IR 05000315/1987004
| ML20205G772 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 03/17/1987 |
| From: | Burgess B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20205G692 | List: |
| References | |
| 50-315-87-04, 50-315-87-4, 50-316-87-04, 50-316-87-4, NUDOCS 8703310539 | |
| Download: ML20205G772 (13) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Report Nos. 50-315/87004(DRP); 50-316/87004(DRP)
Docket Nos. 50-315; 50-316 Licenses No. DPR-58; DPR-74 Licensee: American Electric Power Service Corporation Indiana and Michigan Electric Company 1 Riverside Plaza Columbus, OH 43216 Facility Name: Donald C. Cook Nuclear Power Plant, Units 1 and 2 Inspection At: Donald C. Cook Site, Bridgman, Michigan Inspection Conducted: January 27, through February 23, 1987 Inspectors:
B. L. Jorgensen J. K.
Heller 3/7!87 Approved By:
u ief Projects Section 2A Dat'e Inspection Summary Inspection on January 27, through February 23, 1987 (Reports No. 50-315/87004(DRP);
50-316/87004(DRP))
Areas Inspected:
Routine unannounced inspection by the resident inspectors of previously identified items; operational safety; maintenance; surveillance; NRC Region III requests; and, Safety Committee activities.
Results: Of the six areas inspected, no violations or deviations were identified in five areas. One violation was identified (Severity Level IV -
visitor escort deficiency - Paragraph 3) in the remaining area and is addressed in Inspection Report No. 315/87008; 316/87008.
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DETAILS 1.
Persons Contacted W. Smith, Jr., Plant Manager
- A. Blind, Assistant Plant Manager, Administration
- J. Rutkowski, Assistant Plant Manager, Production
- L. Gibson, Assistant Plant Manager, Technical Support
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B. Svensson, Licensing Activity Coordinator T. Kriesel, Technical Superintendent, Physical Sciences K. Baker, Operations Superintendent
- E. Morse, Quality Control Superintendent i
- T. Beilman, I&C/ Planning Superintendent
- J. Allard, Maintenance Superintendent
- T. Postlewait, Technical Superintendent, Engineering M. Horvath, Quality Assurance Supervisor R. Clendenning, Radiation Protection Supervisor
- W. Hodge, Security Manager The inspector also contacted a number of other licensee and contract employees and informally interviewed operations, maintenance, and technical personnel.
- Denotes some of the personnel attending Management Interview on
February 25, 1987.
2.
Actions on Previously Identified Items
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a.
(Closed) Unresolved Item (315/83001-08; 316/83001-08): The administrative procedures did not adequately provide for testing i
the following plant modifications or significant procedure changes.
During the time frame of 1983 through 1985 the licensee conducted an administrative procedure review. During that review this unresolved item was addressed.
PMI-2010 " Plant Managers and Department Head Instructions, Procedures and Associated Index," Revision 12 at Paragraph 3.9.6 requires that the originating department walkdown the original issue and each major revision to procedures. The walkdowns are intended to verify nomenclature, equipment type j
and ranges, locations, and the practicability of the procedure.
PMI-5040 " Design Changes," Revision 8 at Paragraph 4.14 specifies
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that functional and operability testing are required to verify that
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the design change will operate as designed. This paragraph assigns
responsibility to the lead engineer to determine functional testing and assigns the design change coordinator (with plant input) the
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responsibility to determine operability testing requirements, b.
(Closed) Unresolved Item (315/83001-13; 316/83001-13): Evaluate applicability of PMI-8010 to operation phase activities, especially with respect to verifying suitability of operating procedures.
'r PMI-8010 was considered applicable to Unit 2 preoperational testing
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and has been cancelled. As shown in the previous unresolved item the licensee has made several programmatic enhancements to PMI-5040 which should resolve this matter, c.
(Closed) Unresolved Item (315/83001-14; 316/83001-14): PMI-5040 does not reflect the design verification requirements of ANSI N45.2.11.
The latest revision of PMI-5040 appears to contain these design verification requirements.
d.
(Closed) Unresolved Item (315/85028-05; 316/85028-05): Determination of the acceptability of the licensee's actions during a period of indicated Quadrant Power Tilt Ratio (QPTR) in excess of 1.02 on January 13 and 14, 1985.
This unresolved item had three attributes.
First, actions by the licensee to reduce nuclear power range high power trip and reset setpoints on January 13, 1985 were in doubt.
This was based on data sheets for restoration on January 14 which apparently showed the "as-found" setpoints not reduced, but at normal values. This interpretation was in error. The data in question are not "as-found" values; rather, they constitute instructions concerning what the "as-left" shall be.
