Flow Resistance k-factor of Vent Pipe K2 = Flow Resistance k-factor of All Down Stream Piping, Valves Turns, etc. The venting rates will be a function of jP , 2P , P 3 , y , y and type of fluid being unted. Since venting rates are a function'of so many variables, the following restrictions were put on the calculations: , (1) Minimum vent area is at the system exit (A) on figure). . (2) P,A,g,andP3 j 2 are such that P is always less than critical pressure 2 for sonic flow rates. .. (3) Non condensible gas temperatures equal temperature of saturated steam. Minimum flow rates (with sonic velocities) of saturated steam and non condensible gas as a function of Kj and P) are shown on Figures 1 and 2. Superheated steam will have approximately the same Cp/Cv ratio as saturated steam and the flow rates will therefore be equal to the flow rates on Figure 1 times a correction factor.. This correction factor is equal to the square root of the density ratio of superheated to saturated i steam. Figure 3 shows.ventina rates of water as a function of the whole systern k.-factor and pressure per one square inch of vent area. .Two phase flow will be assumed
. ~
1755 063
OO 11UJJdU ' as the linear ratio of saturated steam and water weighed Dy the percent quality (i.e., zero percent quality would be all water, 50% quality would be half water
.:alf steam and 100% quality would be all steam).
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. _ . -q.755 066
Figure 3 FLOW MTE OF SATURATED LIQUID VS p K-FACTOR AND SYSTEM PRESSURE
- a PER ONE SQUARE INCH OF VENT AREA o P G
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ANCHOR MOTIONS, SEISMIC RESPONSE SPECTRA AND ALLOWABLE N0ZZLE LOADS I 1755 068
.36-1106601-00 INTRODUCTION Contained herein are structural design data to be used in the design and analysis of the proposed high point vent valves. It is being proposed that
~
"' remotely operated valves be made available:
- - one at the top of the hot leg U bend, and another atop the pressurizer. The data presented in this doctanent are worst case data anu should be used for the design and analysis of all proposed valves.
The data consist of anchor motians, seismic response spectra, and allowable nozzle k ads. Anchor motions are presented for various loading conditions including deadweight, thennal expansion, steady state hydraulics, operating basis earthquake (OBE), and design basis earthquake (DBE). The latter tenni-nology, DBE, is synonymous to SSE or Safe Shutdown Earthquake. Steady state hydraulic loads are those loads induced by pressure expansian and fluid motion within the system and should be imposed on all components except the pressurizer for all loading conditions requiring the inclusion of operating loads. Thennal and deadweight anchor motions are self-explanatory; however, the thennat motions provMed are the maximtsn experienced during operation at 8%,15% and 100% power. A seismic response spectra curve is provided for excitation in each of the X, Y, and I directions shown in Figures 1 and 2 for both the OBE and SSE. Since a 1" diameter pipe is be:ing considered, the values of critical damping reconnended by Regulatory Guide 1.61 are adopted. Excitation at 1% critical damping is provided for the OBE and 2% for the SSE. Allowable nozzle loads have been provide for a 1" schedule 160 pipe constructed of Ni-Cr-Fe 5B167 material. Forces obtained from structural analyses should be compared by the method indicated on page 10 to the allowable loads listed there for the various load conditions. It is important to note that the forces and allowable loads are defined in the coordinate system shown in the accompanying sketch at the bottom of page 10. 1755 069
86-1106601-00 ANCHOR MOTIONS 9The coordinata system is defined in Figures 1 and 2. e Easultants are obtained by the " square root of the sum of the squares" method. eDisplacements are in inches. 6Estations are in radians. 0 0 0 4 4 4 v z y : x ME z _ _
.001 .00001 .00028 .00375 Z-Earthquaka .122 .001
.008 .0001; .00001 .00001 T-Earthquatca .002 .008
.003 .026 .148 .00359 .00003 .00001 Z-Earthquake
.027 .155 .00359 .00028 .00375 Easultant .122 DEE(SSE)_
.001 .003 .00002 .00053 .00674 Z-Earthquake .218
.014 .015 .00023 .00002 .00002 T-Earthquake .004
.048 .273 .00654 .00006 .00002 Z-Earthquaks .006
.050 .273 , .00634 .00053 .00674 Easultant .218
.012 .047 .020 .00040 .00000 .00005 DEADurICHT
.049 3.643 .534 .00139 .00002 .00005 THEILv.AI. EDANSION
.000 .096 .024 .00023 .00000 .00000 STEADT-STATE ETDRAUI.ICS 1756 070
. z.
C:-1106601-00 RISPONSE SPECTRA Easponse spectra are provided on the following pages for the conditions listed below: 03E - Operating Basis Earthquake I - Direction T - Direction Z - Di ection SSE - Safe Shutdown Earthquake I - Direction T - Direction 2 - Direction 9 1755 071
.s.
86-1106601-00
*ESPONSE SPECTRA I-DIRECTION EARTHQCAKE - OBE 12 C2ITIcAL DAe. Ixc 100.0- L i -
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TABLE OF CONTENTS Page Section 1
1.0 INTRODUCTION
2 2.0 SHIFT OPERATIONS . . . . . . . . . ......... . SHIFT COMPLD(ENT , . . . . . . . . . . . . . . . . . 2 2.1 3 2.2 ADDITIONAL SHIFT PERSONNEL . . .......... 2.3 AUTHORITIES AND RESPONSIBILITIES FOR SAFE 3a OPERATION AND SHUTDOWN . . ... ..........
.............. . 4 2.4 SHIFT WNCTIONS . . .
Sa 2.3 CONTROL CENTER RECORDS . . ... .......... Sa 2.6 ON-CALL ............ ... ...... . 6 2.7 PROCEDURES . .................... Adherence to Procedures ... . ......... 6a 2.7.1 7 2.8 OPERATOR RETRAINING . ..... ... ....... 7 2.9 PLANT MODIFICATION AND SET POINT REVISION NOTIFICATION 2.10 PLANT OPERATING QUALITY ASSURANCE MANUAL 7 REVISION NOTIFICATION ........ ...... . 8 2.11 SUPPLEMENTAL LABORATORY ANALYSIS REPORT ......
.. 8 2.12 GENERAL PRACTICES FOR COLLECTING SHIFT RECORDS .
SHIFT LOGS . .................... 10 2.13
... ...... . 12 2.14 REACTOR TRIP AND PLANT SHUTDOWN
.... .......... 13 2.15 SHORT-TERM INSTRUCTIONS
.... ........... 13a 2.16 CONTROL CENTER ACCESS SHIFT RELIEF . ................... 14 2.17
.... 14a l 2.18 INSTRUMENT READINGS AND CONTROL INDICATIONS
......... ......... .. 14a 2.19 KEY CONTROL CONTROL CENTER STATUS BOARD ............ 15a l 2.20
...... ..... .. 16 2.21 CONTROL CENTER REFERENCES NOTIFICATION . . .. . ......... ...... 17 2.22
............ 18 2.23 CONTROL CENTER HOUSEKEEPING 18 2.24 EQUIPMENT OUT-OF-SERVICE LOC . .... ...... .
2.25 UNUSUAL OPERATING EVENTS REPORT ... ....... 18a
... ........... 19 3.0 NON-CONFORMING OPERATIONS 4.0 WORK REQUESTS ..-r................. 20 RADIATION WORK PERMITS . . . . . . . . . 21 5.0 .......
6.0 EQUIPMENT OUTAGES . .. ...... ......... 21a 1755 083 t Page (i) AI-500 Date 10/25/79 Rev. 28
TABLE OF CONTENTS Section (Cont'd) g ENCLOSURES Enclosure 1-Forn 912235, Change Notification . . . . 23 Enclosure 2-Form 912225, Supplemental Laboratory Analysis Raquest Form ......... 24 Enclosure 3-Form 912212, Shutdown Report ...... 25 Enclosure 4-Form 912220, Kay Control Log . . . . . . 26 Enclosure 5, Short-Tenn Instruction ........ 27 Enclosure 6, Short-Term Instructions Index . .... 28 Enclosure 7. Equipment Out-of-Service log ..... 29 Enclosure 8, Unusual Operating Event Summary ... 30 32 Enclosure 9. Shift Relief Checklist ........ 34 Enclosure 10, Operational Status Checklist ..... Enclosure 11. Critical Plant Equipment /Paramer9ts Checklist ........ ...... 35 1755 084 t Fage (ii) AI-500 Date 10/25/79 Rev. 28
9
- 1. 0 INTRODUCTION Plant operations are conducted by the Operations Section with the coope-ration and assistance of personnel from other plant sections, corporate headquarters, and non-Company organizations as required. The instructions containad in this section do not supersede or nullify applicable legal re-quirements or regulations. It is the purpose of these instructions to insure that plant operations are conducted in conformance with applicable legal requirements anJ regulations, and dictates of good operating practice.
1755 085 i AI-500 Date 11/7/73 Page 1 Rev. 1
t 2.0 F_IIFT OPERATIONS Plant operations are conducted by shift operating personnel assigned to the Operations Section under the direction of the Shift Supervisor. 2.1 SHIFT CCMPLDIENT Each operating shif t for Unit 3 will normally consist of a Shif t Super-visor, an Assistant Nuclear Shift Supervisor, a Chief Nuclear Operater, a Nuclear Operator, two Assistant Nuclear Operators, a Nuclear Auxt:iary Operator, and an Assistant Nuclear Auxiliary Operator. During cold shut-down or refueling conditions, the shift complement may be reduced to a minimum of a Shift Supervisor, a Nuclear Operator, and a Nuclear Auxiliary Operator. The total number and classification of personnel assigned to each shift will be determined by the Operations Superintendent as dic-tated by plant conditions and anticipated operations. In all cases, the minimum shift composition shall be in compliance with the provisions of Technical Specifications Table 6.2-1 below: MINIMUM SHIFT CREW LICENSE APPLICABLE MODES CCMPOSITION - CATECORY 1.2,3,4 5 and 6 SOL 1 1* OL 2 1 Non-Licensed 3 1 i NOTE: Shiftcrewcompositionmaybelessthantheminimumr9qu[rement for a period of time not to exceed 2 hrs in order to acconnodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1. 1755 086 i *Does not include the licensed Senior Reactor Operator or Senior Reactor Operator LIMITED TO FUEL HANDLING individual supervising core alterations after the initial fuel loading. Fase 2 AI-500 Date 10/25/79 Rev. 28
4
- s. In addition to the minimum shift requirements of Table 6.2-1, two licensed Reactor Operators shall be in the Control Center during startup and scheduled shutdown of the reactor and during recovery from reactor trips caused by transients or emergencies.
- b. At least one licensed Reactor Operator shall be in the reactor building when fuel handling operations are in progress in the reacter building. An operator holding a Senior Reactor Operator's License or a Senior Reactor Operater LIMITED TO FUEL HANDLING and assigned no other concurrent operational duties shall be -
in direct charge of refueling activities.
- c. (deleted) 2.2 _ ADDITIONAL SHIFT PERSONNEL Such additional operating personnel, as may be required because of unusual plant conditions or operational needs, will be obtained as directed by the Shift Supervisor. Normally, the Operations Superintendent is 1755 087 Page 3 AI-500 Date 4/13/78 Rev. 15
consulted when arrangements are nada for large numbers of operating per-sonnel, but this notification shall not restrict the Shift Supervisor fron obtaining such personnel as required for plant operations. It should be noted, however, that operations requiring sdditional personnel vill not Be undertaken until the shift is sufficis .Ay manned. 2.3 AUTHORITIES AND RESPONSIBILITIES PVR SAFE OPERATION ANi, SHUTDOWN The person charged with the duty of reactor operation has the authority and responsibility for shutting the reactor down to a safe condition by approved procedures when he determines that the safety of the reactor / plant is in jeopardy or when operating parameters exceed any of the reactv: pro-tection set points and automatic shutdown does not occur. Upon determina-tion of shutdown, he shall notify the Shif t Supervisor in charge who will insure that the plant is shut down in safe condition by approved procedures. The Shift Supervisor in charge has the responsibility to notify the Opera-tions Superintendent or, in his absence, the person on-call and together they determine the circumstances, analyze the cause, and determine that opera-tions can proceed safely before the reactor is returned to power af ter a trip or unscheduled or unexplained power reduction. Approval to take the reactor critical following a trip rests with the Operations Superintendent or, l in his absence, the persou on-call, but the actual startup shall be autho-rized by the Shif t Supervisor in charge and the authorization shall be doctmented in the reactor startup procedure. The Shift Supervisor in charge has responsibility to be present at the plant and to provide direction for returning the reactor to criticality o power following a trip or unscheduled or unexplained power reduction. 1755 088 Page 3a AI-500 Date 4/13/78 tav. 15
All persons charst.d with the duty of operating the plant have the respon-sibility to:
- a. Believe and respond conservatively to instrument indications unless they are proven to be incorrect by instrument channel check or instrument channel test.
- b. Adhere to Technical Specifications.
The Shift Supervisor in charge and the Operations Superintendent have the responsibility to review routine operating data to assure safe operation. 1755 089 Fage 3b AI-500 Date 4/13/78 Rev. 15
2.4 SHIFT FUNCTIONS The personnel assigned to shift operations perform or are prepared to perform three general functions:
- a. Continuous normal operation of the plant and its associated equipment, including normal planned power changes, startups, and shutdowns.
- b. Maintain the plant in a safe condition during abnormal conditions.
- c. Protect plant personnel, the health and safety of the public, and plant equipment during and following an emergency situa-tion.
