ML20008E221

Forwards Request That TMI-2 Lessons Learned Radiological Tech Specs Be Modified.Submits Info Required Per 800702 Request

ML20008E221
Person / Time
Site:	Cooper
Issue date:	10/15/1980
From:	Pilant J NEBRASKA PUBLIC POWER DISTRICT
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
NUDOCS 8010240431
Download: ML20008E221 (7)

Text

.- .

Nebraska Public Power District " #bbbbhb .^1 " ^ " '

• October 15, 1980 Darrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

TMI-2 Lessonc Learned -

Proposed Radiological Technical Specifications Cooper Nuclear Station NRC Docket No. 50-298, DPR-46 Reference (1) Letter from D. G. Eisenhut to All Boiling Water Reactor Licensees dated July 2,1980, same subject (2) Letter from T. A. Ippolito to J. M. Pilant dated April 10, 1980, " Evaluation of Licensee's Compliance with Category A Items of NRC Recommendations Resulting from TMI-2 Lessons Learned" (3) Letter from D. G. Eisenhut to All Operating Power Reactor Licencees dated August 7, 1980, same subject In accordance with the guidance and model technical hecifications (TSs) provided in Reference 1 and our letter of September 15, 1980, Nebraska Public Power District requests that the Cooper Nuclear Station Radiological Technical Specifications be modified as indicated on the enclosed revised pages. Revisions are marked by a vertical bar in the margin.

The following addresses each of the items required by Reference 1 and either enumerates where this item is provided for by the present Technical Specifications o; on which revised page it has been added.

1. Emergency Power Supply / Inadequate Core Cooling The requirements for operability and surveillance of the reactor vessel water level instrumentation are presently adequately addressed in Tables 3.2.F (page 65) and 4.2.F (page 80). The bases section (page 87) for these tables is revised to reflect the accident accessment purpose of this instrumentation.

Y L

0010240 l .)

Darr ll G. Eis:nhut October 15, 1980 Page 2

2. Valve Position Indication The operability and surveillance requirements for the safety valve position fndicating thermocouples and the relief valve position indicating pressure switches are specified on revised page 136. The bases section for this specification on page 150 has also been revised.

3. Containment Isolation Table 3.2.a (pages 50, 51, 52, 52a, and 52b) defines the operability requirements for the containmert isolation instrumentation which provides for the necessary diverse isolation signals. Specification 4.7.D.1.a (page 166) assures that simulated automatic initiation is conducted and that closure times are recorded for primary containment isolation valves. Table 3.7.1 (page 168) contains the acceptable isolation times for the valves. Table 4.2. A (page 68 and 69) define the appliable surveillance requirements for the initiating sensors and associated trip channels which assure the capability for automatic isolation. The bases for the containment isolation times are as stated en page 183 and 184.

4. Shift Technical Advisor (STA)

Presently the position of Shift Technical Advisor is fulfilled in an interim manner as committed by the District and as agreed to by the staff in Reference 2, Item 2.2.1.b. In accordance with Enclosure 1 of the September 5,1980, letter from D. G. Eisenhut to all licensees, the District will submit the final STA program including training and requalification plans by January 1,1981. Technical Specifications for the STA will be proposed prior to the time that the final STA program at CNS is implemented. The subject of minimum shift crew composition as reflected on page 18 of the TSs will be addressed in the District's response to D. G. Eisenhut's letter to all licensees dated July 31,1980, " Interim Criteria for Shift Staffing."

5. System Integrity and Improved Iodine Measurements Program The Federal Register advance notice of proposed rule making July 8,1980, (45916) discussed proposed changes to NRC regulations pertaining to technical specifications. In agreement with this effort, the District fully supports the contention that increasing the volume of technical specifications decreases the effectiveness of these specifications to focus attention on matters of more immediate importance to safe operation of any facility. In this regard we are presently reviewing the CNS Technical Specification in an effort to identify and request deletions of those ineffective

Darrell G. El:Inhut October 15, 1980 Page 3 passages which do not contribute directly to safe operations.

