ML20062H637
| ML20062H637 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 07/26/1982 |
| From: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20062H632 | List: |
| References | |
| 50-409-81-23, NUDOCS 8208160121 | |
| Download: ML20062H637 (3) | |
Text
Appendix NOTICE OF VIOLATION Dafryland Power Cooperative Docket No. 50-409 As a result of the inspection conducted on December 22, 1981 through March 31, 1982, and in accordatcc with the NRC Enforcement Policy, 47 FR 9987 (March 9, 1982), the following violations were identified:
1.
Technical Specification 4.2.2.4.e states:
"The Forced Circulation loops shall not be pressurized unless their temperature is above 70 F and shall not be pressurized above 280 psig unless their temperature is at least 130'F."
Technical Specification 6.8.1 states, in part, " Written procedures shall be established, implemented and maintained."
Step 16 of Section 2.3.2 of the normal plant shutdown procedure in Volume I of the Operations Manual states, in part, "Every 15 minutes during reactor cooldown record the " Reactor Vessel Wall Temperature Recorder" on the " Reactor Heatup and Cooldown Data Sheet." Do not exceed the limits of Section 2.1.3 Step 2."
Section 2.1.3 Step 2 restricts cooldown rate at the forced circulation suction to 60*F/hr.
Annunciator alarm response in Volume I of the Operations Manual for Alarms E2-1 and E2-2 (FC Pump 1A and IB Tripped) states, in part, under Immediate Action Step 3, " Ensure temperature in the shutdown loop remains greater than 200*F by reducing seal inject flow into the shutdown pump to minimum allowable D/P."
Contrary to the above, on December 24, 1981, while recovering from the scram on December 23, 1981, an unintentional cooldown of the 1B forced circulation loop to 86*F occurred while loop pressure was greater than 280 psig. This violation of Technical Specification 4.2.2.4.e resulted from the following violations of the above mentioned procedures.
Plant personnel failed to take the required 15 minute reading for the 1A and 1B forced circulation loop suction temperatures.
Plant personnel failed to restrict the cooldown rate on the IB forced circulation loop to 60 F/hr.
@208160121 820727 PDR ADOCK 05000409 G
Appendix 2
Plant personnel increased rather than decreased the seal injection flow to the shutdown pumps.
This is a Severity Level IV violation (Supplement I).
2.
Technical Specification 6.8.1 requires, in part, that written procedures be implemented covering activities referenced in Appendix A of Regulatory Guide 1.33, Revision 2.
Appendix A of Regulatory Guide 1.33, Revision 2, references in Paragraphs 4.1 and 4.m operation of the main steam system.
Plant procedures contained in Sections 2.1.3, 2.3.2, 3.3.1.3 of Volume I of the Operations Manual and response procedures for Alarms Al-3, B14-3 and D2-3 provide for the use of the main steam bypass valve during normal plant startups and shutdowns and during times when the reactor coolant system pressure is increasing above an established value but indicate that the valve is normally closed during power operation.
(These plant procedures are consistent with the procedures and system descriptions contained in the safety analysis report.) Neither the plant procedures nor the safety analysis report indicate that the main steam bypass valve will be used to bypass all of the generated steam directly to the condenser during power operation.
Technical Specification 2.10.3.3 requires that the nuclear instrumenta-tion be capable of initiating scram actions as specified in Technical Specification Table 1.
Table 1, Item 1, requires the setpoint of the scram for Reactor Power-High to be in accordance with Table 4.0.2.2.1-1.
Table 4.0.2.2.1-1, Item 6.b, requires the trip setpoint for the Reactor Power-High to be 5 120% of rated thermal power wher.ever indicated power is above 15%.
Contrary to the above described plant procedures and safety analysis report provisions for use of the main steam bypass valve, on March 16, 1981, the reactor was operated at approximately 87% indicated power for approximately 12 minutes with all of the generated steam bypassed to the condenser via the main steam bypass valve. Additionally, a violation of Technical Specification 2.10.3.3 resulted since this mode of operation involved a loss of all feedwater heating which rendered the nuclear instrumentation incapable of accurately detecting power (i.e., the actual power was 102% when the indicated power was 87%) and initiating a scram at 5 120*a of rated thermal power.
This is a Severity Level IV violation (Supplement I).
Appendix 3
3.
10 CFR 50.59 states, in part, "(a)(1) The holder of a license author-izing operation of a production or utilization facility may...(ii) make changes in the procedures as described in the safety analysis report...
without prior Commission approval, unless the proposed change... involves a change in the Technical Specifications incorporated in the license or an unreviewed safety question...(b) The licensee shall maintain records of... changes in Irocedures made pursuant to this section...
These records shall inulude a written safety evaluation which provides the bases for the decermination that the change...does not involve an un-reviewed safety question..."
Sections 8.3 and 13.4 of the safety analysis repo t specify the operational uses of the main steam bypass valve and those uses do not include use of the valve to bypass the total steam flow directly to the condenser during power operation.
Procedures and changes thereto that are required for the operation of the main steam system are among those procedures which are required by Technical Specification 6.8.2 to be reviewed by the Operation Review Committee prior to implementation.
Contrary to the above, on March 17, 1981, Revision 1 to Operations Memorandum No. DPC-86 was implemented (issued) changing the procedure for the operation of the main steam bypass valve (i.e., allowing the valve to be used to bypass the total steam flow directly to tha condenser during high power operation) without a safety evaluation being performed to determine if an unreviewed safety question was involved and without review by the Operation Review Committee.
This is a Severity Level IV violation (Supplement I).
Pursuant to the provisions of 10 CFR 2.201, you are required to submit to this office within thirty days of the date of this Notice a written statement or explanation in reply, including for each item of noncompliance:
(1) cor-rective action taken and the results achieved; (2) corrective action to be taken to avoid further noncompliance; and (3) the date when full compliance will be achieved. Consideration may be given to extending your response time for good cause shown.
7d'll L O
6 Date8
/d James G. Keppler U
Regional Administrator