ML19257B939

Forwards Summary Listing Used to Implement All Lessons Learned Task Force short-term Requirements,In Response to NRC 791030 Ltr

ML19257B939
Person / Time
Site:	Cooper
Issue date:	01/11/1980
From:	Pilant J NEBRASKA PUBLIC POWER DISTRICT
To:	Harold Denton Office of Nuclear Reactor Regulation
References
RTR-NUREG-0578, RTR-NUREG-578 TAC-12408, NUDOCS 8001210277
Download: ML19257B939 (8)

Text

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' CcNcRAL OFFICE

\ P. O. BO X 499. COLUMBUS, Nc BR ASKA 686g!

ic Power Distr.ic t reLe ~ o~e m n s.4-es.1

) @ ._ Nebraska Publ.- . - _ - -

January 11, 1980 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Lessons Learned Requirements Implementation Methods Cooper Nuclear Station NRC Docket No. 50-298, DPR-46

Reference:

1) Letter from H. R. Denton to All Operating Nuclear Power Plants, dated October 30, 1979, " Discussion of Lessons Learned Short Term Requirements"

2) Letter from T. D. Keenau (Vermont Yankee - BWR Owners Group) to D. G. Eisenhut, dated October 17, 1979, "BWR Owners Group Positions on NUREG 0578"

3) Letter from H. R. Denton to J. M. Pilant Dated January 2, 1980

Dear Mr. Dencon:

Per the NRC request made in Reference 1, enclosed is a summary listing of the methods used by Nebraska Public Power District to implement all of the Lessons Learned Short Term Requirements at Cooper Nuclear Station.

As usual, complete documentation of the analyses discussed hetain is available at the station for ISE review.

If additional clarification is ne essary regarding any of the enclosed information, please do not hesitate to contact me.

Sincerely, M _.

Jay .. Pilant Director of Licensing and Quality Assurance JDW/cauc Enclosure

- h 17.74 198 S i/l

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Page 1 of 7.

Nebraska Public Power District NdREC 0578 Implementation Methods ,

NUREC 0578 Section 2.1.1 There is no need for action in response to this requirement for the reasons stated in Emergency Power Reference 2.

Supply Requi.ement 2.1.2 Nebraska Public Power District endorses the relief and safety valve testing program being Relief and Safety developed by the CE BWR Owners Group.

Valve Testing 2.1.3.a Pressure switches have been ordered which will be installed on the discharge piping of each Direct Indication safety / relief valve. Details of the procurement efforts conducted to date were provided in of Valve Position response to the staff's telephone request January 7, 1980. Additional information regarding the date by which the pressure switches will be installed will be submitted in the response to the Show Cause Order issued by Reference 3.

2.1.3.b Additional hardware to identify inadequate core cooling on BWR's has not been determined to Instrumentation for In- be necessary at this time. Analyses and operator guidelines for the detection and mitigation adequate Core Cooling of inadequate core cooling are currently being developed per Requirement 2.1.9 and questions from the Bulletins and Orders Task Force.

2.1.4 The design of Cooper Nuclear Station is such that there is diversity of parameters sensed Diverse Containment for the initiation of containment isolation. A review has been completed of all non-essential Isolation systems penetrating the primary containment and it has been determined that these systems do in fact receive automatic isolation signals.

(( Containment isolation provisions upon reset of the isolation signal was reviewed and it was

,,g.s determined that 14 air operated isolation valves require modification to their control systems.

43, Administrative controls are presently in effect to prevent these valves from reopening.

. Applicable hardware has been ordered (i.e. switches, relays, etc.) and the details of the pro-

---_ curement efforts conducted to date were provided in response to the staff's telephone request s4) January 7, 1980. Additional information regarding the date by which the hardware will be

'43 ' installed will be submitted in the response to the Show Cause Order issued by Reference 3.

2.1.5.a . This position is not applicable to Cooper Nuclear Station since the licensing basis did Dedicated Ilydrogen not include requirements for external recombiners or purge systems for post-accident Control Penetrations combustible gas control of the containment atmosphere.

2.1.5.c This position is not applicable to Cooper Nuclear Station since the licensing basis did Recombiner Procedures not include requirements for hydrogen recombiners.

Page 2 of 7.