Thus, data from January 13 documenting the reduction of the applicable setpoints are not contradicted by any data from the following day.
"As-found" data were not required on January 14 and none were recorded. The documentation of January 13 (that action requirements were met) stands unrefuted.
Second, a concern was expressed that setpoints were changed utilizing monthly surveillance test procedures (STP's - which accomplish
" channel functional test" requirements) rather than channel calibration procedures.
Procedure PM1-6030 permits use of the STP if a bistable is found out of specification during performance of the STP. This does not mean, however, that it prohibits use of the STP in other circumstances. For the case in question, the licensee processed (proposed, reviewed and approved) a specific Change Sheet to each of four STP's (one per instruaent channel) specifically authorizing their use for the purpose of adjusting bistable setpoints.
On January 13 this was accomplished by Change Sheet CS-2, which provided the specific new setpoints established in the instant case.
On January 14, Change Sheet CS-3 cancelled CS-2 and established a generic reference to the applicable Technical Specification for determining setpoint changes on a case basis as required by the Technical Specification.
This change has since become permanent.
The inspector examined the channel functional test procedures (**2 THP 4030 STP.127 through STP.130) and the calibration procedure (**2 THP 6030 IMP.231) during this inspection and concluded they all provide adequate instructions and documentation for adjustment of bistable setpoints.
Third, there was a question as to whether the cause of the indicated flux tilt was properly understood and corrected. The licensee ascribed the cause to " drift" on the part of the N-41 channel lower l
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detector. During this inspection, the licensee's corrective and preventive action documentation (Condition Report 2-01-85-0081)
was examined. The following facts are cogent:
(1) An incore flux map was taken which showed no actual tilt; this was verified by no Delta-T tilt.
(2) Both the amount and rate of change of detector N-41 behavior were quantified from historical data.
(3) A correction factor to reconcile N-41 output with known flux conditions from the flux map was developed and applied using methodology developed for beginning of cycle calibration factors.
(4)
Indicated flux tilt from the excore detectors then matched tilt determined from the flux map.
The licensee resumed power escalation following the above, but continued twice daily hand calculations (in addition to computerized monitoring) pending full calibration of all four excore channels; an action completed within the following week. The inspector concluded the licensee did understand the cause of the indicated tilt and used a valid mears to correct the errant indication on a temporary basis followed by more permanent (e.g. calibration is considered good for 90 days) resolution.
e.
(Closed) Open Item (315/86004-03): The Containment Spray System Engineering Safety Features Response Time is 45 seconds or less, however the containment spray pump outlet valves have a stroke time of 60 seconds. The licensee was asked to verify that the outlet valves will be able to perform their intended safety function after 45 seconds. The licensee provided the inspector with an engineering evaluation showing the valves will provide negligible flow resistance after 45 seconds.
No violations, deviations, unresolved or open items were identified.
3.
Operational Safety Verification During the inspection period, the inspector observed control room operation, manning, shift turnover, approved procedures and Limiting l
Condition for Operation (LCO) adherence, and also reviewed applicable
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logs and conducted discussions with control room operators. Observations of the control room monitors, indicators, and recorders were made to verify the operability of emergency systems, radiation monitoring l
systems, and nuclear and reactor protection systems, as applicable.
Reviews of surveillance, equipment condition, and tagout logs were
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conducted.
Proper return to service of selected components was verified.
Tours of the auxiliary building, turbine building, and screenhouse were made to observe accessible equipment conditions, including fluid leaks,
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potential fire hazards, and control of activities in progress. During discussions with operations personnel, and while observing operations personnel in conduct of activities, the inspector evaluated their knowledge and capabilities as a reflection of their training.
a.
The inspector reviewed selected plant operating procedures as follows:
(1) 2 OHP 4021.056.001 " Filling and Venting Auxiliary Feed Water System and Placing System in Standby Recdiness," Rev. 6, 6/20/86 through Change Sheet No. 2, 2/11/86.
(2) 2 OHP 4021.056.002 " Operation of the Auxiliary Feed Pumps During Plant Startup and Shutdown," Rev. 6, 11/6/86.
b.
Selected alarm response procedures were reviewed, using the controlled copies of such procedures maintained in the respective
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Unit control rooms.
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(1)
1-0HP 4024.113 " Annunciator No. 13 Response - Steam Generators 1 and 2."
(2)
1-OHP 4024.114 " Annunciator No. 14 Response - Steam Generators 3 and 4."
(3) 2-CHP 4024.213 " Annunciator No. 13 Response - Steam Generators 1 and 2" (Unit 2).
The inspector found instructions in both Items 1 and 3 above, concerning alarm window number 009, "TDFP Locked Out," which stated an erroneous mechanical overspeed trip setpoint.
