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Docket No . 50-346

License No. NPF- 3

Serial No. 578

January 18, 1980

Mr. Harold R. Denton, DirectorOffice of Nuclear Reactor RegulationU. S. Nuclear Regulatory ComissionWashington, D. C. 20555

Dear Mr. Denton:

On December 31, 1979 Toledo Edison provided a status of efforts forcomplying with your staf f's short-term lessons learned recomendationsas a result of the Three Mile Island, Unit #2 incident. This letteridentifies the rnethods by which the recommendations Davis-BesseNuclear Power Station, Unit 1 comitted to for January 1,1980 havebeen implemented. Items for which the previous plant desigr complieswith your recomendations were discussed in Toledo Edison's lettersof October 23 and November 21,1979 (Serial Nos. 546 and 559). Recom-mendations not implemented January 1,1980 or part of the previousplant design will be discussed in our response to your Show CauseOrder of January 2,1980.

This is considered to respond to description of implementation methodsrequested in your letter of January 2, 1980. We are available todiscuss any questions you may have on this submittal.

Very truly yours,

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RPC/TJM h

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THE TOLEDO ED: SON COMPANY ECISCN PLAZA 300 MADISCN AVENUE TOLEDO. CHIO 43652

8001290 3Qd

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Docket No. 50-346License No. NPF-3Serial No. 578January 18, 1980Attachment

Summary of the

Methods of Implementing

Short Term Lessons

Learned Coc:mitments Actually

Implemented by January 1,1980 at

The Davis-Besse Nuclear Power

St at ion, thit 1

182'/138~

Recommendations Committed to by Davis-Besse NuclearPower Station Unit #1, (DB-1) for January 1,1980.

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

Summary

The current design, as discussed in Toledo Edison letter of November 21, 1979(Serial No. 559), identify the pressurizer heaters, pilot operated relief valve(PORV) and pressurizer level indication as meeting the recommendation. The PORVblock valve control power and motor power has been modified as illustrated inFigure 1. Separation and independence of electrical systems has been maintainedin accordance with the criteria of the DB-1 Final Safety Analysis Report.

Recommendation 2.1.2 - Performance Testing for BWR and PWR Relief and SafetyValves (Task No. 6).

Summary

Mr. William Cahill, Jr., Chairman of the Electric Power Research Institute Safetyand Analysis Task Force submitted to the NRC the " Program Plan for the PerformanceVerification of PWR Safety / Relief Valves and Systems" on December 17, 1979.Toledo Edison considers this program to be responsive to this recommendation.

Toledo Edison's understanding of the current projected schedule is to providesubstantive program results by July,1981.

Recommendation 2.1.3b - Instrumentation for Detection of Inadequate Core Cooling

in BWRs and PWRs. (Task No. 8).

Summary

Toledo Edison identified that inadequate core cooling guidelines for loss ofinventory cases would be available by January 1, 1980. These were submittedto you by our letter dated December 27, 1979 and were concurrently implenentedinto the plant procedures.

1827139'

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

Summary

A leakage reduction program at DB-1 has been consolidated for applicable portionsof the systems listed below. The detection method is indicat e d. Correction ofexcessive leakage conditions falls under the current plant preventive maintenanceprogram.

Syst em Detection Method

Reactor Coolant System Combination of :a) Installed instrumentationb) Installed reactor coolant pump

controlled leakage flowins trumentation

Makeup, Letdown, Seal Same as reactor coolant systemInjection and Seal ReturnSystems

Low Pressure Injection Pressure testing method on suctionand Containment Spray and discharge pipingSystems

High Pressure Inj ection Inspection during pressurizedSystem operation

Waste Gas System Nitrogen and Ultrasonic leakdetection method

Primary Sample System Pressure Testing Method

Reactor Coolant System Nitrogen and Ultrasonic leakDrain System (outside detection methodcontainment up to andincluding the drain tank).

The following are excluded from the consolidated leakage reduction programidentified above:

Pressurizer Quench Tank Recirculation SystemClean and Miscellaneous Liquid Radwaste System

The initial effort in the consolidated leakage reduction program not previouslycompleted under DB-1 Technical Specification requirements is currently ongoingat the station in cooperation with Babcock and Wilcox personnel. .

lb.

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

S umma ry

Toledo Edison has completed area dose rate calculations outside the containmentvessel. For evaluation purposes a normalized case assuming no isotopic decay timewas assumed. Decay curves were also generated to provide detailed h1 formationfor consideration in accessability requirements.

Systems were selected in four categories for the review.

1) Recirculation systems included the applicable portions of the followingsystems:

a. The containment spray systems used to recirculate water from thecontainment vessel emergency sump into the containment vessel.

b. The decay heat removal system used to recirculate the water fromthe containment vessel emergency sump into the containm..nt vessel.

c. The high pressure injection system used to recirculate water fromthe low pressure injection system into the containment vessel.

2) Containment atmosp'are extensions include the applicable portions of:

a. Containment ventilation systems external to the containment up tothe first isolation valve which could contain atmosphere from thecontainment.

b. Sampling system used to obtain a containment atmosphere sample.

3) The liquid sample stegory included portions of rSe sampling system usedto meet post accident liquid sampling conditions f rom the reactor coolantsystem.

