ML19210D616

Responds to NRC 791030 Ltr Re short-term Lessons Learned Recommendations from TMI-2 Incident.Items Identified Which Require Outages Have Been Scheduled Concurrent W/Presently Planned Outages

ML19210D616
Person / Time
Site:	Davis Besse
Issue date:	11/21/1979
From:	Roe L TOLEDO EDISON CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
RTR-NUREG-0578, RTR-NUREG-578 559, NUDOCS 7911270397
Download: ML19210D616 (12)

Text

. .

Docket No. 50-346 Totsoo License No. NPF-3 Serial No. 559 LOWELL E. ROE November 21, 1979 7,,7,','$',",*,,,,

[4191 259-5242 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Dear Mr. Denton:

Toledo Edison indicated its intent to comply with your staf f's short term lessons learned recommendations from the ihree Mile Island, Unit 2 (TMI-2) incident of March 28,1979. Our letter of October 23, 1979 (Serial No. 546) identified pre-liminary schedules and some basic information about our implementation of the recommendations at the Davis-Besse Nuclear Power Station Unit 1 (DB-1) . We identi-fled them, and reiterate now, that we consider the established design of the Davis-Besse Station combined with the modifications and procedural changes made in recent months adequate to ensure there be no undue risk to public health and safety.

Your generic letter of October 30,1979 (Log No. 454) identifies concern over implementation schedules of utilities for items recommended to be implemented by January 1, 1980. Additionally, it identifies new, more detailed recommendation criteria. The schedule concern was reinforced during a telephone conversation between our staffs on November 9, 1979. Toledo Edison has been reviewing and further defining our projected activities in this area and have gone back and reassessed all our pending activities in light of your concern. The result has been a prioritization process that delays several non-related ongoing regulatory activities and therefore allows e::paditing of the subject tasks by our personnel as well as our vendors and consultants. We will attempt to minimize the impact on other TMI-2 related items being pursued in the bulletins and orders and emergency preparedness areas.

~

The results on short term lessons learned of our reassessment are provided here.

The items attached only include those schedule items whose projected implementation dates have been revised or justified and those whose implementation methods or de-tailed information modify our letter responses of October 23, 1979.

7911270 3 7 ) l P t THE TOLEDO EDISON COMPANY EDISON PLAZA 300 MADISCN AVENUE TOLEDO. OHIO 43652

}X99 MB

Docket No. 50-346 License No. NPF-3 Serial No. 559 November 21, 1979 Page Two These projected dates are reficceed as realistic and are based on aggressive schedules for Toledo Edison, its vendors and consultants considering all regulatory activities currently defined. As a result, these dates will be impacted by each new higher priority request of your staff. As the activities progress we will continue to expedite and improve schedules when possible. Those items requiring outages have been scheduled concurrent with presently planned outages. Installation may be completed during an earlier outage of sufficient duration af ter the equipment is available and installation engineering complete. As always, we are available, at your request to discuss any aspect of our submittal.

Very truly yours,

}

.( ,  ? [-J l C" LER:TJM 0

}399 O

Docket No. 50-346 License No. NPF-3 Serial No. 559 No*/ ember 21, 1979 TOLEDO EDISON RESPONSE TO NRC LETTER OF OCTOBER 30, 1979 FOR DAVIS-BESSE NUCLEAR POWER STATION UNIT 1 (DB-1) 3x99 050

Docket No. 50-346 License No. NPF-3 Serial No. 559 November 21, 1979 Toledo Edison Response to NRC letter dated October 30, 1979 NOTES: 1) All schedule dates are projections and subject to change based on outcome of reviews under 10CFR50.59, hardware availability and support manpower availability.

2) Items in the attachment correspond to the identification schemes of the September 13 and October 30, 1979 NRC letters and NUREG 0578.

Task number designations are for administrative cross-referencing within Toledo Edison.

Letter Items:

Item d Three additional instrumentation requirements for short-term action were developed during the ACRS review of NUREG 0578. These items relate to containment pressure, containment water level and containment hydrogen monitors designed to follow the course of an accident. (Task Nos. 37 - 39)

Response

CONTAINMENT PRESSURE - Safety grade control room indication for an extended range of containment vessel pressure will be provided at Davis-Besse Unit 1. The revised projected schedule identifies installation by Janu'ary 1,1981.

CONTAINMENT WATER LEVEL - Safety grade control room indication of narrow and wide range containment water level will be provided at Davis-Besse Unit 1.

Several options to provide this level are currently being evaluated. For commitment purposes, the most schedule restrictive option requires instrument installation inside the containment building. The planned outage when the equipment would be available for installation and testing is the Spring 1981 refueling outage. If a design is selected that does not require containment entry for full or. partial implementation, the schedule may be accelerated.