NUREC 0578 Section 2.1.6.a The following systems ahich may contain radioactive fluids post-LOCA have been reviewed '

Systems Integrity for for leakage rate reduction (the p..renthetical numbers denote flow diagrams):

liigh Radioactivity

a. Misc. drains, vents, and seal systems (B&R 2005).

b. Reactor Building and drywell equipment drains (B&R 2028).

c. Floor drain processing (B&R 2032 Sheets 1, 2, 3, and 4),

d. Equipment drain processing system (B&R 2033 Sheets 1, 2, 3, and 4).

e. Reactor Building and drywell floor drain system (B&R 2038).

f. R11R System (B&R 2040) .

g. Main Steam Reactor Building (B&R 2041),

h. Reactor Water Cleanup System (B&R 2042 Sheets 1, 2, and 3).

i. RCIC System'(B&R 2043).

J. IIPCI System (B&R 2044) .

The following corrective actions will be taken:

a. The gland seal leakoffs will be removed from the following valves which are located outside Primary Containment. This medification was recommended by Anchor Valve, and Minor Design Change 78-22 has been written to document the modification:

(1) IIPCI MO-16 (2) IIPCI MO-19 (3) IIPCI MO-20 (4) CS MO-11A (5) CS MO-12A (6) RCIC MO-20 (7) RCIC MO-21 (8) WIR 110-18 (9) RilR MO-25A & B (10) RilR MO-27A & B (11) RilR MO-33 (12) MSIV 863 & C 2[h.

The leakoff port will be plugged and welded and the valves packed with Crafoil during sa . . the 1980 refueling outage.

4

b. Valves in lines that dump to the floor drain system or the main condenser will be pro-IN) cedurally closed in the event of an accident that has the potential of fuel clad failure.

O These valves are as listed:

O (1) llPCI 559AV(A039) and 560AV(A040) - floor drain (2) RCIC 580AV(A012) and 584AV(A013) - floor drain

Page 3 of 7-NUREC 0578 Section 2.1.6.a (3) MS 783 AV(A042) and 782AV(A041) - condenser ,

(continued) (4) MS 780AV(A034) and 781AV(A035) - condenser (5) MS 786AV and 787AV - condenser (6) MS 784AV and 785AV - condenser

c. The vents from the liquid collecting tanks will be evaluated for the best method of venting during post-accident recovery operations.

The leakage rate from components within the Reactor Building has been measured and/or estimated. Total leakage is approximately 2 ML/ min. The following is a breakdown of the leakage:

a. IIPCI MO-16 1.5 M1/ min. (This valve will have the seal leakoff line removed and will be repacked with Grafoil during the March 1980 refueling outage.)

b. CS A pump seal 0.5 Ml/ min.

Preventive Maintenance - The station operators will continue to make routine tours of the Reactor Building at least three times per day. Management personnel will continue to make tours periodically. If leaks are noted on these tours, they will be identified and work requests written for repair and/or corrective action. These work requests are identified so they can be followed by the Operations and Maintenance Supervisors for prompt repair.

2.1.6.b A TID 14844 radioactivity release was assumed into the primary containment. The lines Plant Shielding Review penetrating the primary containment have been identified by a drawing review and a physical walk through. The vital areas of the reactor building have also been *dentified by a drawing review and a physical walk through. Radiation levels during accident conditions at the various vital areas have been calculated. These calculated radiation levels have been indicated on plant drawings that will be located in the Technical Support Center. At this time, calculations show that radiation readings could be extreme in the reactor y building during accident conditions and that these readings could come from multiple lines N* and sources. Feasible modifications and/or temporary shielding to reduce such readings N are being investigated.

BWR plants are specifically designed to mitigate major design basis events with no access g outside the control room. With lesser accidents, personnel access into the reactor building s may be possible providing monitoring instruments are used to determine radiation readings at the planned path of entry. The drawings in the Technical Support Center will be used for this planning.

Page 4 of 7 -

14UREC 0578 Section 2.1.7.a This requirement is not applicable to Boiling Water Reactors. f Auto Initiation of Auxiliary Feed 2.1.7.h This requirement is not applicable to Boiling Water Reactors.