All major plant operations are conducted from the Control Center in accordance with the Shift Supervisor's instructions. The Shift Super-visor, consistent with the provisions of P0QAM (plant Operating Quality Assurance Manual), effects the Load Dispatcher's orders and the direc-tions of pisnt staff supervisors. The Chift Supervisor hes the respon-sibility to maintain a broad perspective of operational conditions affecting plant safety as the matter of highest priority at all times when on duty. Although the Shif t Supervisor's normal duty ststion is the Control Center, he may be anywhere in the plant his attention is required. In his absence, the Assistant Nuclear Shift Supervisor shall assume duty in the Control Center and have responsibility for interpreting the Shift Supervisor's instructions and directing operations. 1755 090 t l Page 4 AI-500 Date 10/25/79 Rev. 28
During abnormal or emergency conditions where multiple operations are required, the Shift Supervisor shall assume a "consnand" role. He shall base his decisions on an overview of the condition and direct the activi-ties of the Control Center Operators to insure reactor safety. The Shift Supervisor shall remain in the Control Center at all times during as abnormal or emergency condition until properly relieved. Th3 persons authorized to relieve a Shif t Supervisor of his "consnand" duties are another Shift Supervisor, an Assistant Nuclear Shift Supervisor, or the Operations Superintendent. Emergency conditions may also require the Shift Supervisor to act as temporary Emergency Coordinator. Responsibilities and authorized relief for this function are defined in EM-100, Emergency Plan. During all conditions, the operating instructions of the Shift Supervisor may not be superceded except by the Nuclear Plant Manager, the Operations Super-intendent, or the person on-call. During all operational conditions, shift operations personnel will be guided by P0QAM, supervisory direction, and technical advice and assis-tance. The Assistant Nuclear Shif t Supervisor assine s.hr 5P.iit i@tr- 4 visor.by directing and overseeing the routine operations of the plant in accordance with the Shift Supervisor's general instrue:. ions and by performing the preparation and review of changee to P0QAM. He also provides assistance by assuring the timely completien of required administrative functions such as Radiation Work Permits (RWP's), Work Raquests, and Surveillance Procedures. The Chief Nuclear Operator provides direction and assistance to the operators in the field in the performance of assigned tasks. He is available to provide assistance 1755 091 Page 4a AI-500 Date 10/25/79 Rev. 28
in any area of the plant and makes periodic in-plant rounds to observe operating equipment. A Nuclear Operator and one Assistant Nuclear Operator will normally be in the Control Center and will be operating the plant controls and equipment and maintaining the operating logs. The other Assistant Nuclear Operator and a Nuclear Auxiliary Operator will be in the plant maHng rounds to observe operating equipment, record operating data, and execute routina operations. In "on-the-job" training functions, Assistant Nuclear Auxiliary Operators will be in the plant assisting the Nuclear Auxiliary Operator or performing work func-tions as dictated by plant conditions. At least one CR-3 licensed operator shall be in an area in front of the control board at all times with an unobstructed view of and access to the operational control board. The general area is defined as that area enclosed by the installed red tile stripping in front of the control board. In the event of an emergency affecting the safety of operations, the operator at the controls may momentarily be absent from the general area in front of the control board in order to verify the receipt of an annunciator alarm or initiate corrective action provided he remains within the confines of the Control Center and maintains an unobstructed view of the operational control panels. The operator at the controls should not under any circumstances leave the red lined general area for any non-emergency reason without obtaining a qualified relief operator at the controls. Both the Nuclear Operator and the Assistant Buclear Operator on duty in the Centrol Center shall at all times be prepared to assume the responsibility of the " operator at the controls" to allow either to leave the red lined general area for non-emergency reasons. 1755 092 Page 5 AI-500 Date 10/25/79 Rev. 28
The controls of the plant shall be manipulated by a licensed operator or Senior Operator except that an individual may manipulate the controls as part of his trainios to qualify for an operator license under the direc-tion and in the presence of "a licensed operator or Senior Operator". 2.5 CONTROL CENTER RECORDS Certain records are saintained either in the Control Center or the Shif t Supervisor's Office to facilitate consrunications and operations. These records will consist of but not be limited to the following:
- a. Control Center notebook containing short-term instructior.s, "on-call" notification, Reactor Trip Log, Generator Under Frequency Log, and In-Plant and System Switching and Tagging List.
- b. Shift Supervisor's Log
- c. Operator's Log
- d. Key Control Log
- e. Equipment Out-of-Service Log
- f. (deleted) 3 Current Radiation Work Permits
- h. (deleted)
- 1. Jumper Log
- j. Equipment C3earance Order Log 2.6 ON-CALL Either the Nuclear Plant Manager or his authorized designee shall be at the plant or on-call at all timas. When the Nuclear Plant Manager or his authorized designee is absent from the plant, during backshifts,
' 1755 093 AI-500 Date 4/26/79 Fame 5a Rev. 24
weekends, or holidays, he shall be provided with a voice pager by which the Shift Supervisor can contact him regarding plant problems. At times when use of the voice pager is impossible, he shall advise the Shif t Super-visor of the location and telephone number at which he may be contacted. Any Responsible Supervisor may be designated as on-call provided he holds a valid NRC Senior Reactor Operator's License. 2.7 PROCEDURES Operating Procedures are divided into two basic groups, those whi
- require step-by-step sign-off or initials for proper verification of vital proce-dural steps and those which are written as operator guides and require no sign-off's or approval.
Procedures dealing with systems which directly affect the reactor, reactor coolant (RC) system, and engineered safeguards (ES) vill be of the sign-off and app
- oval type while procedures dealing with secondary and auxiliary systems not directly affecting the reactor, RC system, or ES will be u'ai-lized as operator guides.
The procedures used as operator guides may be checked off as the step or section is completed, but procedure verification and/or retention is not required. Procedures requiring proper verification will be handled in the following way. When an operator has performed an operation or verified the status of the item as required by the specific procedural step, he initials the space provided for that step in the procedure or on a Check-Off List.
' 1755 094 AI-500 Date 3/22/79 Page 6 Rev. 23
Sp uences or steps in procedures that cannot be performed due to clear-ances, abnormal conditions, or emergency operating modes shall be marked as Not Applicable" (N/A) and the specific reason for not performing the step shall be noted and initialed by the Shift Supervisor prior to pro-caeding. When all procedural steps have been completed and the required spaces initialed, the operator signs and dates the procedure. The Shift Super-visor or Assistant Shift Supervisor attaches a Procedure Approval and Transmittal Sheet (Form 912019 - RE: AI-400, Section 5.6) to the com-plated procedure. He then reviews the completed procedure, completes and signs the Procedure Approval and Transmittal Sheet, and forwards the completed procedure to the Operations Superintendent. The Operations Superintendent reviews che completed procedure and forwards it to the Compliance Section for review. For the purposes of documentation and record retention, the Procedure Approval and Transmittal Sheet attached to the completed surveillance data sheets or the applicable procedural section(s) dealing with the function performed shall be considered a complete procedure. 2.7.1 Adherence to Procedures Refer to Sections 5.3, 5.4, and 6.0 of AI-400, Plant Operating Quality Assurance Manual Control Document.
}f)5 095
( Page 6a AI-500 Date 9/28/79 Rev. 27
2.8 C;ZRATOR RETRAINING See FSAR, Appendix 12C. 2.9 PLANT MODIFICATION AND SET POINT REVISION NOTIFICATION To promptly and effectively advise shift personnel of plant modifications and set point revisions, the Operations Superintendent will transmit to the Shift Supervisors a completed copy of Form 912163, Modification Approval Record (MAR), along with a Form 912235, Change Notification. Each Shif t Supervisor will insure that each member of hir shif t has been informed of the plant modification or set point revision and initial the Change Notification. Forms which have been initialed by all Shif t Supervisors shall be maintained in the Control Center until the modification is complete, then returned via the Operations Engineer to the Operations Superintendent for filing. 2.10 PLANT OPERATING QUALITT ASSURANCE MANUAL REVISION NOTIFICATION Temporary and permanent revisions are made to POQAM in accordance with Section 8.0 of AI-400, Plant Operating Quality Assurance Manual Control Document. All revisions made will use Form 912101, Procedure Review Record (PRR). When the revisions are approved, they will be transmitted along with a Form 912235, Change Notification, to the Shift Supervisor who will initial the Change Notification, signifying that each member of his shift has been informed of the revision. It is the responsibility of each Shift Supervisor to insure that his shift is informed. Forms which have been signed by all Shift Supervisors shall be returned via the Operations Engineer to the Operations Superintendent who will file the change form for the latest revision in the POQAM Revision log. ( 1755 096 Fase 7 AI-500 Date 10/25/79 Rev. 28
2.11 SUPPLEMENTAL LABORATORY ANALYSIS REQUEST Occasionally plant modes or conditions dictate that supplementary data be collected to support recovery from off-standard conditions. To insure that the additional data is collected and documented, Form 912225 (Enclo-sute 2), Supplemental 14boratory Analysis Raquest Form, shall be sent to the CaenEad Laboratory as conditions dictate. Because initial laboratory sampling may be required on an "as soon as possible" basis, verbal requests will be accepted from the Nuclear Operator, Chief Nuclear Operator, Assis-tant Shift Supervisor, or Shift Supervisor with the understanding that Form 912225 (Section 1 complete) will be provided to the techrJ eian by the time results are available for logging. If requasted in Stetion 1 of Form 912225, results can be verbally transmitted as each is logged on the form. Upon completion of Section 2, the form will be obtained from the ChenRad Offics. Se.ction 3 of the form is then completed in accordance with AI-500, Conduct of Operations, Sections 2.5(c) (Control Center Records), 2.7 (Procedures), and 2.20 (Control Center Status Board). 2.12 CENERAL PRACTICES FOR COLLECTING SHIFT RECORDS Shift records are comprised of the logs, data sheets, recorder charts, and computer printouts which describe or record operating information and action. The information obtained from these records is useful for current operations and for analysis of previous operations. The follow-ing general practices are applicable to shif t records:
- a. All los entries, data sheets, and chart notations must be legible, accurate, complete, and understandable.
1755 097 i Page 8 AI-500 Date 3/22/79 Rev. 23
- b. The individuals responsible for mafataining logs and data sheets are responsible for signing and dating the portions of the records which cover their shift assignments.
- c. Each in-service recorder chart shall be checked at least once per shift (normally during the first hour of the shif t) to verify that the marking is legible and the timing correct.
I 1755 098 Fase Sa AI-500 Date 12/22/77 Rev. 13
- d. Before the chart is placed on or removed from the recorder, it shall be marked with the date, recorder identification symbol, and parameter being recorded. It shall be date stamped daily on the 0000-0800 shift.
- a. When significant events or unusual trends in transients occur, the resulting raccrder traces are to be identified as to the time and event notation to assist in operations analysis,
- f. The Shif t Supervisor on each shift will review the data sheets and los records compiled and recorded during that shift. As a part of his tour through the plant, the Assistant Nuclear Shift Supetvisor will review log records and data sheets at operating stations outside the Control Center. This review is to detect unusual or abnormal trends or readings that require investiga-tion or remedial action and to check on the completeness and accuracy of the records. The Shift Supervisor or the Assistant Nuclear Shift Supervisor signifies completion cf this review by initialing the data sheet or log record and recording the time the review was made.
- g. At midnight of each day, the shift records are assembled and checked for completeness by the Shift Supervisor on duty. The shift records are then sent to the Operations Superintendent.
- h. The Operations Superintendent is responsible for reviewing shift records. He advises the respective Shift Supervisor of any deficiencies in cespiling the information and initiates measures to eliminate the deficiencies.
- i. The Assistant Nuclear Shift Supervisor is responsible for insuring that all supplies necessary for keeping shif t records I
(charts, ink, procedures, forms, logbooks, check-off sheets, etc.) are in ample supply. Page 9 AI-500 Date 10/25/79 1755 099
- j. Shif t operating logs will be taken once per shif t (normally during the first hour of the sh*f t) unless otherwise speci-fied.
2.13 SHIFT LOGS 1he narrative log notations of plant conditions, operations, and events are a vital portion of the shif t records. (Errors in a shift log are corrected by drawing a single line through the incorrect information and writing the correct information adjacent to or in space available with reference to the deleted information. The individual making the correc-tion shall initial and date the deleted information.) All entries shall be made in "olack ink.
- a. The Operator's Log is maintained on a shift basis to record the plant status and events in chronological order. Log entries may include but are not limited to the following:
- Date
- Names of Shift Personnel
- Plant Status
- Water Quantity Used for Makeup
- Waste Disposal System Status
- Starting and Stopping of Equipment
- Change of Auxiliary System Configuration
- Water Quantity Used for Dilution
- Watet Quantity and Its Source Used for Boration
- Performance of Surveillance Tests
- Completion of Specified Check-Off Lists
- Maintenance Activities Affecting Operations 4
- Occurrence of Significant Annunciator Alarms 1755 100 Page 10 AI-500 Date 12/22/77 Rev. 13
- Performance of Special Inspections or Checks (Overspeed Trip, Oil Filters, etc.)
- Reactor Trips
- Instrument or Equipment Malfunctions or Failures
- Unusual Trends or Cenditions Observed
- Company Electrical Grid Events Affecting Operations
- Relay Operaticas and Targets
- Starting and Stopping Caseous or Liquid Waste Disposal Dischargas (List release permit number.)
- Alarm Tests
- The number of the procedures used to perform any of the above operations.
At the end of each shift the Nuclear Operator signs the Operator's Log, signifying that the entries are a complete and accurate recoed of plant operations. The current logbook and the most recent "back copy" are to be retained in the Control Center.
- b. The Shift Supervisor's Logbook is a multiple-sheet record form in n hardbour.d book. The Shift Supervisor sununarizes plant conditions and events during his shift. Events entered may include any of those noted in the Operator's Log, however, the shift repert need not repeat routine items which have no safety significance or little operational importance, but will include a detailed explanation of major events.
The Shift Supervisor's Log shall begin with plant status infor-metion and should include any changes in status of the avail-ability of systems, unusual occurrences, results of liquid or ' gaseous releases, sample analysis, and changes of major aux-iliary equipment service. At the end of each shifc, the Shift Page 11 AI-500 Date 7/13/78 1755 101
Supervisor signs the log signifying that the report is a com-prehensive, accurate summary of plant events and activities. At the end of the 24 hr. period, the copy of the Shift Supervisor's logsheet is detached from the book and forwarded to the Operations Superintendent for review. The original temaining in the book is retained in the Control Center for reference and record. All log-books except the current one and the most recent "back copy" are transmitted to the Administrative Supervisor for disposition. 2.14 REACTOR TRIP AND PLANT SHUTDOW'N A reactor trip is any reactor protection system (RPS) action, manual or automatic, which causes the de-energizing of the control rod drive necha-nisms (CRDM's) and allows the control rod assemblies to drop into the core. A plant shutdown is the opening of generator breakers 1661 and 1662. When a reactor trip or plant shutdown occurs, the Shift Supervisor takes the following action:
- a. Insures that the plant is placed in a safe condition by having the necessary operations performed in accordance with approved procedures.
- b. Notifies the Operations Superintendent or person on-call.
- c. Determines the subsequent action to be taken.