Both identified the setpoint for the lockout at 25% overspeed (i.e. 125% rated).
The trip is now set, however, at 111% plus/minus 1% rated. The adjustments were made during respective Unit 1 and Unit 2 outages in late 1985 and mid-1986, apparently by implementation of special procedures through Job Orders. On questioning, two of three control room operators remembered the setpoint as 125% rated, while a third was uncertain but thought it to be "around 112%." This situation l
is discussed further in Paragraph 5 below, " Surveillance," which i
addresses review of the special procedures, c.
The inspector walked down the Unit 1 East motor driven and the turbine driven auxiliary feedwater pump lineups using valve lineup sheet No. I to 1-OHP 4030 STP.017E " East Motor Driven Auxiliary Feedwater System Test" and STP.017T " Turbine Driven Auxiliary Feedwater System Test" (See also Paragraph 5, Surveillance). No significant problems or operability questions were identified, nor were any conditions noted which indicated any components were materially degraded.
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During the walkdown, the inspector found Valve 1-FW-146 chained and locked open. Valve 1-FW-146 "TDAFP Cooling Water Supply" is listed as simply open.
It appears that a 1986 procedure change revised the valve position of two valves down stream of 1-FW-146 from open to locked open, but Valve 1-FW-146 was not addressed. The addition of a chain and lock to 1-FW-146 does not appear to hinder operability, however the valve lineup sheet should be changed to reflect plant conditions or the lock and chain removed, whichever is appropriate.
This is an open item pending licensee action on this item (0 pen Item 315/87004-01).
d.
During a review of 1-0HP-4023E-0 " Reactor Trip or ' Safety Injection" the inspector found that 1-CCR-441 "CCW Fm Pen Cool" was listed as a Phase A isolation valve. A review of logic prints revealed that 1-CCR-441 does not get a phase A isolation signal. This was identified to the Operations Procedure Coordinator; a temporary procedure change was issued to delete the valve from the Phase A isolation list.
e.
The inspector observed an activity which appeared to be in violation of the Modified Amended Security Plan. Discussion of this subject matter is detailed in Inspection Report No. 315/87008; 316/87008.
f.
The inspector assisted NRC Region III specialists in evaluation of an exercise of the licensee's Emergency Plan conducted on February 10, 1987. These matters were documented separately in IE Inspection Report No. 050-315/87006(DRSS); 050-316/87006(DRSS).
One violation and one open item, and no deviations or unresolved items were identified.
4.
Maintenance Station maintenance activities of safety related systems and components listed below were reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides and industry codes or standards and in conformance with Technical Specifications.
The following items were considered during this review: the Limiting Conditions for Operation were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures; and post maintenance testing was performed as applicable. During observation of activities in progress, the inspector evaluated employee training in the maintenance area as reflected by performance knowledge and capabilities.
The following maintenance activities were reviewed or observed.
c.
Observed and Reviewed (1) JO 15309 (File:PS35)
Repair leak on piping upstream of 2-CCW-184W (CCW to "W" CCP shutoff valve).
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(2) JO 12282 (File:PS35)
Repair leak on CCW line to 2W RHR Hx.
The inspector verified that the licensee was in compliance with the maintenance / weld procedure, post maintenance hydrostatic test procedure and the freeze seal procedure (if applicable).
The repairs were among several involving leaks in Unit 2 CCW piping which are of unknown cause and which involved removal / replacement of piping sections. The licensee has sent the removed sections to the lab for analysis and has radiographed additional selected sections of the piping.
To date the licensee has not identified the reasons for the leaks nor identified any additional leaks. The licensee investigation is continuing. This item is an open item pending completion of the licensee investigation (0 pen Item 316/87004-01).
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b.
Observed JO 00462 (File: performance)
Helium leak test of new vent piping located in the CVCS pipe tunnel, c.
Reviewed (2) J0 702335 MRV-242 (Steam piston Trn B (File: ME-VRV-2-MRV-242)
Dump Valve for No. 4 Steam Stop) seat is leaking by.
(2) JO 015363 Set stroke of MRV-242 after
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(File: IN-C&I-2-M200)
repair of JO 702335 These Job Orders were reviewed because logs for 12/23/86 could be interpreted to suggest the licensee attempted to do eight hours of work in a four hour LC0 and accomplished this by entering, exiting and re-entering the LCO time clock. The inspector discussed this item with the Maintenance Superintendent and found that all work was done within the LC0 time clock and that no manipulating of the LC0 time clock occurred.
(3) J0 85571 Repair the IE CCW pump (File: ME-PP-PP-10)
because parts of the suction strainer damaged the rotating element.
This J0 was reviewed as a followup to Violation 315/87003-03.
