ML19257C529

Forwards Summary of Methods of Implementing Short-Term Lessons Learned Commitments Actually Implemented by 800101 at Facility, in Response to NRC 800102 Request

ML19257C529
Person / Time
Site:	Davis Besse
Issue date:	01/18/1980
From:	Crouse R TOLEDO EDISON CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
RTR-NUREG-0578, RTR-NUREG-578 TAC-12453, NUDOCS 8001290344
Download: ML19257C529 (8)

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Docket No . 50-346 License No. NPF- 3 Serial No. 578 January 18, 1980 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Comission Washington, D. C. 20555

Dear Mr. Denton:

On December 31, 1979 Toledo Edison provided a status of efforts for complying with your staf f's short-term lessons learned recomendations as a result of the Three Mile Island, Unit #2 incident. This letter identifies the rnethods by which the recommendations Davis-Besse Nuclear Power Station, Unit 1 comitted to for January 1,1980 have been implemented. Items for which the previous plant desigr complies with your recomendations were discussed in Toledo Edison's letters of October 23 and November 21,1979 (Serial Nos. 546 and 559). Recom-mendations not implemented January 1,1980 or part of the previous plant design will be discussed in our response to your Show Cause Order of January 2,1980.

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

Very truly yours, ffh: 22=w RPC/TJM h 1829137 THE TOLEDO ED: SON COMPANY ECISCN PLAZA 300 MADISCN AVENUE TOLEDO. CHIO 43652 8001290 3Qd

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Docket No. 50-346 License No. NPF-3 Serial No. 578 January 18, 1980 Attachment 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 Nuclear Power Station Unit #1, (DB-1) for January 1,1980.

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).

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 PORV block valve control power and motor power has been modified as illustrated in Figure 1. Separation and independence of electrical systems has been maintained in accordance with the criteria of the DB-1 Final Safety Analysis Report.

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

Summary Mr. William Cahill, Jr., Chairman of the Electric Power Research Institute Safety and Analysis Task Force submitted to the NRC the " Program Plan for the Performance Verification 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 provide substantive 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 of inventory cases would be available by January 1, 1980. These were submitted to you by our letter dated December 27, 1979 and were concurrently implenented into the plant procedures.

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Recommendation 2.1.6a - Integrity of Systems outside Containment Likely to Contain 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 portions of the systems listed below. The detection method is indicat e d. Correction of excessive leakage conditions falls under the current plant preventive maintenance program.

Syst em Detection Method Reactor Coolant System Combination of :

a) Installed instrumentation b) Installed reactor coolant pump controlled leakage flow ins trumentation Makeup, Letdown, Seal Same as reactor coolant system Injection and Seal Return Systems Low Pressure Injection Pressure testing method on suction and Containment Spray and discharge piping Systems High Pressure Inj ection Inspection during pressurized System operation Waste Gas System Nitrogen and Ultrasonic leak detection method Primary Sample System Pressure Testing Method Reactor Coolant System Nitrogen and Ultrasonic leak Drain System (outside detection method containment up to and including the drain tank).

The following are excluded from the consolidated leakage reduction program identified above:

Pressurizer Quench Tank Recirculation System Clean and Miscellaneous Liquid Radwaste System The initial effort in the consolidated leakage reduction program not previously completed under DB-1 Technical Specification requirements is currently ongoing at the station in cooperation with Babcock and Wilcox personnel. .

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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 containment vessel. For evaluation purposes a normalized case assuming no isotopic decay time was assumed. Decay curves were also generated to provide detailed h1 formation for consideration in accessability requirements.

Systems were selected in four categories for the review.

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

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

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

c. The high pressure injection system used to recirculate water from the 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 to the first isolation valve which could contain atmosphere from the containment.

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

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

4) The letdown category included the portion of the letdown system from the reactor coolant system past the failed fuel detector up to the inlet valves to 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 onsite interim technical support center areas were verified to have direct dose rates less than 15mr/hr allowing continuous occupancy.

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

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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 liquid and containment vessel gaseous samples with the current plant sample points. These initial 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 analysis of the sample containers.

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

Onsite chemical analysis is not considered essential in the two hour time frane for event assessment. Off site laboratory analysis is an acceptable alternative f or the interim period until onsite capability is established to provide chemical analysis. Plant modification for onsite capability is a category "B" item and will 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 releases for the case when existing instrumentation goes off scale. These guidelines, currently being incorporated into plant procedure AD 1850.04 " Post Accident Radiological Sampling and Counting", provide for an extension of f the current station vent sample line to a partially shielded area for counting with a portable monitor. A calculational conversion method then is used to provide radioactive

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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 low volume air sample to provide in- plant iodine detection. The data analysis is done utilizing previously available equipment.

Recommendation 2.2.la - Shif t Supervisor's Responsibility Summary Toledo Edison has provided responsibility definition consistent with your staff's October 30, 1979 letter in revision 6 of station procedure AD 1839, " Station Operations". In addition, the shift foreman responsibilities have been underscored by 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 is providing an operating experience program separate from the shif t technical advisor and accident assessment function.

An operating experience program has been established with the Davis-Besse Station Technical Section as the lead organization. Details of review and reporting functions were 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 a rotating 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> shif t whenever the plant is not in a cold shutdown condition. The STA is provided with onsite living quarters to guarantee access to the control room within 10 minutes of notification. The responsibilities and details of his activities were formalized for January 1,1980 and are currently in the procedure review process.

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

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

Recommendation 2.2.2a - Control Room Access Summary -

Toledo has revised station procedure AD 1839, " Station Operations" to establish clear 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 floor of the Davis-Besse Unit 1 of fice building. A communication system is available to the control room and other key locations. S t atiot. procedure AD 1827.00 " Emergency Plan" has been revised to incorporate this interim center.

The preliminary design of the permanent technical support center was submitted to your 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 the assembly room on the turbine room floor. Communication with the control room and the onsite technical support center is provided. This designation was formalized in station procedure AD 182 7.00, " Emergency Plan".

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

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,_ Reversable Notes:

1) Essential power to

]-Fbtor Starter Motor Control Center sgajj 430/120V Contacts F12A is from Emergency 77 -r)

Control Power Diesel Generato r No.2.

Trans fo rme r 2) For details of AC b' "

electrical distribution Switch Mounted see Figure 8-4B of the in Control Room Davis-Besse Nuclear Power Station Unit 1, Final Safety Analysis Report.

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• me n Valve Limit t r Operator Switch f r the pressurizer Contacts pilot operated relief valve block valve.

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Figure 1