~
- d. Completes Steps 1 thru 8 on Form 912212 (Enclosure 3), Reactor Trip and Shutdown Report, assigns the next consecutive report number, and forwards it to the Operations Superintendent for
. disposition. A log of Reactor Trip and Shutdown Report dates and types will be maintained in the Control Cancer notebook.
- a. Insures the reactor trip Laformation is entered in the Reactor i
Operator's Log and Shif t Supervisor's Log (reactor trip only).
'.ge 2 17551])2
^'-5 i' !?"'
The res cor will not be taken critical following a reactor trip until a determination of the cause of the trip has been made, corrective action taken, and approval to take the reactor critical has been obtained from t.he Operations Superintendent or person on-call. In the interim between trip and approval for recovery, the Shift Super-visor may authorize the withdrawal of Safety Group 1 provided a > 1% ok/k shutdown margin is maintained and rod withdrawal is not prohibited by any RPS action statements of Standard Technical Specifications. 2.15 SHORT-TERM INSTRUCTIONS Short-term instructions are any miscellaneous instructions that may arise and shall be used for routine maintenance and personnel instruction where the plant safety is not affected. All short-term instructions shall auto-antically expire in 90 days if not previously cancelled. It is the Shift Supervisor's responsibility to review, audit, and maintain the short-term instructions in a current and up-to-date condition. If it is necessary to continue the short-term instruction, it shall be reissued using a new document number. Short-term instructions shall not be amended. If a change is necessary to an issued short-term instruction, it shall be cancelled and reissued in its correct. form. To complete the Short-Tern Instruction Form (Enclosure 6), insert the document number. The document number will be prefixed by the year, then the next progressive short-tcra instruction number; insert the date.
' 1755 103 AI-500 Date 4/13/78 Page 13 -Rev. 15
Com-The length of effect sh'all be as conditions dictate up to 90 days. plate the instruction and action as required. The short-term instruction is imunediately effective when signed by a Shift Supervisor or the Opera-tions Superintendent. Any deviation from short-term instructions shall be logged in the Control Cancer Iogbook and reported to the Shift Super-visor on duty. The instructions are logged on the Short-Term Instructions Index (Enclosure 6). log entry is completed by inserting the document number, subject, and date entered. When it is necessary to cancel a short-term instruction, this may be performed by completing the "Date Removed" and " Removed By" portions of Enclosure 6. Official instruction removal occurs with the initials of a Shift Supervisor or the Operations Superintendent. Cancelled short-term instructions shall be destroyed. 2.16 CONTROL CENTER ACCESS The Shif t Supervisor is responsible for maintaining control of personnel entering the Control Center whether entry is for observing operations, conducting tests, or performing maintenance. The Shift Supervisor is authorized to refuse entry or direct personnel to leave the Control Center. When an operational transient or accident occurs, the Shift Supervisor, Assistant Nuclear Shift Supervisor, and other Operations personnel in the Control Center shall have an unobstructed view of and insnediate access to the operational controls. Additionally, these personnel shall have access to operational auxiliaries such as the computer line printer, the IBM-5100 computer, and the Environmental Monitoring Panel. The Shift Supervisor or Assistant Nuclear Shift Supervisor i 1755 104 AI-500 Date 10/25/79 Page 13a Rev. 28
shall limit the Control Cente- to only those personnel who are essential for the direct operation of the plant and to technical advisors relative to support the particular operating condition. Based on the specific operational condition and personnel essentiality to that condition, the Nuclear Plant Manager or Operations Superintendent any authorize reprogramming of the security key-card system to restrict access to the Control Center. i 1755 105 Fase 13b AI-500 Date 10/25/79 Rev. 28
2.17 SHIFT RELIEF Shift operating personnel, when on duty, are to remain on duty with full responsibilities of their position until properly relieved. A Shift Supervisor, Assistant Nuclear Shift Supervisor, Chief Nuclear Operator, Nuclear Operator, Assistant Nuclear Operator, Nuclear Auxiliary Operator, or Assistant Nuclear Auxiliary Operator is considered to be properly relieved eben the individual assigned to relieve him is properly quali-find and licensed (if r, , aired) to assume the position and has been informed of the status of the plant, operations in progress, and special instructions if any are applicable. The relieving individual has the responsibility of (1) reviewing the Operator's Log, status board, Equip-ment Out-of-Service Log, and data sheets; (2) discussing operations with on-duty personnel; (3) reading special instructions if any are applicable. These responsibilities shall be carried out as soon after relief as plant conditions will allow. Each Shift Supervisor, Assistant Nuclear Shift Supervisor, Chief Nuclear Operator, Nuclear Operator, and Assistant Nuclear Operator will be responsible for reading the Operator's Log entries back to his last scheduled shift and so indicate by placing his initials in the space provided at the end of the shif t immediately pre-ceding his return to operational duties. Shif t Supervisors and Assis-tant Nuclear Shift Supervisors shall also review the Jumper Log, Shift Supervisor's 143. Equipment Out-of-Service Img, and short-term instruc-tions. The oncoming Shift Supervisor shall document his relief by com-plating Enclosure 9. Shift Relief Checklist, while dicussing operations with the on-duty Shift Supervisor. Immediately after turnover, the on-duty Shift Supervisor shall complete Enclosures 10 and 11 to assure operational and emergency systems' status. Page 14 AI-500 Date 10/25/79 Rev. 28
2.18 INSTRUMENT READINGS AND CONTROL INDICATIO g Plant operations are conducted by shift operating personnel under the direction of the Shift Supervisor. They have the authority and respon-sibility to perform the operations necessary to limit plant operations or shut down the plant when such action is warranted by plant conditions, unusual circumstances, or unidentified events. Such actions may be war-ranted on the basis of instrument readings and/or control indications which are not consistent with expected plant conditions. When analyzing such situations, shift operating personnel must consider the instrument readings and correct indications to be true unless they are proven to be incorrect. Operating personnel will not manipulate instrument, control, or alarm set points other than those available on the control console or those normally required during routine operations. 2.19 KEY CONTROL Keys are required for the operation of certain switches, valves, switch-gear, and for access to particular rooms and areas. The components or 1755 107 Page 14a AI-500 Date 10/25/79 Rev. 28
areas requiring kef" control are those with special operational i portance or safety significance. A list identifying each key-controlled item and area and the respective key identification is maintained in the Control Center. All keys on this list are identified as " Controlled Keys". These keys are maintained in a locked cabinet under the control of the Shift Supervisor. Permission of the Shift Supervisor or Assistant Nuclear Shif t Supervisor is required for the use of a " Controlled Key". Before authorizing the use of a " Controlled Key", the Shif t Supervisor must assure himself that the individual intending to use the equipment or enter en area under key control understands and appreciates the particular operational or safety requirements associated with the equipment or area. At the completion of the operation or task, the Shift Supervisor vill assure himself that conditions are proper for placing the lock control in the proper position or locking a door or barrier and returning the key to the key control cabinet. The individual using the equipment or area under key control vill be responsible for insuring that unauthorized personnel do not use the equipment or gain access to the area during the period when the control device is unlocked.
" Controlled Keys" are issued to certain "0N DUTY" operating classifications; possession of such keys is necessary to perform normal and emergency uuties within their area of responsibility. The " Controlled Keys" remain on the operator's person during his assigned hours of duty and are not permitted to be used by unauthorized individuals.
1755 108 Page 15 AI-500 Date 2/16/78 Rev. 14
Entries are made in the Key Cc strol Log, form 912220 (Enclosure 4), which show the porpose for which a " Controlled Key" was required, who used it, badge number, and the time if was returned to the key control cabinet. When the Key Control Logsheet has been completely filled out and all keys listed on that sheet have been returned, the logsheet shall be sent to the Operations Superintendent for disposition. 2.20 CONTROL CENTER STATUS BOARD The Control Center Status Board provides a visual display of information for immediate reference by Control Center personnel. This information includes the condition of engineered safeguards equipment, storage tank levels, and the chemical analysis of the coolant. The status board infor-nation is derived from the Equipment Out-of-Service Log and other plant i 1755 109 AI-500 Date 4/26/79 Page 15a Rev. 24
records. The Shift Supervisor is responsible for insuring that the status board is updated as new information becomes available. 2.21 CONTROL CENTER REFERENCES To assist shif t operating personnel in the conduct or their duties, ref-erence information related to any aspect of plant operation, safety, and administration is permitted in the Control Center. To meet these needs, the Operations Superintendent will insure that current copies of the reference information listed below are available in the Control Center.
- a. Final Safety Analysis Report (FSAR)
- b. Technical Specifications
- c. System Flow Diagrams and One Line Electrical Diagrams
- d. Working Copies of POQAM Addit,ional reference matter may be kept in the Control Center at the discretion of the Operations Superintendent.
1755 110 Page 16 AI-500 Date 4/26/79 Rev. 24
Typical of this category of reference documents are vendor technical man-uals, training course notes, engineering handbooks, and technical texts. All such material shall be subject to approval by the Operations Superintendent. 2.22 NOTIFICATION Many plant conditions and operating situations are of such a nature that it is necessary or prudent to promptly advise the Operations Superintendent or person on-call of the circumstances. The Shift Supervisor must utilize his judgement and experience in assessing the need for such notification which implicitly includes obtaining advice, assistance, and direction from the Operations Superintendent or person on-call. The following situations require prompt, verbal notification:
- a. Raaetor i' rip
- b. Inadvertent (radioactivity bearing) liquid or gaseous releases.
- c. Major equipment failure or malfunction (includes all safe-guards equipment).
- d. Unexplained reactivity changes.
- e. Loss of off-site power.
- f. Employee injury or radiation overexposure.
- g. Accidents occurring on plant property (except minor injuh).
- h. Events requiring reports within 24 hours to the NRC Region II Office. l (Terhaie=1 Specifications 6.6 and 6.9)
- 1. Turbine Trip
- j. Load Restrictions
- The Shift Supervisor shall note in his log when he notifies the Operations i
Superintendent or person on-call. 1755 111 Page 17 AI-500 Date 9/28/79 Rev. 27
2.23 BOUSEKEEPING In the interest of safe and efficient operation, plant equipment and arena aust be semintained in a clean and orderly manner. The responsi-bility for keeping the Control Center in this condition is that of the shift operating personnel. All dusting and cleaning of control consoles, instrument panels, and computer consoles, and the orderly storage of becks, drawings, and records, will be parformed by shift operating per-sonnel. In all areas outside the Control Canter, the responsibility for insuring the Operations Section's work areas are maintained in a clean and orderly condition rests with the Shift Supervisor or, as in the case of fuel handling operations, the Refueling Supervisor. 2.24 EQUIPMENT OUT-OF-SERVICE LOG The Equipment Out-of-Service Log shall be updated and maintained by Control Center personnel. Any safety-related equipment taken out of service shall be recorded in this log en Enclosure'7. The information provided shall include the nature of the problem, the date taken out of service, the date the equipment must be returned to service (if applicable), clearance number, Technical Specification reference (if applicable), and the date the equipment is returned to service. The Shift Supervisor shall ensure this log is being maintained and up-to-date. This log shall be reviewed at shift turnover as required by Section 2.17. When the Equipment Out-of-Service Ing has been completely filled out and all listed equipment returned to service, the logsheet l shall be transmitted to tho Operations Superintendent for review / dispositin. 1755 112 Fage 18 AI-500 Date 8/9/79 Rav. 26
Engineered safeguards systems or equipment shall have independent veri-fication of proper valve alignment per Step 6.1.9 of CP-115. In-Plant Equipment Clearance and Switching Orders, and applicable surveillance performed before returning equipment to service. 2.25 UNUSUAL OPERATING EVENTS REPORT Unusual operating events include, but are not limited to, reactor trips, operating events involving significant unexpected behavior, or signifi-cant unusual operating situations such as changing RC pump combinations at power or inadvertent equipment operations. When an unusual operating event occurs, the Shift Supervisor takes the following action:
- a. Insures that the plant is placed in a safe condition by having the necessary operations performed in accordance with approved procedures.
NOTE: If the event is a reactor trip, refer to Section 2.14. i
- b. Notifies the Operations Superintendent or person on-call.
- c. Determines subsequent action to be taken.
- d. Insures the unusual operating event information is entered ,
in the Reactor Operators Log and Shift Supervisors Log.
- a. Completes Steps 1 thru 9 of Enclosure 8, Unusual Operating Event Sumary, and forwards Enclosure 8, including the required information of Step 8, to the Operations Engineer.
The Operations Engineer will collate the Unusual Operating Event Summary { and prepara an Unusual Operating Events Report. The report will be divided into four major sections: event synopsis, evaluation and i 1755 113 Fage iba AI-500 Date 9/28/79 Rev. 27
ecoseendations, event details and data, and transient classification. The " Evaluation and Reconnendations" section will be further subdivided to discuss expected plant performance, performance deviations, and recommended corrective action. The Unusual Operating Events Report will be routed through the Operations Superintendent and Rasults Engi-neer to the plant files (3-0-1-e) and will also be transmitted via AI-500, Conduct of operations.
' 1755 114 AI-500 Date 9/28/79 Fage 18b Rev. 27
3.0 NON-CONFORMING OPERATIONS Non-conforming operations are defined and their identification and report-ing discussed in CP-lli, Procedure for Documenting the Reporting and Review of Non-Conforming Operations. I i i 9 I i l I ' 1755 115 AI-500 Date 12/2/76 Fase 19 Rev. 10
4 4.0 WORK REQUESTS Work Raquests are used by plant personnel to report to the appropriate Maintenance Section equipment or system deficiencies. The Work Request, Form 912104, is described in detail in CP-113, Procedure for Handling Work Raquests, Including Discrepancies and Corrective Actions. 4.1 If the Work Raquest is originated by the Operations Section, the operator originating the Work Request completes Part I and submits four copies to the Shift Supervisor on duty. The Shift Supervisor then determines the following:
- a. Is the request of sufficient urgency to call out maintenance personnel? If so, notify the respective supervisor of the need for maintenance.
- b. Will the performance of the work by maintenance personnel have an impact on plant safety and/or operations? If so, Lunediately no':1fy the Responsible Supervisor of any precau-tions or limitations which need to be imposed.
- c. Has work been requested before? If so, how long ago? If time seems too long, notify Operations Engineer, but void new Work Raquest.