4) The letdown category included the portion of the letdown system from thereactor coolant system past the failed fuel detector up to the inlet valvesto the purification system demineralizers.

The radioactive source terms used were in the following release f ractions:

a. Containment atmosphere : 100% noble gases, 25% halogens

b. Reactor coolant: 100% noble gases, 50% halogens, 1% of Cesium and Rubidium

c. Containment sump liquid: 50% halogens, 1% of Cesium and Rubidium

The results of the preliminary analysis showed that the control room and onsiteinterim technical support center areas were verified to have direct dose ratesless than 15mr/hr allowing continuous occupancy.

The procedural actions required outside these areas were identified as allowinginf requent access doses below the guidelines identified in your staff's letterof October 30, 1979. More detailed evaluations are underway to optimize the f actorsof sampling procedures and techniques with proposed design modifications and willbe part of our proposed modifications submittal.

182;7I41

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

Summary

Guidelines have been generated to obtain highly radioactive reactor coolant liquidand containment vessel gaseous samples with the current plant sample points. Theseinitial guidelines provide for liquid samples f rom the pressurizer liquid space.These guidelines are currently being formalized into plant procedure AD 1850.04" Post Accident Radiological Sampling and Counting".

Prompt onsite radiological analysis capability is provided by spectral analysisof the sample containers.

More detailed evaluations are underway to optimize sampling techniques withshieldtag enhancements and will be reported as described in the summary ofrecommendation 2.1.6b. This is required to handle the referenced source terms.

Onsite chemical analysis is not considered essential in the two hour time franefor event assessment. Off site laboratory analysis is an acceptable alternativef or the interim period until onsite capability is established to provide chemicalanalysis. Plant modification for onsite capability is a category "B" item andwill be discussed in our design modification submittal later this month.

Recommendation 2.1.8b - Increased Range of Radiation Monitors

S umma ry -

Guidelines have been previded to quantify high level accidental radioactive releasesfor the case when existing instrumentation goes off scale. These guidelines,currently being incorporated into plant procedure AD 1850.04 " Post AccidentRadiological Sampling and Counting", provide for an extension of f the currentstation vent sample line to a partially shielded area for counting with a portablemonitor. A calculational conversion method then is used to provide radioactive

effluent release rate.-

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

Sunna ry

Toledo Edison now has available silver zeolite collection cartridges for a lowvolume air sample to provide in- plant iodine detection. The data analysis isdone utilizing previously available equipment.

Recommendation 2.2.la - Shif t Supervisor's Responsibility

Summary

Toledo Edison has provided responsibility definition consistent with your staff'sOctober 30, 1979 letter in revision 6 of station procedure AD 1839, " StationOperations". In addition, the shift foreman responsibilities have been underscoredby letter from Toledo Edison's corporate management.

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Recommendation 2.2.1.b - Shif t Technical Advisor

Summary

As indicated in our letter of October 23,1979 (Serial No. 546), Toledo Edison isproviding an operating experience program separate from the shif t technical advisorand accident assessment function.

An operating experience program has been established with the Davis-Besse StationTechnical Section as the lead organization. Details of review and reporting functionswere formalized January 1, 1980 and are currently in the procedural review process.

The shift technical advisor (STA) at Davis-Besse Unit 1 is currently on duty for arotating 24 hour shif t whenever the plant is not in a cold shutdown condition. TheSTA is provided with onsite living quarters to guarantee access to the control roomwithin 10 minutes of notification. The responsibilities and details of hisactivities were formalized for January 1,1980 and are currently in the procedurereview process.

Recommendation 2. 2. lc - Shif t and Relief Turnover Procedures

Summary -

Toledo Edison has revised station procedure AD 1839, " Station Operations" to beconsistent with this recommendation.

Recommendation 2.2.2a - Control Room Access

Summary -

Toledo has revised station procedure AD 1839, " Station Operations" to establishclear lines of authority in the control room that extend to access control.

Recommendation 2.2.2b - Onsite Technical Support Center

Summa ry

The interim unsite technical support center has been designated to be the fif th floorof the Davis-Besse Unit 1 of fice building. A communication system is available tothe control room and other key locations. S t atiot. procedure AD 1827.00 " EmergencyPlan" has been revised to incorporate this interim center.

The preliminary design of the permanent technical support center was submitted toyour staf f on December 27, 1979 (letter Serial No. 571).

Recommendation 2.2.2c - Onsite Operational Support Center

Su=marv

The onsite operational support center for Davis-Besse Unit 1 is designated as theassembly room on the turbine room floor. Communication with the control room andthe onsite technical support center is provided. This designation was formalizedin station procedure AD 182 7.00, " Emergency Plan".

1812{/ 143

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480V, 36, 60 Hz., Essential Motor Control Center F 12 A

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,_ Reversable Notes:

]-Fbtor 1) Essential power toStarter Motor Control Center

sgajj 430/120V Contacts F12A is from Emergency77 -r) Diesel Generato r No.2.Control Power

Trans fo rme r 2) For details of ACb' electrical distribution"

Switch Mounted see Figure 8-4B of thein Control Room Davis-Besse Nuclear Power

Station Unit 1, FinalSafety Analysis Report.

.*

me n

Valve Limitt r OperatorSwitch

f r the pressurizerContacts

pilot operated reliefvalve block valve.

I82j7144.

Figure 1