IR 05000263/1987003
| ML20207R628 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 03/11/1987 |
| From: | Jackiw I NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20207R618 | List: |
| References | |
| 50-263-87-03, 50-263-87-3, NUDOCS 8703170469 | |
| Download: ML20207R628 (6) | |
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U.S NUCLEAR REGULATORY COMMISSION
REGION III
Report No. 50-263/87003(DRP)
Docket No. 50-263 License No. DPR-22 Licensee: Northern States Power Company 414 Nicollet Mall Minneapolis, MN 55401 Facility Name: Monticello Nuclear Generating Station Inspection At: Monticello Site, Monticello, Minnesota Inspection Conducted: January 23 through February 13, 1987 Inspector:
P. L. Hartmann Approved By:
IT a i b$1 3 -//'[7 Re,ac,torProjectsSection2C Date
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Inspection Summary Inspection on January 23 through February 13, 1987 (Report No. 50-263/87003(DRp))
Areas Inspected:
Special, announced inspection by the resident inspector to review the circumstances surrounding the isolation of Control Rod 22-07 for maintenance.
Results: One violation was identified relating to an inadequate procedure used during preventive maintenance.
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DETAILS 1.
Persons Contacted a.
Northern States Power Company (NSP)
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Eliason, General Manager, Nuclear Plants W. A. Shamla, Plant Manager
- M. H. Clarity Assistant to the Plant Manager D. E. Nevinski, General Superintendent, Engineering & Radiation Protection D. D. Antony, General Superintendent, Operations W. J. Hill, Superintendent Technical Engineering W. W. Albold, General Superintendent, Maintenance B. D. Day, Superintendent, Operating Engineering b.
U.S. Nuclear Regulatory Commission (NRC)
- I. N. Jackiw, Chief, Projects Section 2C
- P. L. Hartmann, Senior Resident Inspector, Monticello The inspector also contacted other licensee employees including members of the technical and engineering staffs, and reactor and auxiliary operators.
- Denotes the licensee and NRC representatives attending the management exit interview.
2.
Event On January 22, 1987, while at 100 percent power, Control Rod 22-07 was made inoperable while fully withdrawn in preparation for minor maintenance on a nitrogen isolation valve (No. 111).
The isolation oftheHydraulicControlUnit(HCU)wascompleteat6:10a.m. This isolation rendered the rod incapable of being moved by manual rod notion or being scrammed from HCU pressure (accumulator, control rod dri,a) and reactor pressure.
At about 6:40 a.m., the oncoming site superintendent and the offgoing site superintendent discussed whether the control rod should have been made inoperable with the rod fully withdrawn.
The oncoming site superintendent determined that Technical Specification (Tech Spec) 3.3.A was incorrectly interpreted and the rod could not be made inoperable without a prior calculation of adequate shutdown margin (SDM). At about 7:30 a.m., efforts were begun to return Rod 22-07 to operable status.
At 8:06 a.m., the rod was capable of being scrammed.
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O Plant management was aware of the problem and ordered a SDM calculation by the nuclear engineering department and independently by GE.
The licensee informed the inspector at 2:25 p.m., January 22, 1987, that adequate SDM existed for Rods 22-07 and 22-11 (most reactive rod) at the fully withdrawn position with the core in its most reactive condition.
The GE calculation, which is code certified by the NRC, showed acceptable SDM.
3.
Analysis a.
Introduction A number of weaknesses were noted that contributed to this occurrence.
The procedure for isolating the HCU did not provide sufficient warning to the user to verify that adequate SDM existed before isolaticn; the work request authorization (WRA) did not list the applicable tech specs and the operating shift supervisors did not correctly interpret Tech Spec 3.3.A.1.
These weaknesses are discussed below.
b.
Inadequate Work Request and Procedure
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The Monticello Plant uses several methods to control maintenance and isolate systems for maintenance.
For some routine maintenance activities, an Operations Controlling) Document (OCD) is used as in the case of isolating a HCU (OCD 4201 and for depressurizing a HCU nitrogenaccumulator(0CD4208).
Typical past practices at Monticello have been to isolate the HCU only when the reactor is shut down or with the associated control rod inserted. OCD 4208, which cnly depressurizes the control rod accumulator, is usually used to work on Nitrogen Valve No. 111.
The rod is capable of being scrammed from reactor pressure in this situation. An accumulator can be made inoperable with the control rod withdrawn in accordance with Toch Spec. 3.3.D.
However, in this case the system engineer designated this work to be done by OCD 4201 which makes the HCU inoperable and the rod unavailable to scram because of the HCU isolation.
Work Request 87-0206 was written by the system engineer for the purpose of repairing Nitrogen Valve No. 111.
The WRA required isolation (per OCD 4201) of the liCU for Rod 22-07. Administrative Control Directive 4 ACD 15.5 titled WRA Preparation and Content, Part 6.4, Sections 2 and 3, Item 6 (Tech Spec) states in part:
"If technical specification requirements are involved, then the appropriate page number stould be identified and the apprupriate technical spet.eiication section shall be identified. The " Tech Spec" block in Section 2 shall be checked "Yes,"
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On WRA 87-0206, the tech spec block was checked "no" by the preparer and no page was referenced.
The preparer erroneously believed this was not necessary as the OCD listed the technical specification.
In accordance with 10 CFR Part 2. Appendix C,Section V.A., a Notice of Violation will not be issued for the violation because it meets all of the following tests:
(1)
It was identified by the licensee.
(Thelicenseeidentified the problem to the resident inspector.)
(2)
It fits in Severity Level IV or V.
(This violation would be classified as Severity Level IV.)