4.1.1 The Shift Supervisor signs the Work Request in the " Responsible Super-visor" blank of Part I. However, if the Responsible Supervisor is indeed the originator, then the " Responsible Supervisor" blank should'be marked ,
"N/A" (CP-113, Procedure for Handling Work Requests, Including Discrepan-1755 116 AI-500 Date 10/25/79 Fage 20 Rev. 28
cies and Corrective Actions; Step 4.1.4). He then files copy "D" (gold) in the Control Canter Work Raquest File. The renaining copies are sent to the Planning Coordinator (Maintenance) or Work Supervisor (documenta-tion) for review and distribution per CP-113, Procedure for Handling Work Raquests, Including Discrepancies and Corrective Actions. If Operations personnel are to accomplish the work activity, the Shif t Supervisor shall complete Part II of the Work Request as required by Section 4.2 of CP-113, Procedure for Handling Work Requests, Including Discrepancies and Correc-tive Actions. NOTE: During off-hours, the evaluation of Part II of the Work Request shall be completed by the Shif t Supervisor for various depart-ments as priorities dictate. Refer to Section 4.1 above and Section 4.2 of CP-ll3, Procedure for Handling Work Requests, i Including Discrepancies and Corrective Actions. < 4.1.2 When the Quality Control Inspector completes his review of the Work Raquest (Steps 4.6.2 thru 4.6.4 of CP-il3, Procedure for Handling Work Raquests, Including Discrepancies and Corrective Actions), the complaced copy "C" (pink) is returned to the Shift Supervisor. The Shift Super-visor will remove copy "D" (gold) from the Outstanding Status File, destroy it, and route copy "C" (pink) back to the originator for infor-nation. ,
' 1755 117 Page 20e AI-500 Date 6/15/78 Rev. 16
3.0 RADIATION WORK PERMITS Radiation Work Permits (RWP's) are used to authorize specific individuals to enter areas to work on equipment where the possibility of radiation exposure or contamination exists. After the RWP has been initiated by the department responsible for the work to be performed and approved by the ChenRad Section, the three forms are taken to the Control Center where the Shif t Supervisor determines if the work can be performed without interfering with plant operations, makes arrangements ior any clearances needed, and notes any special conditions or restrictions required by the Operations Section on the RWP. When the clearances are complete, the Shif t Supervisor signs the RWP and work can begin. The Control Center copy of the RWP is posted in the Con-trol Center. The ChemRad Section vill notify the Control Center when the RWP is closed out and the Control Center copy shall be removed from the Control Center and sent to the Health Physics Supervisor. ' 1755 118 Page 21 Al-500 Date 12/22/77 Rev. 13
6.0 EQUIPMENT OUTAGES When necessary the Operations Section will recommend and schedule the This is plant operating configuration required for equipment repair. the responsibility of the Operating Engineer who is the Staff Engineer reporting to the Operations Superintendent. In addition to his function as a technical advisor, the Operating Engineer will coordinate the functions of Operations, Technical Services, Maintenance and outside contractors during scheduled outage periods. . 1755 119 Page 21a AI-500 Date 2/8/79 Rev. 21
6 8 ENCLOSURES Enclosure 1 Form 912235, Change Notification Enclosure 2 Form 912225, Supplemental Laboratory Analysis Request Form Enclosure 3 Form 912212, Shutdown Report Enclosure 4 Form 912220 Kay control Log Enclosure 5 Form 912226, Short-Term Instruction Enclosure 6 Form 912227, Short-Term Instructions Index Enclosure 7 Equipment Out-of-Service Log Enclosure 8 Unusual Operating Event Summary Enclosure 9 Shift Relief Checklist Enclosure 10 Operational Status Checklist Enclosure 11 Critical Plant Equipment / Parameters Checklist 1755 120 AI-500 Date 10/25/79 Fase 22 Rev. 28
ENCI.0SURE 1
' 8 CHANGE NOTIFICATION
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SUPERVISORS: - The attached is an approved change. Af ter noting and informing your people, so indicate by dating and initialing below: MAR NO./ TITLE REV. PROCEDURE NO./ TITLE PERMANENT PAGES AFFECTED Details: Shift Supervisor .Tnittels/Date Asst. Nuclear Shift Supv. Initials /Date Operations Engineer
- Operations Superintendent i .
1755 121 Page 23 AI-500 Date 10/25/79 9122 Rev. 28
.5UPPLEMENTRL LRBDRRTDRY RNRLYSIS REGUEST FORM SECTION I REQUEST SYSTEM / SRMPLE PRlGIN:
RER5DN FOR REGUE5T: RNRLY515 REGUESTED: BORDN DTHERe RND SPECIFY; SPEClRL INSTRUCTIDN5: DR151NRTDR: DRTE: SHIFT SUPERV150R: DRTE: TIME: TIME: SECTlDN 2 RESULT 5 OflTE TIE INIT. PfRttCTG
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IMS GY R IDMS (151T5) IlQUUtK5: 17s; 122 SECTION 3 RCCEPTRNCE * (RESULT 5 IN DPERRTOR5.LDG: INIT. STRTUS BORRD UPDRTED: INIT.
- RTTRCH TD PROCEDURE RPPROVRL RND TRRNSMITTRL 5HEET RND RDUTE FOR STORRGE UNDER RI-EBB Fage 24 AI-500 Date 3/22/79 Rev. 23 gl}}}{
' ENCLOSURE 3 FLORIDA POWER CORPORATION CRT3TAL RIVER UNIT 3 SHUTDOWN REPORT NO.
- 1. TIME AND DATE OF SHUTDOWN
- 2. TTPZ OF SHUTDOWN: Forced Scheduled NOTE: Forced shutdowns are th6se which must be initiated no later than the weekend following discovery of an abnormal system condition.
Manual Trip
- 3. METHOD OF SHUTDOWN: Manual Runback Automatic Trip Other 2 (Explain in Step 9.) !
- 4. POWER PRIOR TO SHUTDOWN %
Power % MWe
- 5. PIANT STATUS AFTER SHUTDOWN: Mode
- 6. APPARENT CAUSE OF SHUTDOWN
- 7. DURATION OF OUTAGE
- 8. CORRECTIVE ACTION
- 9. REMARKS Shift Supervisor Operations Superintendent 1755 123 Results Engineer
# ***"* ""* *I" **
Page 25 AI-500 Date 2/8/79 Rev. 21 912 212
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ENCLOSURE 5 SHORT-TERM INSTRUCTION Date Document No. to Length of ; tact: From Instruction: Action: Issued By 1755 125 1 Fage 27 AI-500 Date 2/8/79 Rev. 21 912 22
- ENCLOSURE 6 SHORT-TERM INSTRUCTIONS INDEX h l
N NO. SUBJECT DATE ENTERED DATE REMOVED RDiOVED BY' I i i i l 912 227 1755 126 Page 28 AI-500 Date 2/8/79 Rev. 21
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ENCLOSURE 8 (Page 1 of 2) UNUSUAL OPERATING EVENT
SUMMARY
- 1. TIME AND DATE OF EVENT:
- 2. TYPE OF EVENT (i.e. , reactor trip, significant unexpected behavior during a planned evolution or transient, etc.):
Power % MWe
- 3. PLANT STATUS PRIOR 'ID EVENT: Mode Mode Power % MWe
- 4. PLANT STATUS AFTER EVENT:
- 5. APPARENT CAUSE OF EVENT: (Use attachments as required.)
- 6. DURATION OF EVENT (i.e., time in degraded condition, approximate time from event to discovery, etc.):
- 7. CORRICTIVE ACTION: (Use attachments as required.)
- 8. REQUIRED INFORMATION CHECKI,IST:
NOTE: This information should be as complete as practical and should include data from a minimum of 1 hr. prior to the svent.
- a. Operating Strip Charts:
[0TSG Levels, T(ave), Pressurizer Level, FW Flows, WR/NR, RC Pressure. Header Pressure, Reactor Power, Thot, MWe} Denote point of event on each strip chart and assure time, date, and recorder number are on each strip chart.
- b. Computer Alarms Summary Princout:
- c. Annunciator Events Printout:
1755 128 Page 30 AI-500 Date 9/28/79 Rev. 27
4 ' ENCLOSURE 8 (Page 2 of 2) UNUSUAL OPERATDIG EVENT SID0fARY-(Cont'd)
- 8. REQUIRED INFORMATION CHECKLIST:
- d. Post-Trip Review Summary: (Reactor Trip Only)
- e. Operators Log:
- (Include at least one shift prior to event.)
- f. Shift Supervisors Log:
(Include at least one shift prior to event.)
- g. Event Susunaries:
NOTE: Each operation classification that observed or took part in the event shall summarize the segrence prior to, during, and after the unusual event to the best of his recollection. Each person should prepare his own reconstruction of the event in detail. If plant conditions permit, this sumary should be written immediately after the event, otherwise it shall be written after shift turnover and prior to leaving the plant site.
- 9. REMARKS:
1755 129 Fage 31 AI-500 Date 9/28/79 Rev. 27
ENCLOSURE 9 (Page 1 of 2) SHIFT REI'EF CHECKLIST Information obtained from off-going shift:
- a. Plant Status Mode NI Power Level %
MWe MWth (Group 21)
- b. Operations in Progress (specify)
- c. Special Instructions (explain)
- d. Equipment in Degraded Mode
- e. " Action Statement" in Effect and Total Time Interval
, in " Action Statement"
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NOTE: Compare with allowed time interval. 1755 130 Page 32 AI-500 Date 10/25/79 Rav. 28
0 ENCLOSURE 9 (Page 2 of 2) SHIFT RELIEF CHECKLIST (Continued)
- f. Equipment Under Maintenance or Test (specify)
- g. Commener Time /Date/ Signature Off-Going Shift / /
On-Going Shift / / 1755 131 AI-500 Date 10/25/79 Page 33 Rev. 28
ENCLOSURE 10 OPERATIONA1, STATUS CHECKLIST Check (/)
- a. Master Surveillance Plan
- b. Jumper Iag
- c. Equipment Out.-of-Service Log
- d. Short-Tens Instructions
- e. Shift Supervisors Ing
- f. Operators Iog 3 Plant Status Board
- h. Annunciator Alarms Time /Date/ Signature
/ /
NOTE: The " check" on each item above signifies completion of review of that item.
. 1755 132 AI-500 Date 10/25/79 Page 34 Rev. 28
' ENCI.0SURE 11 (Page 1 of 4)
B CRITICAL PLANT EQUIPMENT / PARAMETERS CHECKLIST A. The Shift Supervisor shall assure by direct observation of controls or indication that the following equipment / parameters are aligned as speci-fied. NOTE: A check (/) signifies status as specified. To be completed at the beginning of each shif t.
- 1. Radiation Monitorina Panel Check for abnormally high readings, alarms of bypassed channels.
- 2. ES Channel "A" "th nnel Function Enabled" lights lit.
" Bypass / Reset" lights lit.
No Tripped Channels All ES equipment in normal standby status. DEP-1A SWP-1A RWP-3A MUP-1B running DCP-1A AHF-1A MUP-1A AHF-15A RWP-2A BSP-1A AHF-1C RB sump pumps in normal standby. All valves on ECCS panel properly aligned with power available.
- 3. ES Channel "B"
" Channel Function Enabled" lights lit.
" Bypass / Reset" lights lit.
No Tripped Channels All ES equipment in standby status. DEP-1B AHF-lc INP-35 - AHF-15B - DCP-13 BSP-1B MUP-1C IWP-2B SWP-13 All valves on ECCS panel properly aligned with power available. 1755 133 Page 35 AI-500 Date 10/25/79 Rev. 28
ENCLOSURE 11 . (Page 2 of 4) CRITICAL PLANI EQUIPMENT / PARAMETERS CHECKLIST (Continued)
- 4. PSA Panel EFP-1 in normal standby.
ASV-5 in " Auto". MSIV's air supply test switches in " Normal". ItS line rupture matrix status lights normal. MU&P loop proper lineup. No abnormal unexplained trends on MUT level recorder.
- 5. ICS Panel ,
All ICS stations in " Auto". Pressurizer heaters and spray valve in " Auto". MUV-31 in " Auto". No abnormal trends on strip chart recorders. Recorders are at proper tire. Feedvater valves in " Auto": FW-14 FW-31 FW-30 FW-15 FW-36 FW-28 FW-32 FW-29 FW-33 All four pump recirc. valves. No air failures lit.
- 6. TG Panel Turbine backup lube oil and EH pumps in normal standby.
Turbine in ICS control. All turbine drains closed. H2 Pressure Normal
- 7. Electrical Distribution Voltaga on 230 kV line normal.
If 4160V or 6900V unit buses are being fed from auxiliary transformer: Verify " Auto Transfer" switches in " Auto": ng.. 3.. ( 1755 134 AI-500 Date 10/25/79 Fage 36 Rev. 28
ENCLOSURE 11 (Page 3 of 4) 4*
.M CRITICAL PLANT EQUIPMENT / PARAMETERS CHECKLIST (Continued)
I
- 7. Electrical Distribution (Cont'd)
ES buses being fed from SU transformer: "A" "B" Emergency diesel generator start circuit lights lit:
. "A" Diesel Generator "A" "B"
"B" Diesel Generator "A" "B"
Emergency diesel generator high speed light on.