CONTAINMENT HYDROGEN - Safety grade control rcom indication of 0 - 10% hydrogen con-centration will be provided at Davis-Besse Unit 1. The revised projected schedule identifies installation by January 1,1981.

1]) 0 Docket No. 50-346 License No. NPF-3 Serial No. 559 November 21, 1979 Item e An additional requirem. following issuance of NUREG 0578, which concerned a remotely operable high soint vent for gas from the reactor coolant system, was developed (high point and vessel head vents, Task No. 40)

Response

A generic design effort is underway by B&W to provide a functional description of the construction, location, size and appropriate power supply for reactor coolant system high point vents. Appropriate safety analyses and reviews considering the effects of such vents are also being pursued concurrently. Additionally, an evaluation of the necessity of a reactor vessel head vent on the B&W nuclear steam supply system is being performed. It is presently projected that a preliminary design of the pro-posed venting system should be completed and forwarded to you for review and conceptual approval by January, 1980.

Provided the evaluations are completed as expected and no potential unreviewed safety issues are identified, these vents could be installed during the planned Spring outage of 1981. This is contingent upon NRC approval and qualified equipment availability within the time frame between NRC approval and the 1981 outage.

Item f NUREG 0578 recommendations as modified by the errata of Enclosure 5.

(Note that recommendations are numbered consistent with NUREG 0578).

Recommendation 2.1.1 - Emergency Power Supply Requirements for Pressurizer Heaters, Power Operated Relief Valves and Block Valves, and Pressurizer Level Indicators in PWRs. (Task No. 001 - 005).

Response

PRESSURIZER HEATERS - The current Davis-Besse design provides manual loading of 126 kw of pressurizer heater capacity to each emergency diesel. Procedures are already in place covering the manual loading of these electrical loads. All actions to energize these heaters from the emergency diesel generators are in the control room and are made by the control room operator. Davis-Besse currently meets this recommendation.

No further action is required.

PILOT OPERATED RELIEF VALVE (PORV) - The current Davis-Besse design has its non-safety related pilot operated relief solenoid valve and control circuitry powered from a

. 125 vDC bus supplied by a safety grade battery system. A safety grade battery charger is provided to this DC system. The valve position indication is provided from the station battery ria a 250 vDC to 120 vAC inverter. Isolation of safety grade buses from non-safety grade devices is provided by qualified devices at the motor control c ent er s. Davis-Besse meets the emergency power recommendation. No further action is required.

}T)) Jb2

Docket No. 50-346 License No. NPF-3 Serial No. 559 November 21, 1979 PILOT OPERATED RELIEF VALVE BLOCK VALVES - Davis-Besse will provide an AC safety grade motive and control power source to the PORV block valve. The power source will,be appropriately isolated from non-safety grade portions of the system. The current schedule projects this modification will be completed by January 1,1980.

PRESSURIZER LEVEL INDICATORS - The current Davls-Besse design has two channels of safety grade pressurizer level indication on a safety grade power supply. Davis-Besse currently meets this recommendation. No further action is requirsd.

Recommendation 2.1.3a - Direct Indication of Power Operated Relief Valve and Safety Valve Position for PWRs and BWRs. (Task No. 7)

Response

Toledo Edison will provide positive primary system safety "alve and PORU position indication by the use of an acoustic monitoring system. Discussions with vendors supplying these devices indicate that the system can be powered from a safety related power supply with sensors that are seismically and environmentally qualified. However, the entire system will not meet safety grade requirements. Due to projected equipment delivery and installation engineering requirements this system will be installed and tested prior to start-up af ter the Spring refueling outage currently scheduled to start in March 1980. Davis-Besse currently utilizes quench tank level, quench tank pressure and PORV electrical position indication to aid in operator diagnostics. These methods are currently discussed in the DB-1 emergency procedures. Toledo Edison's current procedures and operator training are adequate to provide diagnostic informa-tion for the operator and will be utilized as a backup to the acoustic system when installed.

Recommendation 2.1.3b - Instrumentation for Detection of Inadequate Core Cooling in BWRs and PWRs. (Task Nos. 8 - 10).

Response

Toledo Edison is participating with B&W in the development of procedural guidelines for operator recognition of inadequate core cooling under the conditions identified in Section 3 of recommendation 2.1.9 (NUREG 0578, page A-44) . These guidelines are currently being pursued with the NRC Bulletin and Orders Task Fo ce. Inadequate core cooling for the refueling and loss of inventory cases will be submitted by January 1, 1980.

The design of any new instrumentation, if deemed necessary by a joint B&W Owners evaluation, would be submitted for NRC review and approval in April 1980. A scheduled installation date for any such instrumentation will be provided along with the April 1980 proposal.