The inclusion, at this time, of technical specifications addressing the System -Integrity and Iodine Measurements programs would be counterproductive to this important effort.

It should be noted that these changes alter the page and specification numbering sequences submitted in the District's proposed Radiological Effluent (Appendix I) Technical Specifications of January 30, 1980.

In accordance with the request made in Reference 3, this change has been reviewed by the appropriate District Safety committees and has been judged to be a Class II amendment. Payment in the amount of $1,200 is enclosed.

Should you have any questions or require further clarification, please contact me.

In addition to three signed originals, 37 copies are also submitted for your use.

Sincerely, M

Jay . Pflant Director of Licensing &

Quality Assurance JMP/jdw:bn17/3 Enclosure

Darrell G. Eistnhut October 15, 1980 Page 4 STAi'E OF NEBRASKA)

)ss PLATTE COUNTY )

Jay M. Pilant, being first duly sworn, deposes and says that he is an authorized representative of the Nebraska Public Power District, a public corporation and political subdivision of the State of Nebraska; that he is duly authnrized to submit this request on behalf on Nebraska Public Power Elstrict; and that the statements in said application are true to the best of his knowledge and belief.

W CJaV M. Pilant Subscribed in my presence and sworn to before me this 20 # day of

$fdd ,/PFD

~

l -) _ o- /

NO .AY PE SLIC f M ut3AL N0tAST-Onese d Asheebe avmAw.cArner My Commission expires ._ * * * " 7' " _ .

. .i 3.2 . BASES'(cont'd.)

Trip settings of <100 mr/hr for the monitors in the ventilation exhaust ducts are based upon initiating normal ventilation isolation and standby gas treat-ment system operation so that none of the activity released during the re-fueling accident leaves the Reactor Building via the normal ventilation path but rather all the activity is processed by the standby gas treatment system.

Flow transmitters are used to record the flow of liquid from the drywell sumps. . An air. sampling system is also provided to detect leakage inside the primary containment.

For each parameter monitored, as listed in Table 3.2.F, there are two (2) channels of instrumentation. By comparing readings between the two (2) channels, a near continuous ' surveillance of instrument performance is available.

ble. Any deviation in readings will initiate an early recalibration, there-by maintaining the quality of the instrument readings.

The recirculation pump trip has been added as a means of limiting the con-sequences of the unlikely occurrence of a failure to scram during an antici-pated transient. The response of the plant to this postulated event falls within the envelope of study events given in General Electric Company Topical Report, NED0-10349, dated March, 1971.

The liquid radwaste monitor assures that all liquid discharged to the discharge canal does not exceed the limits of Section 2.4.1.b of the Enviromnental Technical Specifications. Upon sensing a high discharge level, an isolation signal is generated which closes of radwaste discharge valve. The set point is adjustable to compensate for variable isotopic discharges and dilution flow rates.

The main control room ventilation isolation is provided by a detector monitoring the intake of the control room ventilation system. Automatic isolation of the normal supply and exhaust and the activation of the emergency filter system is provided by the radiation detector trip function at the predetermined trip level.

The mechanical vacuum pump isolation prevents the exhausting of radioactive gas thru the 1 minute holdup line upon receipt of a main steam line high radiation signal .

The operability of the reactor water level instrumentation in Tables 3/4.2.F ensures that sufficient information is available to monitor and assess accident situations.

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.D Safety and Relief Valves 4.6.D Safety and Relief Valves

1. During reactor power operating condi- 1. Approximately half of the safety valves tions and prior to reactor startup and relief valves shall be checked or f rom a Cold Condition, or whenever replaced with bench checked valves reactor coolant pressure is greater once per operating cycle. All valves than atmospheric and temperature will be tested every two cycles.

greater than 2120F, all three safety valves and the safety rodes of all The set point of the safety valves relief valves shall be operable, ex- shall be as specified in Specification cept as specified in 3.6.D.2. 2.2.