"A" "B"
Emergency diesel generator start mode and voltage adjust in " Auto":
"A" "B"
Emergency diesel generator breakers' position indication light on and target matched. Ihe following bus voltages are normal: 4160V ES "A" 480V ES "A" 480V ES "AB" 4160V ES "B" 480V ES "B"
- 8. HVAC Panel Ihe following recorders are reading normal, inking properly, and are at proper time:
AH-35-FR AH-717-FR AH-294-FR AH-32-FR No unexplained target mismatch on any ventilation equipment. ( 1755 135 Page 37 AI-500 Date 10/25/79 Rev. 28
4 ENCLOSURE 11 (Page 4 of 4) 4 g CRITICAL PLANT EQUIPMENT / PARAMETERS CHECKLIST ( (Continued)
- 3. If any equipment / parameter is found contrary to that specified in the checklist:
- 1. Describe immediate action taken to correct the condition. (and)
- 2. Explain actions taken, instructions given, or recommendations to prevent recurrence, Time /Date/ Shift Supervisor Signature f
0000-0800 / / 0800-1600 '/ / 1600-2400 / / 4 1755 136 Page 38 AI-500 Date 10/25/79 Rev. 28
ATTACHMENT VI Rodda
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INTEROFFICE CORRESPONDENCE productton (Offitti C-4 III All topt) SutitCT: Management Responsibility of Nuclear Shift Supervisor 10: Nuclear Plant Personnel part: December 3, 1979 Alpha 19 The purpose of t his direct ive is to emphasize the primary m.magement responsibility of the Shif t Supervisor and clearly establish his command duties. The primary management responsthIIity of the NucIcar Shit t Supervisor is to provide direct command oversight of plant operatlong and perform a management review of ongoing operations, maint enance, and support funettons import ant to safety (i .e. , to maintain an overview of the situation, to make decisions, and to direct operations). During back shifts and weekends, he is the Senior Management Representative on site and all personnel on site report to him. The Nuclear Shift Supervisor's command duties require that he be on duty in the Control Room or Shift Supervisor's office and that he maintain the broadest perspective of operational conditions affecting the safety of the plant as a matter of highest priority. During accident situations, the Shift Supervisor shall remain in the Control Room to direct the activities of Nuclear Operators until proper-ly relieved by another Shift Supervisor or Assistant Shift Supervisor. If the Shift Supervisor is temporarily absent from the Control Room during routine operations, the Assistant Shift Supervisor shall assume the Control Room command function with all the responsibilit ies and authority of the Shift Supervisor. When functioning as Temporary Emergency Coordinator, the Shift Supervisor has full authority to evaluate and classify the emergency and initiate appropriate actions to mitigate the consequences. Should his evaluation indicate t hat extreme measures must he taken, he - has the authority to direct any or all personnel to evacuat e t he plant site, to place any or all generating plants in a safe shutdown condition, and to notify all applicable agencies of the plant status or required outside assistance. The Nuclear Shif t Supervisor is fully supported by corporate and plant management in carrying out the above responsibilities.
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xc: NRC Region 1 0'G. C. Moore N - gr j 77 Office of inspection Assistant Vice president J3 iJI and Enforcement Power Production ese reg <ti
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ATTAQ9ENT Wf Rev. 13 8/9/79 t ( \ - (.' Section U NFOR 1c. R. Nuclear ON M L
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mulxI.Tunvi nsTRucTIOuS AI-200 , FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 b OaGANIZATION AND RESPONSIBILITY
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TABLE OF CONTENTS Page Section 1
1.0 INTRODUCTION
2 2.0 ORGANIZATION . . . . ...... ............ NUCLEAR GENERAL REVIEW COMMITTEE , ........... 3 3.0
......... ........ 4 4.0 PRODUCTION DEPARTMENI 4
4.1 SYSTD4 PLANI OPERATIONS ............... 4 4.2 PRODUCTION DEPARTMENT STAFF ............. 5 5.0 PLANT ORGANIZATION . ... ............... OPERATIONS SUPERINTENDENT .............. 5 5.1
...... . ,......... 9 5.2 OPERATING ENGINEER .
10 , 5.3 MAINTENANCE SUPERINTENDENT . ............. TECHNICAL SERVICES SUPERINTENDENT .......... 12 5.4 12 5.5 NUCLEAR QA/QC COMPLIANCE MANAGER . .......... CHDilSTRY AND RADIATION PROTECTION ENGINEER ..... 14 5.6 PERFORMANCE ENGINEERING SUPERVISOR . ......... 18 5.7 TECHNICAL SU?? ORT ENGINEER , ............. 20 5.8 21 5.9 ADMINISTRATIVE SUPERVISOR .............. PLANT REVIEW COMMITTEE . ................ 27 6.0 . 28 7.0 REQUIRED STAFF QUALIFICATIONS BY POSITION ....... NUCLEAR PLANT MANAGER ................ 28 7.1 TECHNICAL SERVICES SECTION . . ............ 28 7.2 28 7.2.1 Technical Services Superintendent ......... PERFORMANCE ENGINEERING ............... 29 7.2.2 29 7.2.2.1 Performance Engineering Supervisor . . . . . . . . . 29 7.2.2.2 Results Engineer . . . . . . . . . . . . . . . . . . Computer & Controls Engineer . . . . . . . ..... 29 7.2.2.3 29 7.2.2.4 Reactor Engineer . . . . . . . . . . . . . . . . . . 31 7.2.3 CHEMISTRY AND RADIATION PROTECTION . ......... ChemRad Protection Engineer ............ 31 7.2.3.1 31 7.2.3.2 Assistant ChemRad Protection Engineer ....... Plant Engineer (ChemRad) . . ............ 31 7.2.3.3 31 7.2.3.4 Radiation Waste Supervisor . . . . . . . . . . . . . 1755 139 Page (i) AI-200 Date 8/9/79 Rev. 13
TABLE OF CONTENIS Page Section (Cont'd) 32 7.2.3.5 Health Physics Supervisor . . . . . . . . . . . . . 32 2 2.3.6 Chief ChemRad Technician ............ . 32 7.2.3.7 ChemRad Technician ............... . 32 7.2.3.8 Assistant ChamRad Technician ........... 32 7.2.4 TECHNICAL SUPPORT . . . .......... ..... 33 7.2.4.1 Technical Support Engineer ............ 33 7.2.4.2 Inservice Inspection Engineer . . . . . . ..... 33 7.2.4.3 Nuclear Technical Specifications Coordinator . . . 33 7.2.4.4 Plant Engineers . . . . . . . . . . . . . . . . . . 33 7.2.4.5 Engineering Assistant ............... 34 7.3 MAINTENANCE SECTION . ........ ........ 34 7.3.1 Maintenance Superintendent ... ......... 34 7.3.2 Maintenance Staff Engineer ... . ... ..... 34 7.3.3 Hechancial Supervisor . . . . . . ........ . 34 7.3.4 Electrical Supervisor . . . . . .. ........ Instruments and Controls Supervisor . . . . . . . . 34 7.3.5 35 7.3.6 Contractor Supervisor . . . . . . . . . . . . . . . 35 7.3.7 Nuclear Master Mechanic . . . . . ......... 35 7.3.8 Nuclear Certified Welder Mechanic . . . . ..... 35 7.3.9 Nuclear Mechanic ........ .... ..... Nuclear Apprentice Mechanic . . . 35 7.J.10 ......... 35 7.3.11 Nuclear Chief Electrican .... .... ..... 36 7.3.12 Nuclear Electricians . ..... . ........ 36 7.3.13 Nuclear Apprentice Electricians . ......... 36 7.3.14 Chief Nuclear Instruments & Controls Technician . . 36 7.3.15 Nuclear Instruments & Controls Technician . . . . . 7.3.16 Assistant Nuclear Instruments & Controls 36 Technician .............. ..... 37 7.3.17 Planning Engineer . . . . . . . . . . . . . . . . . 37 7.3.18 Planning Coordinators . . . . . .......... 37 7.3.19 Maintenance Engineers . . . . . . . . . . . . . . . 37 7.3.20 Materials Coordinator . . . . . . . . . . . . . . . 38 7.3.21 Administrative Technicians ... .... ... .. 38 7.4 OPERATIONS SECTION ........ ......... 38 7.4.1 Operations Superintendent .... .... ..... 38 7.4.2 Operating Engineer ...........'..... 38 7.4.3 shift Supervisors . . . . . . . . . . . . . . . . . 39 7.4.4 Assistant Shift Supt.rvisors . . . . . . . ..... 39 7.4.5 Chief Nuclear Operators . . . . . . . . . . . . . . i 1755 140 Page (ii) AI-200 Date A/9/79 Rev. 13
TABLE OF CONTENTS Sectica (Cont'd) Page 7.4.6 Nuclear Operators . . . . . . ........... 39 7.4.7 Assistant Nuclear Operators . ........... 39 7.4.8 Nuclear Auxiliary Operators . ........... 39 7.4.9 Assistant Nuclear Auxiliary Operators . . . . . . . 40 7.5 ADMINISTRATIVE SECTION 40 7.5.1 Administrative Supervisor . . . . . . . . . . . . . 40 7.5.2 office Manager .......... ........ 40 7.5.3 office Clerks . . . . . . . . . . . ........ 40 7.5.4 officer of the Guard ............... 40 7.5.5 Cuards ...................... 41 7.5.6 Training Supervisor . . . . . . . . ........ 41 7.5.7 Training Records Clerk .............. 41 7.5.8 Training Specialist . ............... 41 7.5.9 Building Servis es Supervisor. . . . . . . . . . . . 41 7.5.10 Administrative Planner. . . . ........... 42 7.5.11 Building Servicemen . ............... 42 7.6 NUCLEAR QA/QC COMPLIANCE SECTION .......... 42 7.6.1 Nuclear QA/QC Compliance Manager .... ..... 42 7.6.2 Nuclear QA/QC Supervisor ............. 42 7.6.3 Nuclear Compliance Supervisor . ........... 43 7.6.4 Nuclear Compliance Auditors . . . . ........ 43 7.6.5 Nuclear QA/QC Inspector . . . . . . . . . . . . . . 43 1755 141 Page (iii) AI-200 Date 8/9/79 Rev. 13
- 1. 0 INTRODUCTION _
The purpose of this section of the Plant Operating Quality Assurance Manual (P0QAM) is to:
- a. Describe the organization and responsibilities of the sections and individuals responsible for all aspects of the operational Quality Assurance effort.
- b. Delineate the qualifications required for positions in CR-3.
1755 142 Page 1 AI-200 Dat e 1/16/76 Rev. 3
2.0 ORGANIZATION The organization as related to operational Quality Assurance is as shown on Technical Specifications Figures 6.2-2, Naclear Plant Organization, and 6.2-1, Offsite organization Chart. 1755 143 Page 2 AI-200 Date 4/13/78 Rev. 4
3.0 jm_ CLEAR GENERAL REVIEW COMMITTEE The Senior Vice President-Engineering and Construction shall appoint a Nuclear General Review Committee having the responsibility for verify- , ing that the operation of the plant is consistent with company policies, rules, approved procedures, and license provisions. This Comittee meets the requirements of ANSI N18.7-1976, " Administrative Controls for Nuclear Power Plants", Section 4. 1755 144 Page 3 AI-200 Date 11/2/78 Rev. 7
4.0 PRODUCTION DEPARTMENT The Director-Power Production is responsible to the Senior Vice President-Engineering and Construction and is given overall authority to staff, operate, and maintain the Nuclear Plant. (Rafer to Technical Specifications Figure 6.2-1, Offsite Organization). The Director-Power Production is informed of significant problems or occt3rrences which relate to safety or Quality Assurance through established administrative procedures. 4.1 ySTD4 PLANT OPERATIONS System plant operations are under the supervision of the Nuclear Plant Manager and the Manager-Fossil Operations. The Nuclear Plant Manager reports to the Director-Power Production. In addition to the expertise in the Nuclear Plant, there is available at the discretion of the Director-Power Productied' assistance from other plants with people thoroughly trained in power plant operations, maintenance, and chemistry, as well as the Production staff. Refer to Technical Specifications Figure 6.2-1, Offsite Organization. 4.2 V STAFF PRODUCTION DEPARTME.r The Production Department is divided into various sections as shown on Technical Specification Figure 6.2-1, Offsite Organization. These sections are staffed with personnel experienced in specialized plant operating and maintenance problems, and provide technical support to the Nuclear Plant staff. 1755 145 Page 4 AI-200 Date 4/13/78 Rev. 4
5.0 Pl. ANT ORGANIZATION The plant organization is shown on Technical Specification Figure 6.2-2, Facility Organization. The Nuclear Plant Manager is direc'cly responsible for the safe operation of the. Nuclear Plant. In all matters pertaining to actual operation and maintenance, and to Technical Specifications, the Nuclear Plant Manager shall report to and be directly responsible to the Director-Power Production. The Nuclear Plant Manager is responsible for the activities of the following: 5.1 OPERATIONS SUPERINTENDENT The Operations Superintendent is responsible to the Nuclear Plant Manager for the operation of the plant. Plant operations are performed by the Operations Section under the general supervisien of the Operations Super-intendent. Under his supervision the Operations Section shall conduct nuclear plant operations to ensure both short and long range safe, efficient, and timely production of electric power consistent with company and department policies and requirements and with governmental regulations. He shall co-ordinate overall plant operating plans and schedules with system requirements and coordinate unit analyses to minimize equipment downtime. The responsi-bilities of the Operations Superintendent include the following:
- a. Develop and Laplement procedures, operating instruc-tions, and emergency procedures required for plant operation.
- b. Support the development of and implement uniform policies, procedures, operating instructions, and emergency plans developed by others for plant operation.
' 1755 146 Date 4/13/78 Page 5 AI-200 Rev. 4
- c. Provide lead responsibility for the development and implementation of new unit pre-operational and start-up procedures.
- d. Provide the overall plant planning, operational scheduling, and coordination required to ensure minimum equipment downtime during unit outages.
- e. Coordinate.overall plant operating plans and schedules with the system dispatching group.
- f. Perform continuing reviews and appraisals to ensure.the safe and efficient condition of planc equipment.
- g. Originate and approve work requests for the maintenance of operating equipment.
- h. Perform operator maintenance.
- 1. Prepare and coordinate need statements for initiating betterment projects for the plant.
- j. Implement requirements of the quality assurance program as related to plant operations.
- k. Implement requirements of plant personnel safety programs.
- 1. Develop and coordinate recommended requirements for training programs related to plant operations with the plant administrative services staff, and support the implementation of operator training and relicensing program.
1755 147 Page 6 AI-200 Date 4/13/78 Rev. 4
- m. Provide operations input to plant manpower, expense, and
, capital budgets.
- n. As required, provide company representation to the local com-munity and represent the needs and interests of the local community to the plant and the Power Production Department.
- o. Represent Florida Power Corporation on and participate in the activities of appropriate industry-related committees, professional societies, and codes and standards groups.
- p. Assuring that plant operations are conducted in accordance with the requirements of the Operating License and the procedures of P0QAM.
- q. (deleted)
- r. Evaluating Radiation Work Permits (RWP's) as required by RP-101, Radiation Protection Manual.
- s. Scheduling and reviewing the results of tests, calibrations and inspections required by SP-300 and 400 series Surveillance Pro-cedures, and determining the status and operability of certain safety-related equipment and systems.
- c. Insuring through proper scheduling that each shift is properly manned for the expected operational activities. Shift manning 1755 148 Page 7 AI-200 Date 4/19/79 Rev. 12
schedules must include compliance with Technical Specifi-cation requirements (Table 6.2-1) and the dictates of good operation practice.