DB-1 will install a primary c )olant saturation meter which will provide on-line control room Ladication of coolant saturation condition. The lack of wide range safety grade temperature inputs available is currently under review, as well as the quality assurance program to which the system is built. This will identify if the overall system can qualify as fully safety grade. Due to expected projected equipment delivery and installation engineering requirements the system will be installed during the Spring refueling outage currently snheduled to start in March, 1980. The following pages contain additional information On this subcooled meter.

1x99 053

. , , Docket No. 50-346 License No. NPF-3 Serial No. 559 November 21, 1979 INFORMATION ON THE SUBC00 LING METER Display Infomation Displayed (T-Tsat, Tsat, Press, etc.) AT = (Tsat - Th)

Display Type (Analog, Digital, CRT) . Digital continuous for selected Continuous or on Demand parameter Single or Redundant Display single Location of Display Control Room Area klarms (include setpoints) A T450 F Overall uncertainty (*F, PSI) Note (1)

Range of Display Note (1)

Qualifications (seismic, environmental, IEEE323) Note (1)

Calculator -

Type (process computer, dedicated digital or analog calc.) Dedicated Digital If process computer is used specify availability. (% of time) N/A Single or redundant calculators single Selection Logic (highest T., lowest press) Note (1)

Qualifications (seismic, environmental, IEEE323) Note (1)

Calculational Technique (Steam Tables, Functional Fit, ranges) steam Tables Inout Temperature (RTD's or T/C's) RTDs Temperature (number of sensors and locations) 2 (1/ Hot leg)

Range of temperature sensors Note (1)

. ;299 n64

Docket No. 50-346 .

Licence No. NPF-3 Serial No. 559 November 21, 1979

?

Uncertainty

• of temperature sensors (*F at 1 ) Note n3 Qualifications (seismic, environmental, IEEE323) Note n)

Pressure (specjfy instrument used) Note (1)

Pressure (numoer of sensors and locations) Note (1)

Range of Pressure sensors Note (1)

Uncertainty

• of pressure sensors (PSI at 1 ) Note (1)

Qualifications (seismic, environmental, IEEE323) Note (1)

Backup Capability Availability of Temp & Press yes Availability of Steam Tables etc. Yes Training of operators Yes Procedures Yes

• Uncertainties must address conditions of forcea flow and natural circulation NOTE (1) The selection of specific equipment is still under review.

Docket No. 50-346 License No. NPF-3 Serial No. 559 November 21, 1979 Davis-Besse currently provides operator saturation curves mounted in the control room as well as in the station's emergency procedures. Additionally there is a computer alarm point for approach i.o saturation conditions. Toledo Edison's current procedures and extensive operator training in the area of reactor coolant saturation concerns provide an adequate diagnostic cool to reactor operators and will remain available after the saturation meters are installed.

Recommendation 2.1.5a - Dedicated Penetrations for External Recombiners or Post-Accident Purge Systems (Task No. 13).

Response

Davis-Besse utilizes the concainment hydrogen dilution system for post-accident combustible gas control. This is a dedicated, safety grade, accident mitigation system whose penetrations meet redundancy and single failure criteria. Davis-Besse complies with this recommendation. No further action is required.

Recommendation 2.1.6a - Integrity of Systems outside Containment likely to Contair Radioactive Materials (Engineered Safety Systems and Auxiliary Systems) for PWRs and BWRs (Task No.15) .

Response

A leakage reduction program, including an initial effort as well as an ongoing pre-ventive maintenance effort, is being developed for DB-1. By January 1,1980 we will provide a summary description of this program. This summary will include a list of the systems involved and identification of testing methods. Currently we expect difficulty in the area of gaseous systems testing if helium leak detection methods are utilized. Alternate methods are being evaluated.

Recomendation 2.1.6b - Design Review of Plant Shielding of Spaces for Post-Accident Operations ( Task No. 16).

Response

Toledo Edison's consultant is providing a preliminary analysis of current plant shielding effectiveness of sampling, letdown and accident mitigation systems under the assumed source terms in your letter b January 1,1980.

These results will then be evaluated based on area accessability requirements. Any proposed design modifications would be provided with an implementation schedule by February 20, 1980.

9 1T99 056 Docket No. 50-346 License No. NPF-3 Serial No. 559 November 21, 1979 Recommendation 2.1.7b - Auxiliary Feedwater (AFW) Flow Indication to Steam Generators for PWRs (Task No. 18).

Response

DB-1 has installed control grade AFW flow indicators as described in Toledo Edison's letter Serial No. 507, dated May 22, 1979. Additionally, DB-1 currently has safety grade steam generator level indication which provides the primary means to detect if the auxiliary feedwater systea is working properly. The flow indicators provide a backup to meet the single failure criteria.

With these indications, DB-1 complies with the January 1,1980 control grade instru-mentation recommendation.

Toledo Edison will install cae safety grade AW flow indicator per steam generator to provide a inckup indication of AW flow for the existing safety grade steam generator level. The safety grade AFW flow indicators are projected to be installed prior to January 1, 1981.