Auxiliary Feed Flow Indication 2.1.8.a A design and operational review of existing reactor coolant and containment atmosphere Post-Accident sampling facilities has been completed and a new emergency sampling procedure has been Sampling approved and implemented. Depending on accident severity, this procedure, gives guidance for the emergency sampling of reactor coolant, containment atmosphere, and vent samples from existing sample locations. It has been determined from the 2.1.6.b plant shielding review that.a design modification of the sample locations will be necessary in order to collect samples under extreme post-accident conditions.

The sampling procedure allows for pre-analysis preparation of samples to ensure low back-ground conditions in the laboratory and counting rooms. At this point, boron and chloride analysis can be performed. Once the samples have been prepared for analysis, existing pro-cedures will quantify radioisotopes that are indicators of the degree of core damage, radio-isotopes that are released to the containment, and those that may be released to the envi-

((}, ronment. All sample apparatus required in the emergency sampling procedure have been con-structed.

,%g 4a. .

The analysis of pressurized and un-pressurized reactor coolant samples and dissolved 02 and rs; lig has been evaluated and a sampling apparatus is presently being designed. This sampling CZ) apparatus will require major changes to existing sample point locations. All necessary FNJ plant modifications will be implemented by January 1, 1981.

2.1.8.b Cooper Nuclear Station has developed procedures to quantify release rates up to 10,000 Ci/sec illgh Range Radiation for noble gases from all potential release points.

Monitors For immediate determination (or estimation) of noble gas and radiolodine release rates if existing effluent instrumentation goes off scale, a grab sample, via existing vent samplers or monitors, can be drawn quickly and removed to a low background area. The sample is then evaluated with a dose rate instrument. Calculation to convert dose rate readings to release rates are provided in the emergency sampling procedure. These samples may then be analyzed isotopically.

Sample apparatus has been constructed to allow for quick grab sampling of noble gas, radio-iodines, and particulates on all veut releases. With proper sample preparation, the sample would be isotopically analyzed for actual activity. Existing vent samplers may provide filters and cartridges for radiolodines, and particulates on all vent releases.

Page 5 of 7 -

NUREC 0578 Section 2.1.8.h As discussed in 2.1 8.a post-accident sampling, design modification of certain sample ,

(continued) locations will be necessary due to the plant shielding review, in order to collect samples under the post-accident conditions described in 2.1.6.b. All necessary plant modifications will be implemented by January 1, 1981.

A purchase order was issued to General Electric Co. November 1, 1979 to perform generic prototype equipment design and qualification services for high range radiation monitors.

Per the guidance in Reference 1, design details will be submitted for NRC approval prior to installation.

2.1.8.c Procedures and equipment have been developed which provide CNS with the capability to Improved Iodine accurately assess iodine concentrations. Portable air samplers are used to collect both Instrumentation particulate and iodine samples. These samples can be readily removed to a low background area and purged of entrapped noble gases. The samples are than isotopically analyzed in a laboratory counting room. As discussed in 2.1.6.b, Plant Shielding Review, and 2.1.8.a.

Post-Accident Sampling, dose to health physics personnel has been considered in the procedure.

2.1.9 The specific requirements and schedules relating to this position are being developed in a Transient and continuing series of meetings between the BWR Owners Group and the NRC Bulletins and Orders Accident Analysis Task Force.

The implementation of emergency procedures and retraining will be done on a schedule con-sistent with those established with the Bulletins and Orders Task Force.

ACRS Items - NPPD will meet this position with installation of equipment completed by January 1,1981.

Containment Pressure A purchase order w's issued to General Electric Co. hovember 1, 1979 to perform generic Water Level and prototype equipment design and qualification services for these monitors.

Ilydrogen Monitors

11. R . Denton Item - . NPPD concurs in the consensus of the CE BWR Owners Group that the Cooper Nuclear Station

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RCS Venting existing reactor venting capability is fully satisfactory. Justification for this position

d was presented to the NRC during the October 11, 1979 topical meeting and plant specific

))j design features were submitted in our letter to the NRC of October 30, 1979.

2.2.1.a ps; Plant procedures 1.2 " Station Organization and Responsibilities" and 1.4 " Station Rules Shift Supervisor C2) of Practice" have been reviewed and revised as necessary to assure that reactor operations Responsibilities (/a command and control responsibilities and authority are properly defined.

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