- u. Evaluating shift reports of equipment malfunctions or unusual system behavior to initiate corrective action in the form of maintenance work requests, operational performance tests, special data printout or recording, plant shutdown, or load reduction.
- v. Coordinating alarm tests, fire drills, and evacuation exercises as required by Volume I, Administrative Instructions, and Volume III, Emergency Plan Implementing Procedures, of P0QAM. Scheduling the training and retraining of the shif t personnel. Evaluate the processing of radioactive waste to insuIre that processing, storing, and disposal are being conducted in accordance with established procedures and specific instructions contained in the appropriate Waste Release Permit.
- v. Reviewing and approving the conditions and procedures provided for the performance of operational hydrostatic tests of fluid systems required by a continuing Qurlity Assurance Program.
- x. Participate in Plant Review Committee.
- y. Promote safe work practices in all phases of the operation of the Nuclear Plant.
- z. Participate in Plant Safety Committee. 1755 149 The Operations Superintendent will be in charge of refueling operations under the technical direction of the Performance Engineering Supervisor.
Page 8 AI-200 Date 11/2/78 Rev. 7
The Operationi Superintendent receives and reviews the reports of unit trips, load restrictions, abnormal occurrences, inadvertent liquid or gaseous activity releases, major equipment failure or malfunction (all safeguards), unerplained reactivity changes, loss of off-site power, and deviations from established procedure. He reviews these reports and initiates corrective action and/or additional reports as required. In addition to prescribing and following up remedial action for the fore-going situations, he reports them to the Technical Services Superintendent. The Operations Superintendent prescribes and monitors the training of candidates for shift positions. He also prescribes and monitors the - refresher training of qualified shif t personnel. The Operations Superintendent reviews and approves the rod worth and boron worth curves developed by the Performance Engineering Supervisor for use by Shift Operations personnel. 5.2 OPERATING ENGINEER The Operating Engineer reports to the Operations Superintendent and has a dual function. He provides technical guidance for Nuclear Plant opera-tions, including coordination of operating plans and schedules with system requirements. He also plans and coordinates unit outages. This latter function is a major responsibility. In coordinating unit outages, the Operating Engineer functions in a man-ner similar to that of a Project Engineer. He coordinates the functions of Operations, ChemRad, Technical Support, Outage Planning, Maintenance, and outside consultants. 1755 150 Page 9 AI-200 Date 4/13/78 Rev. 4
5.3 MAINTENANCE SUPERINTENDE _NT The Maintenance Superintendent is responsible to the Nuclear Plant Manager for plant mechanical, instrument, and electrical maintenance. This includes the maintenance of Nuclear Plant equipment, instrumentation, and facilities to ensure both short and long-range safe, efficient, and timely production of electric power consistent with company and department policies and requirements and with Lovernmental regulations. Determine and specify an adequate plant inventory of spara parts, maintenance tools, and operating supplies. Incorporate approved minor plant modifications and additions and support the implementation of plant betterment and other special pro-jects. The responsibilities of the Maintenance Superintendent include the following functions.
- a. Develop and implement unique procedures and methods required for plant maintenance,
- b. Support the development of and implement uniform policies, procedures, and methods approved for plant maintenance.
- c. Support the Power Production Department Maintenance Services Group in the development of a system-vide preventative main-tenance program and develop and implement the necessary pro-cedures and schedules involved for the Nuclear Plant.
- d. Plan and schedule plant maintenance activities and determine and coordinate the use of outside maintenance services.
- a. Initiate, support, and coordinate the development of engineering solutions to maintenance problems with the technical services and other functions.
- f. Supervision of the organization, planning and scheduling of all plant instrument, electrical, and mechanical maintenance.
1755 151 - Page 10 AI-200 Date 4/13/78 Rev. 4
- g. Coordination of the electrical, instrument, and mechani* cal activities of all departments within (and without) the plant in the total maintenance effort.
- h. Organize and conduct the Preventative Maintenance Program on all equipment within the scope of the Maintenance Section's responsibility.
- 1. Assure that all maintenance activities meet or exceed all applicable codes, specifications, standards, FSAR, and Tech-nical Specifications. The requirements include documentation of activities in the performance of modifications, repairs, or replacement of related components and systems.
J. Responsible for safety in all phases of the maintenance effort. Participate in the Plant Review Committee.
- k. Monitor the inventory of spare parts and maintenance supplies to insure min 4= = investments consistent with reliable main-tenance and quality control requirements.
- 1. Compare and updata maintenance budget forecasts.
- m. Review related codes, specifications, and standards for latest revisions or addenda applicable to the operating plant.
- n. Planning scheduled outages.and preparing unit outage reports.
- o. Promote safe work practices.
- p. Participate in Planc Safety Committee.
- q. Organize, develop, and conduct training programs for plant maintenance personnel.
1755 152 Page 11 AI-200 Date 8/9/79 Rev. 13
r. Organize and dir. set the maintenance effort during activities related to initian, and subsequent refuelings. s. Responaible for applicable Maintenance, Refueling, and Sur-veillance Procedures. 5.4 TECHNICAL SERVICES SUPERINTENDENT The Technical 'iervices Superintendent is responsible to the Nuclear Plant Manager for planning, scheduling and supervising the activities o unique t nuclear plant operations to ensure safe, efficient, economical and timely production of electric power consistent with company and department policies , and objectives and governmental regulations. In addition he is responsible for plant support services related to plant performance analysis , environ-ment, and engineering to help meet overall plant production, availability , economics, and efficiency objectives. The Technical Services Superintendent is also responsible for planning for plant emergencies relating u c to p bli health and safety, and damage control as well as plant improve
. u es.u c st di The following functions are included in the Technical ServicesnSuperinte ent d
responsibilities: health physics, power plant chemistry, compliance engineering, performance testing, environmental data acquisition and analysis, systems analysis, engineering support services, plent nimproveme s, t Regulatory Agency liaison, plant accident and other emergency planning and plant records maintenance. 5.5 _ NUCLEAR QA/QC COMPLIANCE MANAGER The NQA/QC Compliance Manager shall be responsibleant to the Nuclear Pl Manager to provide assurance that operation and maintenance e plant of th 1755 153 Page 12 AI-200 Date 8/9/79 Rev. 13
is performed in compliance with P0QAM, Technical Specifications, and applicable portions of the Code of Federal Regulations. The responsi-bilities of the NQA/QC Compliance Manager include the following:
- a. Maintaining an audit system to verify the existence and location of all documentation required by the above documents.
- b. F-=aining the content and acceptance criteria of each required document.
- c. Reviewing each plant-generated Purchase Requisition to verify the safety /non-safety related classification, the Quality and documentation requirements, and to assure that the require-ments for design documents have been specified. Those Purchase Requisitions that require a determination of Quality require-ments or additional Quality requirements will be forwarded to the Director of Production Engineering in accordance with the provisions of CP-101, Procurement of Material, Equipment, and Services.
- d. Directing and approving the preparation, review, and implemen-tation of Quality Control procedures and NQA/QC inspection activities.
- a. Reviewing and approving Quality Control procedures and activities necessary for any plant modification.
- f. Promoting safe work practices and participating in Plant Safety Committee.
1755 154 Page 13 AI-200 Date 8/9/79 Rev. 13
The NQA/QC Compliance Manaeer is responsible for accompanying and assist-ing NRC Directorate of Regulatory Operations Inspectors during routine and special plant inspections as directed by the Technical Services Superintendent. The NQA/QC Compliance Manager will prepara reports, for use by the Nuclear Plant Manager, summarizing the results of his audits and inspec-tions. Where applicable, he vill include suggestions for upgrading the Quality Program or its implementation. The NQA/QC Compliance Manager will assist the Nuclear General Review Com-mittee with their audits of the plant's operational Quality Assurance , program. The NQA/QC Compliance Manager will work with the plant staff in contract negotiations, cost analysis, and planning of required primary system inservice inspections. The NQA/QC Compliance Manager vill observe performance of the Maintenance, Operations, Performance, ChemRad, and Technical Support Departments' field projects to epot check conformance with established procedures and regu-lations. The NQA/QC Compliance Manager will assist the Section Engineers in the areas of planning scheduled outages, post-maintenance testing, and vendor shop inspections. 5.6 CHEMISTRY AND RADIATION PROTECTION ENGINEER The Chemistry and Radiation Protection (ChemRad) Engineer is responsible to the Technical Services Superintendent for matters pertaining to radiation 1755 155 Page 14 AI-200 Date 8/9/79 Rev. 13
protection of all plant site personnel and for water chemistry of all systems within the confines of the Crystal River Unit 3 boundary. These services are performed by the ChemRad Section, which is comprised of the ChemRad Protection Engineer, ChemRad Plant Engineer, Rad Waste Supervisor, Health Physics Supervisor, Assistant ChemRad Protection Engineer, and ChemRad Technicians. The ChemRad Protection Engineer shall be responsible for and have administrative control of the following:
- a. Issuance of Standing Radiation Work Permits to allow routine work to be done in Radiation Controlled Areas (RCA's).
- b. Performing radiological assessments and approving RWP's for non-routine work to be done in RCA's.
- c. Performing routine and special radiation surveys throughout the plant to assure all direct radiation, contamination, and airborne activity areas are properly posted and/or under con-trol, and in compliance with RWP's.
- d. Specifying corrective actic or protective measures to be adopted by plant personnel in order to minimize radiation exposures to personnel,
- e. Assisting in the planning of special operating or maintenance work to be done to minimize radiation exposures to personnel doing the work, and to minimize the amount of airborne activity released to the environment.
i
. 1755 156 Page 15 AI-200 Date 4/13/78 Rev. 4
- f. Maintaining a personnel dosimetry program, such as TLD's, in accordance with 10 CTR 20, on all personnel, and to maintain up-to-date records of all personnel included on the dosimetry program.
- g. Training of all personnel working at the plant in Radiological Control Procedures (RP-100 Series of Volume VIII of P0QAM).
- h. Approving and stipulating the conditions for the release of airborne radioactive materials from Unit 3, and maintaining an up-to-date log on all intermittent and continuous releases in accordance with applicable State and Federal regulations, and Technical Specifications.
- 1. Approving and stipulating the conditions for the release of all radioactive liquids from Unit 3, and maintaining an up-to-date log of all releases in accordance with applicable State and Federal regulations, and Technical Specifications.
J. Maintaining a record of all radioactive materials shipped from the plant for disposal, and assuring that all shipments are in conformance with applicable NRC, DOT, and State regulations.
- k. Sampling and analyzing reactor coolant, auxiliary, and secondary plant water systems on a routine or special basis for chemical analysis, and assuring that all are within the limits of Technical Specifications or other administrative limits. On the basis of results, take correctiv e action to bring analyses within specified limits.
1755 157 AI-200 Date 4/13/78 Page 16 Rev. 4
- 1. Froviding special analysis, such as boron concentration, as requested by the Operations Section or others.
- m. Providing Emergency Plan training to plant personnel, scheduling training drills with outside agencies, and evaluating the results of drills as defined in the Emergency Plan.
- n. Maintaining an on-site radiological surveillance program (TLD's and air sampling) to assure dose rates and airborne concentra-tions are below the 10 CFR 20 criteria for " Unrestricted Areas".
- o. Maintaining an adequate supply of radiation protection instrc-mentation and protective clothing, as well as respiratory and decontamination supplies for use by all plant personnel.
- p. Responsible for providing qualified personnel to monitor and advise work personnel doing operations or maintenance work functions, to assure compliance with work permits and/or administrative or regulatory requirements, and to take necessary action when requirements are in non-compliance,
- q. Responsible for maintaining equipment in the First Aid Room.
- r. Promote safe work practices in all phases of ChemRad area and participate in Plant Safety Committee,
- s. Participate in Plant Review Committee.
The ChemRad Protection Engineer is a member of the Emergency Plan Organi-sation. He may fulfill the duties of " Radiation Team Leader" and act as 1755 158 Page 17 AI-200 Date 4/13/78
- Rev. 4
temporary " Emergency Coordinator" af ter receipt of c Reactor Operator or Senior Reactor Operator's License. He also participates as a Plant Review Committee member. The ChemRad Engineer will evaluate and act upon requests authorizing in-creases in administrative exposure limits as prescribed by RP-101, Radiation Protection Manual. 5.7 PERFORMANCE ENGINEERING SUPERVISOR The Performance Engineering Supervisor is responsible to the Technical Services Superintendent for plant thermal and nuclear performance and maintenance software involving the plant computer. He is responsible for the technical direction of refueling operations and for development of core characteristics, rod worth curves, and boron worth curves for use by Operations personnel af ter review and approval by the Operations Super-intendent. His responsibilities also include any other technical services as required in support of operations and in compliance with safety require-ments. These services are performed by the Performance Engineering Section which is comprised of the Results Engineer, the Computer and Controls Engineer and the Reactor Engineer. The Performance Engineering Supervisor's responsibilities include the following:
- a. Analyze daily logs for evaluation of plant performance and equipment availability.
- b. Supervise and direct work of the Results Engineer, the Computer and Controls Engineer and the Reactor Engineer.
1755 159 . Page 18 AI-200 Date 11/2/78 Rev. 7
- c. Direct activities in emergency support of operation's and coordinate activities with other supervisors.
- d. Direct the Results Engineer in the evaluation of monthly .
plant performance and cost analysis reports.
- e. Participate in the Plant Review Conunittee.
- f. Calculate monthly budget and update yearly forecast.
- g. Promote safe work practices in all phases of the performance engineering effort and participate in Plant Safety Committee.
- h. Direct Rasults Engineer in writing and maintaining Nuclear Plant Reliability Data (NPRD) Reports.
- 1. Evaluate performance of Performance Engineering personnel to insure optimum operating efficiency.
- j. Direct cost analysis studies to determine the effect of plant performance on production costs and make recommendations for corrections to any operating deficiencies.
- k. Organize, develop, and direct training programs for Perfor-mance Engineering personnel.
- 1. Provide information to update training manuals and partici-pate in plant training activities.
- m. Work closely with other supervisors in conducting fuel load-ing and unloading operations, especially in the areas of core nuclear instrumentation.
I755 160 Page 19 AI-200 Date 3/8/79 Rev. 11
- n. Responsible for such research and development projects as may be assigned with a view toward measuring their effect on plant efficiency.
- o. Maintaining complete core history and fuel inventories.
- p. Developing operational follow-up techniques to ensure ade-quate capability to recover from off-design conditions.