Recommendation 2.1.8a - Improved Post-Accident Sampling Capability (Task Nos. 19 - 21).

Response

DB-1 will review appropriate techniques to assure the feasibility of promptly sampling and analyzing reactor coolant and containment atmosphere under accident conditions.

In the short term, guidelines will be provided to the station by January 1,1980 for the drawing of highly radioactive reactor coolant liquid and containment vessel gaseous samples. These sampling techniques will be consistent with the current plant design capabilities which limit sample points to the reactor coolant letdown line and the pressurizer liquid. Guidelines will be provided by January 1, 1980 for on-site radiological analysis of reactor coolant samples as well as radiological and hydrogen analysis of the containment atmosphere.

The capability of on-site chemical analysis of reactor coolant is still under review.

DB-1 b9ing a fresh water site, need rot require chloride analysis. Orre area of concern in liquid chemical analysis is the dissolved gases evaluation. Due to current uncertainties in this area we cannot commit to prompt on-site chemical analysis of highly radioactive reactor coolant. Offsite laboratory analysis provides a short term alternate to this capability.

For the long term, a design review identifying plant modifications that will enhance the onsite sampling and analytical capability for highly radioactive reactor coolant and containment vessel gases will be completed in January,1980. The results of this review, a description of any planned modifict tions and a proposed schedule for implementation of these will be submitted to the NRC by January 25, 1980. ,

1799 057

Docket No. 50-346 License No. NPF-3 Serial No. 559 November 21, 1979 Recommendation 2.1.8b - Increased Range of Radiation Monitors (Task Nos. 22-24) .

Response

Toledo Edison is developing guidelines that will be available by January 1,1980 as an interim method of quantifying high level radioactive releases. Emergency power, independent of off-site power, will not be available until af ter the refueling outage, which is presently scheduled for Spring 1980. These guidelines sre being developed to utilize currently existing and/or portable equipment. Where control room readout capability is not available, communication with the control room will be provided.

Equipment related and procedural details will be addressed in the guidelines.

For the long termja design review identifying plant modifications to enhance this monitoring capability will be completed in January, 1980. The results of this review, a description of any planned modifications and a proposed schedule for implementation will be submitted to the NRC by January 31, 1960.

Recommendation 2.1.8c - Improved In-Plant Iodine Instrumentation (Task No. 25)

Response

Toledo Edison is developing the capability to detect the presence of iodine. Several options are currently being evaluated in sa area of interest following a radiological release. This capability will be available by January 1,1980. A review of planned design modifications to enhance the DB-1 capabilities in this area is underway. The results of the review, including a schedule of any design modifications will be submitted to the NRC by January 31,1980.

Recommendation 2.2. lb - Shif t Technical Advisor (Task No. 31)

Response

Toledo Edison is proposing a separation of the functions described in enclosure 2 of your September 13, 1979 letter. ,

The Shif t Technical Advisor (STA) will fulfill the function of accident assessment.

The specific details of how this individual is to administratively function is pre-sently dependent on manpower availability, further corporate direction and facility changes in progress. As a minimum, a designated individual will be available 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day whenever nue. lear safety systems are required to be operable beginning January 1, 1980. S2ation personnel will be required by procedure to notify the Shift Technical Advisor La the event of:

1. Any anticipated power transient

2. Any reactor trip or load run backs

3. Any loss of safety system function or offsite power supply The STA will be capable of reaching the control room within 10 minutes of receiving a call.

29o, nR8 -

Docket No. 50-346 License No. NPF-3 Serial No. 559 November 21, 1979 The second function, operating experisnce assessment, will be provided by a non-shif t oriented review group within the Toledo Edison organization. By January 1,1980, this function will be provided and its administration documented in procedures.

By January 1,1981, the Shift Technical Advisors will have completed training in the areas of reactor operations, as well as, transient and plant accident response. Additionally, a program will be implemented to expand the general technical edc: Lion level of any Shif t Technical Advisor that does not have a college level education in a scientific or engineering discipline. Completion of this part of the program may not be a prerequisite for Shif t Technical Advisor based on the individuals prior experience and an evaluation of his capability by the Toledo Edison Company. This evaluation will be subject to review by the Vice President of Energy Supply and his staff.

Recommendation 2.2.2b - On-site Technical Support Center (Task No. 34)

Response

As a part of an ongoing review of plant response to emergencies, Toledo Edison is reviewing the requirements of the or. site technical support center. By January 1,1980, an interim center will be established. This center will be in close proximity to the control room, have plant reference documents easily available and will have a communication link to the control room. A permanent location for this center is currently under review and will be determined by the results of an overall evaluation of emergency plant response facilities. The preliminary design of the permanent technical support center (PTSC) will be available by January 1,1980. The projected implementation schedule of the PTSC will be provided with the preliminary design.

e