(3)
It was reported; if required.
(This event was not required to be reported. However, the subject was brought to the attention of the resident inspector during his investigation of the subject event).
(4)
It was or will be corrected, including measures to prevent recurrence, within a reasonable time.
(The violation was inanediately corrected and the responsible individual was retrained. To prevent recurrence, the licensee plans to conduct additional training on this subject.)
(5)
It was not a violation that could reasonably be expected to have been prevented by the licensee's corrective action for a previous violation.
(No previous violation of this type was identified.)
The Procedure used for this work was preventive maintenance OCD 4201, and was identified by the WRA which included a prerequisite that listed:
"1.(p) Technical Specifications 3.3.A and 3.3.D."
The site superintendent and shift supervisor (SR0s) were aware that Tech Spec 3.3.A applied to the isolation.
The first isolation step (Step No. 3) of 000 4201 states:
" Isolate the HCU i.a.w. Operations Manual B.1.3,Section VI.F.1, HCU Isolation for Maintenance, and complete Isolation Table."
This referenced section contains no prerequisite or warning requiring shutdown margin to be verified prior to isolation.
However, in the same operations manual an unreferenced section titled Control Rod System Failure states in part:
" Control Rod System failures of major significance are those which involve rods which are stuck, uncoupled, inoserable, or have failed to scram. The major consideration wit 1 respect to such failures is to assure that reactivity limits pertaining to shutdown margin, individual rod worths, rod drop accidents and scram requirements will not be exceeded."
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Thus, in following OCD 4201, no warning of this type is provided to the user. There is no written guidance in the 000, which is relied upon as the controlling procedure, to alert the user that SDM concerns are involved with the act of isolating a HCU. The OCD was not prepared in a fashion which would ensure Tech Spec 3.3.A.1 would be satisfied. By following the OCD as it exists in the present form, operation outside the limit of Tech Spec 3.3.A.1 was possible by making one or more rods inoperable in a non fully-inserted position.
Tech Spec 6.5.C.3 states in part:
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" Detailed written procedures, including applicable check-off lists and instructions covering areas listed below shall be
prepared and followed.... The following maintenance and test procedures will be developed to satisfy routine inspection, preventive maintenance programs, and operating license requirements:
preventive or corrective maintenance of plant and systems that could have an effect on nuclear safety."
Preventive Maintenance Procedure OCD 4201 was prepared in a manner whereby its use could result in not satisfying Tech Spec 3.3.A.1 when performing preventive maintenance on one or more hydraulic control units for a control rod. This preventive maintenance procedure did not satisfy the requirement of Tech Spec 6.5.C.3 andisaviolation(263/87003-01(DRP)). During the review of Procedure OCD 4201 and 000 4208, the inspector discovered that two fonnats exist for these OCDs:
Form 3092, Rev. O, which OCDs 4201 and 4208 use; and Form 3092, Rev. 2, which OCD 4026-1 uses. The Rev. 2 form has an approval block for the OCD to commence following the shift supervisor's signature. The inspector observed that 0CDs which involve tech spec requirements could be improved by changing to the Rev. 2 format and, thereby, improving the control of OCD work by a shift supervisor's authorization to commence the work.
Conclusion.- Two methods currently in use to ensure that Tech Spec 3.3.A.1wasmet(i.e.(1)thesystemengineers' writing and review of the WRA, and (2) the Procedure OCD 4201) provided no guidance regarding the requirement to determine adequate SDM prior to making a non fully-inserted control rod inoperable.
c.
Application of Tech Specs Tech Spec 3.3.A.1, Reactivity Margin - Core Loading, states:
"The core loading shall be limited to that which can be made subcritical in the most reactive condition during the operating cycle with the strongest operable control rod in its full-out position and all other operable rods fully inserted."
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The basis for this tech spec states in part:
"The core reactivity limitation is a restriction to be applied principally to the design of new fuel which may be loaded in the core or into a particular refueling pattern. Satisfaction of the limitation can only be demonstrated at the time of loading and must be such that it will apply to the entire subsequent fuel cycle.
The generalized form is that the reactivity of the core loading will be limited so that the core can be made subcritical by at least R + 0.25 percent k at the beginning of the cycle, with the stronlest control rod fully withdrawn and all others fully inserted."
OCD 4201 required that Tech Specs 3.3.A and 3.3.D be met. The site superintendent and shift supervisor (SR0s) discussed whether Tech Spec 3.3.A would be met as they realized that the rod would be unavailable to scram.
The conclusion was that 3.3.A would be met if the " strongest control rod in its full-out position" was Rod 22-07. This conclusion, which was debated, indicates that inadequate guidance was provided to operations regarding this tech spec.
The inspector did not review the adequacy of the training on tech specs. However, the licensee is reevaluating the training of licensed operators and intends to significantly improve training in this area.
During followup of this event, the ins]ector learned that the site superintendents and shift supervisors lad difficulty in establishing clearly what the requirements of Tech Spec 3.3.A.1 meant and chose not to utilize other technical expertise available. The operating shift supervisors had some reservation in the accuracy of their interpretation of Tech Spec. 3.3.A.1.
Conclusion - The interpretation taken by several experienced licensed operators of Tech Spec 3.3.A was non-conservative.
Operations supervisors should be encouraged to consult offsite expertise when the correct course of action is unclear.
4.
Exit Interview The inspector met with licensee representatives denoted in Paragraph 1 at the conclusion of the inspection on February 13, 1987. The inspector discussed the purpose and scope of the inspection and the findings.
The inspector also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspector during the inspection. The licensee did not identify any documents /pr0 Cesses as proprietary.
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