5.8 TECHNICAL SUPPORT ENGINEER The Technical Support Engineer is responsible to the Technical Services Superintendent and shall provide technical services as required in sup-port of operations in compliance with safety requirements. The Technical Support Section is comprised of the Inservice Inspection Engineer, the Nuclear Technical Specifications Coordinator, and Plant Engineers. The specific responsibilities of the Technical Support Engineer include the following:
- a. Provide central engineering support for total plant opera-tions in compliance with applicable codes, Regulatory Guides, FSAR, and Technical Specification requirements,
- b. Direct activities of the Inservice Inspection Engineer in planning, scheduling, and performance of the plant's Inser-vice Inspection Program.
- c. Direct activities of the Nuclear Technical Specifications Coordinator in maintaining program which controls plant's compliance with Technical Specifications and FSAR.
- d. Assuring all Nonconforming Operation Reports (NCOR's),
Modification Approval Records (MAR's), and procedure changes comply with Technical Specifications, Regulatory Guides, and FSAR safety considerations.
~ 1755 161 Page 20 AI-200 Date 4/19/79 Rev. 12
- e. Participate on Plant Review Committee (PRC).
- f. Administrate Plant Fire Protection Program and participate in same as Fire Brigade Chief.
- g. Organize, develop, and direct training programs for Technical Support personnel.
- h. Control monthly budget and update yearly forecasts.
- i. Promote safe work practices in all phases of the Technical Support effort and participate in Plant Safety Committee.
J. Evaluate and order additional test equipment, spare parts, or plant equipment as needed, assuring the Quality Program requirements are maintained.
- k. Schedule and conduct the Technical Support Section's outage ,
activities and surveillance requirements as required. ,
- 1. Complete special EEI, EPRI, and monthly operating reports to maintain various operating permits.
- m. Assure compliance with retraining requirements to maintain -
Senior Reactor Operator's License. l e 5.9 ADMINISTRATIVE SUPERVISOR The Administrative Supervisor is responsible to the Plant Manager for site security, training, clerical and accounting functions, drawing and document control, and building services. These functions are performed by four groups, that is, security guard force, clerical operations, training 1755 162 Page 21 AI-200 Date 9/21/78 Rev. 6
t and building services. These four groups are under the general supervi-sion of the Office Manager, the Officer of the Guard, Training Supervisor, and Building Services Supervisor. These individuals are responsible to the Administrative Supervisor for each of their respective areas. The responsibilities of the Administrative Supervisor include the following:
- a. Plant Administration The Office Manager is responsible to the Administrative Supervisor to:
- Formulate, initiate, clear, and administer policy and proce-dures to assure availability, accuracy, and completeness of Quality Assurance records, plant maintenance records, hegith physics records, and all other records as required by commit-ment in the FSAR, ANSI N18.7-1976, or Regulatory Guides.
- Assist in the preparation of reports and permits. (This includes company reports, OSHA reports, NRC reports and permits related to operator licenses, by-products, and special nuclear materials.)
- Coordinate and assist in the preparation of plant operating, construction, and Responsibility Reporting Budgets.
- Coordinate, direct, and administer the clerical services required for recording, transcribing, typing, photocopying, and processing mail.
- Administer the clerical functions and liaison necessary to intra-company relations, for example, the Personnel Depart-ment, Payroll Department, Production Department, Controllers Department, Purchasing Department, etc.
I755 163 Page 22 AI-200 Date 11/2/78 Rev. 7
- Plan, direct, and arrange plant tours and other public relations functions of the plant.
- Coordinate training facilities, equipment, and requirements.
- Act as "Dnergency Liaison Officer" during emergenciea such as fire, bodily injury, radiation and contamination acci-dents, natural disasters, and a reactor accident.
- Arrange timely renewal of operator licenses.
- Must be thoroughly familiar with and make recommendations regarding company policy and regulatory requirements as they apply to records and personnel.
- This section must establish and resolve problems associated in document and drawing control to provide for revisions and accountability of master files and other files in use at the plant.
- b. Plant Security The Officer of the Guard is responsible to the Administrative Supervisor to:
- Develop, evaluate, implement, direct, review, and control an industrial security plan for the Crystal River site. (This plan must include a written, overall description of the secu-rity program designed to protect the Nuclear Plant and written instructions and/or procedures for bomb or other overt threats, civil disturbances, security training, control of incoming packages and materials, response to security alarm systems, etc.). . 1755 164 Page 23 AI-200 Date 9/21/78 Rev. 6
- Coordinate with civil authorities regarding site security.
- Audit use of fuel storage area key and direct the proper handling of documents related to fuel receipt and shipment.
- Must determine requirements and provide solutions to NRC industrial security regulations. This includes writing the Security Plan and Security Procedures.
- Solve the problems associated with entry, key control, alarm systems, security lighting, and fence maintenance.
- Establish guardposts and schedule so guards are always available where needed.
- c. Building and Grounds Maintenance The Building Services Supervisor is responsible to the Admini-strative Supervisor to:
- Plan, develop, coordinate, and direct the administration of building and grounds maintenance and the janitorial services in the sense of housekeeping.
- Must determine requirements and resolve problems associated with manpower and scheduling to provide for custodial ser-vices, insect and rodent control, refuse collection and disposal, preventive maintenance, and fire protection and safety.
- Provide radiation contamination cleanup and laundry services, as required by the ChemRad Section.
1755 165 Page 24 Al-200 Date 4/13/78 Rev. 4
d.. P_,lant Fire and Safety
-- Assist in the formulation and administering of all plant fire and safety regulations. Also, review working con-ditions and recommend changes when hazards exist as per Florida Power Corporation Accident Prevention Manual.
Review industrial and vehicular accidents within and in-cluding damage to the perimeter fence, determine their cause, and initiate and forward accident reports. Participate in Plant Safety Committee. As required, assist the ChemRad Section in the initiation, coordination, and administration of radiation exposure con-trols, by means of film badges and dosimeters, of all plant staff employees, members of cours, and visitors according to NRC, Federal, corporation or other regulations.
- e. Training The Training Coordinator is responsible to the Administrative Supervisor for the planning, scheduling, coordinating, and development of the training, retraining, and replacement training of licensed operators, Electrical and Mechanical Maintenance Repairmen Technical Support Technicians, ChemRad Technicians, and Plant Engineers. He is also responsible for general employee training in the Emergency Plan, security, and radiation protection and for training and requalification
~
documentation. It is desirable that he hold a Senior Reactor Operator's License. His responsibilities include the following: 1755 166 Page 25 AI-200 Date 11/2/78 Rev. 7
Training Programs Develops and schedules training programs to develop and maintain an organization qualified to be responsible for operation, maintenance, and technical aspects of the plant.
- Maintains all training records and documents.
-- Evaluates licensed operators as required by 10 CFR 55 Appendix A, using written tests, oral tests, and operating demonstrations.
Safety
-- Responsible for all emergency training and drills. Special emphasis is given to the safe practices applicable to accidents which might conceivably result in radioactive material release in and beyond the site boundary.
Keep informed of the latest ideas in plant safety equipment. Train power plant employees in safe work practices.
-- Participate in Plant Safety Committee.
Tours Responsible for conducting and/or coordinating all plant tours. Outages
-- Assist as outage supervisor supplying licensed Senior Reactor Operator supervision, if qualified, in the reactor building, spent fuel building, and auxiliary building as required.
1755 167 Page 26 AI-200 Date 4/13/78 Rev. 4
5.0 P_LANT REVIEW COMMITTEE , The Plant Review Comunittee is appointed by the Plant Manager. This comunittee shall provide an on-site review to the Plant Manager in matters of nuclear safety, radiation exposure, and review and audit of plant operation, maintenance, and technical matters. This consnittee , meets the requirements of ANSI N18.7-1976, " Administrative Controls for Nuclear Power Plants", Section 4.5, Refer to AI-300, Plant Review Com-mittee Charter. 1755 168 Page 27 AI-200 Date 11/2/78 Rev. 7
7.0 REQUIRED STAFF QUALIFICATIONS BY POSITION This section describes 'the minimum qualifications of education, skill, and experience required for each Nuclear Plant staff position. These require-ments are consistent with ANSI 18.1-1971, paragraphs 4.2 thru 4.6. Please refer to Technical Specifications Figure 6.2-2, Facility Organization. Each supervisor shall be responsible for assuring .that the employees reporting to hLa meet the mintmum qualifications for the position he is filling. 7.1 Nuclear Plant Manager The Nuclear Plant Manager shall have 15 years of responsible power plant experience of which a minimum of three years shall be nuclear power $ ant experience. The Nuclear Plant Manager shall have acquired the experience and training normally required for examination by the NRC for a Senior Reactor Operator's License whether or not the examination is taken. The Nuclear Plant Manager shall have a recognized Bachelor's Degree or higher degree in an engineering or scientific field generally associated with power pro-duction. 7.2 Technical Services Section 7.2.1 Technical Services Superintendent The Technical Services Superintendent shall have 10 yearc of responsible power plant experience of which a minimum of three years shall be nuclear power plant experience. The Technical Services Superintendent shall have acquired the experience and training normally requirer for examination by the NRC for a Senior Reactor Operator's Licensa Ohp.Xct or not the exami-nation is taken. The Technical Services Supers.arenoets shall have a 1755 169 AI-200
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recognized Bachelor's hegree or higher degree in an engineering or scientific field generally associated with power production. 7.2.2 Performance Engineering 7.2.2.1 Performance Engineering Supervisor The Performance Engineering Supervisor shall have a Bachelor's Degree in an engineering field normally related to power plant work and a minimum of eight years power plant experience. He should have advanced training in Nuclear Engineering, including studies in core analysis. 7.2.2.2 Rasults Engineer The Results Engineer shall have a Bachelor's Degree in an engineering field normally related to power plant work and a minimum cf five year's power plant experience. 7.2.2.3 Computer and Controls Engineer The Computer and Controls Engineer shall have an advanced knowledge of Computer programing, nuclear and power plant instrumentation, principles of nuclear power plant operation, and a minimum of five years power plant experience. He shall have a Bachelor's Degree in Engineering. 7.2.2.4 Reactor Engineer The Reactor Engineer shall have a Bachelor's Degree with advanced training in Nuclear Engineering, including studies in core analysis. He shall have five years of related experience, including two years of Nuclear Plant experience.
' 1755 170 Page 29 AI-200 Date 8/9/79 (next page 31) Rev. 13
7.2.3 Chemistry & Radiation Protection 7.2,3.1 ChemRad Protection Engineer The ChemRad Protection Engineec shall cumulatively have a minimum of five years professional experience in radiation protection. A minimum of three of the five years shall be in a nuclear facility dealing with similar prob-lems as experienced in this station. This position requires a Bachelor's Degree in engineering, chemistry, or physics, or equivalent acceptable experience. 7.2.3.2 Assistant ChemRad Protection Engineer The Assistant ChemRad Protection Engineer shall have a minimum of five years experience in chemistry, radio-chemistry, and radiation protection. A minimum of two years of this five years experience will be related technical training. These positions require a Bachelor's Degree in engineering, chemistry, or physics, or equivalent acceptable experience. 7.2.3.3 Plant Engineer (ChemRad) The Plant Engineer shall hav'e a Bachelor's Degree or equivalent in an engineering or scientific field generally associated with power pro-duction. 7.2.3.4 Radiation Waste Supervisor The Radiation Waste Supervisor shall l ave a minimum of five years experi-ence in radiation protection at a nucler: facility. A minimum of two years of this five years experience should be related technical training. A minimum of four years of this five years experience may be fulfilled by related technical or academic training.
- 1755 171 Page 31 AI-200 Date 8/9/79 Rev. 13
7.2.3.5 Health Physics Supervisor The Health Physics Supervisor shall have a minimum of five years experi- . ence in radiation protection at a nuclear facility. A minimum of two years of this five years experience should be related technical training. A marimum of four years of this five years experience may be fulfilled by related technical or academic training. 7.2.3.6 Chief ChemRad Technician The Chief ChemRad Technician shall have three years of working experience in chemistry and radiation protection, one year of which shall be in a nuclear plant, and must meet the top qualifications of ChemRad Technician. 7.2.3.7 ChemRad Technician The ChemRad Technician shall have two years of working experience involving chemistry, radio-chemistry, and radiation protection; must hava one year of related technical training in chemistry and radio-chemistry or radiation protection; and must meet the top qualifications of Assistant ChemRad Technician. 7.2.3.8 Assistant ChemRad Technician The Assistant ChemRad Technician shall pass the FPC Equivalency Exams in physics, chemistry, and math with a minimum score of 70 in each phase. The Assistant ChemRad Technician will also be screened by a professional consultant to determine his capability for learning basic skills in chemistry, radiochemistry, and radiation protection operations. 1755 172 Page 32 AI-200 Date 8/9/79 Rev. 13
1 7.2.4 Technical Support g 7.2.4.1 Technical Support Engineer The Technical Support Engineer shall have a minimum of eight years of responsible power plant experience of which one year shall be nuclear power plant experience. A marimum of four years of the remaining seven years of experience can be fulfilled by satisfactory completion of academic training. He shall have a Bachelor's Degree in engineering.
'i . 2. 4. 2 Inservice Inspection Engineer The Inservice Inspection Engineer shall have a Bachelor's Degree in an engineering field normally related to power plant work and a minimum of five years power plant experience. He should be knowledgeable of Quality Control practices and Quality Assurance and Surveillance as required by ASME Section II.