2. 2. At least one of the relief valves shall be disassembled and inspected each re-

a. From and after the date that the fueling outage, safety valve function of one relief valve is made or found to be inopera- 3. The integrity of the relief safety valve ble, continued reactor operation is bellows on any three stage valve permissible only during the succeeding shall be continuously monitored.

thirty days unless such valve function is sooner made operable. 4. Tne operaullity of the bellows monitoring system shall be demonstrated once every

b. From and after the date that the safety three months when three stage valves valve function of two relief valves is are installed, made or found to be inoperable, con-tinued reactor operation is permissible 5. Once per operating cycle, with the only during the succeeding seven days reactor pressure > 100 psig, each relief unless such valve function is sooner . valve shall be manually opened until made operable. the main turbine bypass valves have closed to compensate for relief valve

3. If Specification 3.6.D.1 is not met, opening.

an orderly shutdown shall be initiated and the reactor coolant pressure shall 6. The position of the safety and relief be reduced to a cool shutdown condi- valves shall be continuously monitored.

tion within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

7. Operatllity of the relief valve position indicating pressure switches and the safety valve position indicating thermocouples shall be demonstrated once per operating cycle.

-136-

. e *$

3.6.D & 4.6.D BASES (cont'd.)

The relief and safety valves are bench tested every second operating cycle to ensure that their set points are within the + 1 percent tolerance. Additionally, once per operating cycle, each relief valve is tested manually with reactor pressure above 100 psig to demonstrate its ability to pass steam.

The requirements established above apply when the nuclear system can be pres-surized above ambient conditions. These requirements are applicable at nuclear system pressures below normal operating pressures because abnormal operational transients'could possibly start at these conditions such that eventual overpressure relief would be needed. However, these transients are much less severe, in terms of pressure, than starting at rated conditions.

The valves need not be functional when the vessel head is removed, since the nuclear system cannot be pressurized.

The position indicating pressure switches for the relief valves and the thermo-couples for the safety valves serve as a diagnostic aid to the operator in the event of a safety / relief valve failure.

E. Jet Pumps Failure of a jet pump nozzle assembly hold down mechanism, nozzle assembly and/or riser, would increase the cross-sectional flow area for blowdown following the design bases double-ended line break. Therefore, if a failure occurs, repairs must be made.

The detection technique is as follows. With the two recirculation pumps

, balanced in speed to within + 5%, the flow rates in both recirculation loops will be verified by Control Room monitoring instruments. If the two flow rate values do not differ by more than 10%, riser and nozzle assembly integrity has been verified. If they do differ by 10% or more, the core flow rate measured by the jet pump diffuser differential pressure system must be checked against the core flow rate derived from the measured values of loop flow to core flow correlation. If the difference between measured and derived core flow rate is 10% or more (with the derived value higher) diffuser measure-ments will be taken to define the location within the vessel of failed jet pump nozzle (or riser) and the plant shut down for repairs. If the potential blowdown flow area is increased, the system resistance to the recirculation pump is also reduced; hence, the affected drive pump will "run out" to a substantially higher flow rate (approximately 115% to 120% for a single nozzle failure). If the two locps are balanced in flow at the same pump speed, the resistance characteristics cannot have changed. Any imbalance between drive loop flow rates would be indicated by the plant process instrumentation.

In addition, the affected jet pump would provide a leakage path past the core thus reducing the core flow rate. The reverse flow through the inactive jet pump would still be indicated by a positive differential pressure but the net effect would be a slight decrease (3% to 6%) in the total core flow measured.

This decrease, together with the loop flow increase, would result in a lack of correlation between measured and derived core flow rate. Finally, the affected jet pump diffuser differential pressure signal would be reduced because the backflow would be less than the normal forward flow.

A nozzle-riser system failure could also generate the coincident failure of a

-150-r

>< - _ ,