7.2.4.3 Nuclear Technical Specifications Coordinator The Nuclear Technical Specifications Coordinator should have a Bachelor's Degree in an engineering field normally related to power plant work and a minimum of five years power plant experience. He should be knowledgeable of Technical Specifications, FSAR, Facility License, and Code of Federal Regulations requirements and reporting functions. 7.2.4.4 Plant Engineer The Plant Engineer shall have a Bachelor's Degree or higher degree in an engineering or scientific field generally associated with power production, or a minimum of six years of engineering or related work experience of which three years shall be nuclear plant experience. 1755 173 7.2.4.5 Engineering Assistant The Engineering Assistant shall have a minimum of an Associates Degree or i equivalent in an engineering or scientific field generally associated with power production. Page 33 AI-200 Date 8/9/79 Rev. 13
7.3, Maintenance Section 7.3.1 Maintenance Superintendent The Maintenance Superintendent sha11 have a minimum of ten years responsible power plant experience or applicable industrial experience, a minimum of four years of which shall be nuclear power plant experience. He shall have non-destructive testing familiarity, craft knowledge, and in understanding of electrical, pressure vessel, and piping codes. He shall have a Bachelor's Degree in engineering. 7.3.2 Maintenance Staff Engineer The Maintenance Staff Engineer shall have a minimum of seven years of responsible power plant experience or applicable industrial experience, a minimum of one year of which shall be nuclear power plant experience. He shall have non-destructive testing familiarity, craft knowledge, and an understanding of electrical, pressure vessel, and piping codes. He shall have a Bachelor's Degree in engineering. 7.3.3 Mechanical Supervisor The Mechanical Supervisor shall have a high school diploma or equivalent and a minimum of four years experience in mechanical maintenance. 7.3.4 Electrical Supervisor The Electrical Supervisor shall have a high school diploma or equivalent and a minimum of four years experience in electrical maintenance. 7.3.5 Instruments and Controls Supervisor The Instruments and Controls Supervisor shall have a minimum of five years experience in instrumentation and control, of which a minimum of six months 1755 !74 Page 34 AI-200 Date 4/13/78
. Rev. 4
shall be in nuclear instrumentation and control. A minimum of two years of this five years experience should be related technical training. A maximum of four years of this five years experience may be fulfilled by related technical or academic training. 7.3.6 Contractor Supervisor The Contractor Supervisor shall have a college diploma or equivalent and a minimum of four years of experience in the supervision of contractors. 7.3.7 Nuclear Master Mechanic This position must have four years of working experience in related me-chanical systems, one year of which shall be in a Nuclear Plant, and must meet the top qualifications of Nuclear Certified Welder Mechanic. 7.3.8 Nuclear Certified Welder Mechanic This position must meet the top qualificaticus of Nuclear Mechanic and must be certified to FP-81 welding procedure. 7.3.9 Nuclear Mechanic This position must have a minimum of three years of related mechanical caperi-ence; have certification of successful completion of a course in mechanical maintenance; and meet the top qualifications of Nuclear Apprentice Mechanic. 7.3.10 Nuclear Apprentice Mechanic This position must have certification of successful completion of a course in basic shop fundamentals and will be screened by a professional consultant to determine capability for learning basic skills in mechanical maintenance. 7.3.11 Nuclear Chief Electrician This position must have four years of working experience in related
.7 !7h Page 35 AI-200 Date 4/13/78 Rev. 4
electrical systems, one year of which shall be in a Nuclear Plant, and meet the top qualifications of Nuclear Electrician. 7.3.12 Nuclear Electrician , This position must have three years working experience in related electrical systems; have certification of successful completion of a course in electri-cal theory; and meet the top qualifications of Nuclear Apprentice Electrician. 7.3.13 Nuclear Apprentice Electrician _ This position must have certification of successful completion of a course in basic electrical theory and will be screened by a professional consul-tant to determine capability for learning basic skills ..n nuclear electrical maintenance operations. 7.3.14 Chief Nuclear Instruments and Controls Technician This position must have four years of working experience in instrumentation and controls, one year of which shall be in a Nuclear Plant, and must meet the top qualifications of Nuclear Instruments & Controls Technician. 7.3.15 Nuclear Instruments & Controls Technician This position must have three years of working experience in instrumentation and control systems; have one year of related technical training associated with instrumentation and control systems; and meet the top qualifications of Assistant Nuclear Instruments & Controls Technician. 7.3.16 Assistant Nuclear Instruments & Controls Technician This position must pass the FPC Equivalency Exams in physics, chemistry, and math with a =4n4=nm score of 70 in each phase. The Assistant Nuclear t 1755 '76 Page 36 AI-200 Date 9/21/78 Rev. 6
Instruments & Controls Technician will also be screened by a professional consultant to determinc capability for learning basic skills in nuclear instrumentation maintenance operations. 7.3.17 Planning Engineer The Planning Engineer shall have a Bachelor's Degree in an engineering field normally related to power plant work and a minimum of five years power plant experience. 7.3.18 Planning Coordinator The Planning Coordinator shall have a high school diploma plus two years technical school or equivalent. He shall also have five years of power plant operation / maintenance planning of which one of these five years shall have been nuclear power.planc experience and three years shall have been maintenance or maintenance planning experience. 7.3.19 Maintenance Engineer The Maintenance Engineer shall have a Bachelor's Degree in an engineering field normally related to power plant work and a minimum of five years power plant experience. 7.3.20 Materials Coordinator The Materials Coordinator shall have a high school diploma plus two years technical school or equivalent. He shall also have five years of similar experience in maintenance or materials control. 1755 177 i Page 37 AI-200 Date 8/9/79 Rev. 13
7.3.21 Administrative Technician This position requires various general office and clerical skills consistent with FPC personnel policies. 7.4 Operations Section 7.4.1 Operations Superintendent The Operations Sueprintendent shall have a minimum of ten years of responsible power plant experience of which a minimum of five years shall. be nuclear power plant experience. The Operations Superintendent shall hold a Senior Reactor Operator's License. 7.4.2 Operating Engineer The Operating Engineer shall have a minimum of eight years of responsible power planc experience of which a minimum of three years shall be nuclear power plant experience. The Operating Engineer shall hold a Senior Reactor Operator's License. 7.4.3 Shift Supervisors The Shift Supervisor shall have a minimum of a high school diploma or equivalent and four years of responsible power plant experience of which a minimum of one year shall be Nuclear Plant experience. The Shift i Supervisor shall hold a Senior Reactors Operator's License. 1755 '78 Page 38 AI-200 Date 8/9/79 Rev. 13
7.4.4 Assistant Shift Supervisor The Assistant Shift Supervisor shall have a minimum of a high school diploma or equivalent and four years of responsible power plant experi-ence of which a minimum of one year shall be Nuclear Plant experience. The Assistant Shift Supervisor shall hold a Senior Reactor Operator's License. 7.4.5 Chief Nuclear Operator This positio'n must hold an NCR Reactor Operator's License for CR-3; have a minimum of three years power plant experience of which one year must be Nuclear Plant experience; and meet the top qualifications of Nuclear Operator. 7.4.6 Nuclear Operator This position must hold an NRC Reactor Operator's License for CR-3 and meet the top qualifications of Assistant Nuclear Operator. 7.4.7 Assistant Nuclear Operator This position must hold an NRC Reactor Operator's License for CR-3; have a minimum of two years power plant experience of which one year must be Nuclear Plant experience; and have the top qualifications of a Nuclear Auxiliary Operator. 7.4.8 Nuclear Auxiliary Operator This position must be qualified as an Auxiliary Operator or higher at a fossil or nuclear power plant with one year power plant operating experi-ence and must meet the top qualifications of Assistant Nuclear Auxiliary Operator. 1755 179 Page 39 AI-200 Date 4/13/78 Rev. 4
7.4.9 Assistant Nuclear Auxiliarv Operator This position will be screened by a professional consultant to determine . capability for learning basic requirements of nuclear operations and ~ progressing to eventual NRC Licensing. The Assistant Nuclear Auxiliary Operator shall have a high shcool diploma or equivalent. 7.5 Administrative Section 7.5.1 Administrative Supervisor The Administrative Supervisor shall be a high school graduate with a minimum of five years experience in the planning, scheduling, and coordi-nation of clerical functions, building services, and security systems. 7.5.2 office Manager
'ihe Office Manager shall be a high school graduate with a minimum of five years experience in the planning, scheduling, and coordination of clerical functions, building services, and security systemt 7.5.3 office clerks These positions depending upon the classification, require skills developed through formal training, experience, or a combination of the two in general clerical activities such as: filing and retrieval, records processing, exposure to common office machinery and specific skills related to the classification as specified in FPC personnel policies.
7.5.4 Officer of the Guard Security personnel shall have qualifications as stated in Section 7.0 (Security Guard Force) of the Crystal River Nuclear Plant Security Plan and which are consistent with ANSI N18.7-1976, Section 4 (Administrative Controls). Page 40 AI-200 Date 11/2/78 Rev. 7
7.5.5 Guards Security personnel shall have qualifications as stated in Section 7.0 (Security Guard Force) of the Crystal River Nuclear Plant Security Plan and which are consistent with ANSI N18.7-1976, Section 4 (Administrative Controls). 7.5.6 Training Supervisor The Training Supervisor shall have a high school diploma or equivalent and advanced courses in nuclear power, power plant theory, and electronics beyond the level required for a Senior Reactor Operator's License. He shall hold, have held, or be in the process of obtaining a Senior Reactor Operator's License. 7.5.7 Training Records Clerk This position requires skills developed through formal training, experi-ence, or a combination of the two in general clerical activities such as: filing and retrieval, records processing, and other such functional skills as specified in FPC personnel policies. 7.5.8 Training Specialist The Training Specialist shall have a minimum of a high school diploma or equivalent and four years of responsible power plant experience of which a minimum of one year shall be nuclear power plant experience. 7.5.9 Building Services Supervisor The Building Services Supervisor shall have a high school diploma or equivalent and a minimum of four years of experience in decontaminating Nuclear Plant, maintaining office, and other service areas in proper order. 175$ i8I Page 41 AI-200 Date 11/2/78 Rev. 7
*7.5.10 Administrative Planner This position requires various general office and clerical skills, including experience in scheduling and dispatching materials and per-sonnel for building service tasks.
7.5.11 Building Servicemen These positions will require the general ability, maturity, and respon-sibility to perform janitorial, yardwork, laundry, and other service tasks for maintaining offices and other service areas in proper order. They will be scr2ened by a professional consultant to determine their capability for learning basic skills. 7.6 NUCLEAR QA/QC CCNPLIANCE SECTION 7.6.1 Nuclear QA/QC Compliance Manager The NQA/QC Compliance Manager shall have an extensive knowledge of OA practices and regulatory requirements, including Codes of Federal Regu-lations, FPC Quality Programs, CR-3 FSAR, Technical Specifications, and applicable Codes and Standards. Although a Bachelor's Engineering Degree is desirable, it should not be a prerequisite as experience should be able to preclude the requirement. As a minimum, the requirements should be a B. A. and five years of QA-related experience or no degree and 10 yrs. QA-related experience. 7.6.2 Nuclear QA/QC Supervisor The NQA/QC Supervisor shall have an extensive knowledge of QA practices and regulatory requirements, including Codes of Federal Regulations, FPC Quality Programs, CR-3 FSAR, Technical Specifications, and applicable Codes and Standards. Although a Bachelor's Engineering Degree is desir-1755 182 Page 42 AI-200 Date 8/9/79
. Rev 13
e able, it should not be a prerequisite as experience should be able to k preclude the requirement. As a minimum, the requirements should be a B. A. and five years of QA-related experience or no degree and 10 yrs. QA-related experience. 7.6.3 Nuclear Compliance Supervisor The Nuclear Compliance Supervisor shall have an extensive knowledge of QA practices and regulatory requirements, including Codes of Federal Regulations, FPC Quality Programs, CR-3 FSAR, Technical Specifications, and applicable Codes and Standards. Although a Bachelor's Engineering Degree is desirable, it should not be a prerequisite as experience should be able to preclude the requirement. As a minimum .the requirements should be a B. A. and five years of QA-related experience or no degree and 10 yrs. QA-related experience. I, L 7.6.4 Nuclear Compliance Auditors ; l The Nuclear Compliance Auditors shall have a minimum of 4 yrs. of engi- . I neering or related work experience, at least 2 yrs. of which were related { to Compliance or QA work on one or more of the following areas: design, construction, testing, maintenance, or system operation. They should have an Associate's Degree in an engineering field, or equivalent. 7.6.5 Nuclear QA/QC Inspector The NQA/QC Inspector shall have a high school diploma plus 2 yrs. of technical school or college education. He should be certified, or have training, as a non-destructive examiner in ultrasonic, penetrant, magnetic particle, or eddy current methods. He should be capable of meeting the experience requirements of ANSI N45.2-6 (1973) for a Level q 1755 183 Page 43 AI-200 Date 8/9/79 Rev. 13
II person involved in inspections and tests. The'NOA/QC Inspector must have 4 yrs. of experience in QA testing or inspection, 2 yrs. of which have been in a Nuclear Plant. 1755 184 Page 44 AI-200 Date 8/9/79 Rev. 13
o ATTACHMENT VIH LONG RANGE PLAN FOR UPGRADING THE ON-SITE TECHNICAL SUPPORT CENTER Florida Power Corporation will construct a new facility to act as a Tech-nical Support Center (TSC) during emergency situations. This new facility will be located inside the security perimeter, below the berm, northeast of the plant. The size of the TSC will be sufficient to support a minimum of 25 people and provide roon for all required records, drawings, communications, and instrumentation. The instrumentation within the TSC shall include the capability to monitor current and historical data for the primary and secondary plant parameters which are regarded as necessary for plant accident analysis. In addition, plant radiological parameters and site meteorological data is planned to be displayed in the TSC. The CR-3 plant computer Central Processor Unit is not presently capable of providing independent access fran remote locations, but the outputs can be paralleled and monitored in the TSC. FPC is planning to replace the exist-ing computer Central Processing Unit with a unit capable of providing inde-pendent data access from remote terminals. The present schedule for com-pletion of this modification is Fall of 1981. Reliable communications will be available in the TSC and will include dedi-cated lines between the TSC and the Control Room, between the TSC and the NRC. In addition, commercial telephone lines for on-site and off-site com-munications will be available as well as intra-plant lines which include private, party and paging functions. The power supply for the TSC will be designed to be available at all times, including during a loss of offsite power. Where needed, to avoid loss of stored data due to momentary loss of power or switching transients, the power to the applicable instruments and equipment will be continuous (unin-terruptable). Sufficient records and drawings to aid in accident analysis will be perma-nently stored and available for use in the TSC. The structural design of the TSC will take into account the effects of nat-ural phenomena that may occur at the site, and will be designed to meet or exceed existing codes. Protection from radiological hazards, including direct radiation and air-borne contaminants as per General Design Criterion 19 and SRP 6.4, will be provided for in the design of the TSC. Pennanent monitoring systems will be provided to indicate radiation dose rates and airoorne radioactivity concentrations inside the TSC. At the present time, FPC does not have a system of off-site radiological monitoring systems which can provide data, but instead by utilizing known source release and known meteorological information, the release data can be properly evaluated by overlays and equations. If required, a survey team could be dispatched from the TSC to sample and verify the release data. 755 185}}