That violation addressed an error in the surveillance procedure acceptance criteria which was derived from the manufacturer's pump performance curve. The inspector was concerned that the licensee may not have revalidated the pump curve following replacement of the impeller.
The inspector found that the licensee had verified the pump performance curve following replacement of the rotating element.
One open item, and no violations, deviations, or unresolved items were identified.
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5.
Surveillance The inspector reviewed Technical Specifications required surveillance testing as described below and verified that testing was performed in accordance with adequate procedures, that test instrumentation was calibrated, that Limiting Conditions for Operation were met, that removal and restoration of the affected components were properly accomplished, that test results conformed with Technical Specifications and procedure requirements and were reviewed by personnel other than the individual directing the test, and that deficiencies identified during the testing were properly reviewed and resolved by appropriate management personnel.
During observation of activities in progress, the inspector evaluated employee training in the surveillance area as reflected by performance knowledge and capabilities.
The following were observed and/or reviewed, with findings and licensee actions, if applicable, as stated.
a.
- 2 OHP 4030 STP.017E " East Motor Driven Auxiliary Feedwater System Test," Rev. 3, 5/26/86 through Change Sheet No. 4, 10/2/86, b.
- 2 OHP 4030 STP.017R " Auxiliary Feedwater Pump Response Time,"
Rev. 2, 11/13/86.
c.
- 12 THP 4030 STP.218 " Automatic Operation of Auxiliary Feedwater Pumps," Rev. 9, 1/5/87.
d.
- 2 OHP 4030 STP.018 " Steam Generator Stop Valve Dump Valve Surveillance Test."
e.
- 2 THP 4030 STP.127 " Power Range Nuclear Instrumentation Protection Set I (N-41) Surveillance Test (Monthly)," Rev. 6, 6/2/86.
The inspector also briefly reviewed STPs.128,.129 and.130 and results from their implementation during the period January 13 and 14, 1985, as discussed in Paragraph 2.d above.
f.
- 2 THP 6030 IMP.231 " Power Range Nuclear Instrumentation Calibration N-41, N-42, N-43, N-44," Rev. 6, 6/9/86.
g.
- 12 THP 6040 PER.001 " Centrifugal Pump Performance Tests."
This generic procedure was used for establishing performance capabilities of the Unit 2 diesel-driven fire pump following a pump overhaul and prior to its return to service.
h.
The inspector found no procedure for routine verification of the TDAFP overspeed trip settings and operation, which might have been contained in STP.017TV. On further questioning, the licensee produced
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one "special procedure" (**2-MHP-SP-116) which wcs used for testing the Unit 2 TDAFP mechanical and electronic overspeed trip functions following turbine driver overhaul in 1986, and another "special procedure" (**1 MHP-SP-103) which was used to reset and test the mechanical overspeed trip on the Unit 1 pump in late 1985. Review of the special procedures is discussed further in Paragraph k. below.
1.
(1)
- 1 OHP 4030 STP.017E " East Motor Driven Auxiliary Feedwater System Test."
(2)
- 1 OHP 4030 STP.017T " Turbine Driven Auxiliary Feedwater System Test."
For this item the inspector randomly selected 12 Surveillance Tests completed in 1986 and verified that the completed tests complied with the test instructions; no problems were identified.
J.
- 2 OHP 4030 STP.017T " Turbine Driven Auxiliary Feedwater System Test," Rev. 4, 8/28/86 through Change Sheet No. 1, 9/11/86.
This procedure refers operators to the Technical Data Book (a controlled document) for selected criteria and parameters, including selection of a discharge pressure gauge within a range specified in Figure 15.2.
The acceptance range given in Figure 15.2 at the time of inspection was 0-4000 psig. This was slightly excessive, since the 1983 revision to the ASME Code (vice the 1974 edition) reduced the acceptance range from four times reference value to three times reference, and the reference value for the Unit 2 turbine driven pump is 1320 psig.
The inspector identified the discrepancy to the Performance Department and a Condition Report was generated which led to revision of the Technical Data Book acceptance range to 0-3500 psig effective February 4, 1987. The gauge actually used for the subject test was 0-2000 psig, well within Code allowable.
k.
(1) **1-MHP-SP-103 " Unit 1 Turbine Driven Auxiliary Feed Pump Mechanical Overspeed Trip Test," Rev. O, 10/23/85 through Change Sheet No. 4, 11/5/85.
(2)
- 2 MPH-SP-116 " Unit 2 Turbine Driven Auxiliary Feed Pump Mechanical and Electronic Overspeed Trip Tests," Rev. O, 6/5/86 through Change Sheet No. 3, 6/26/86.
(3)
Rev. O, 8/15/85 though Change Sheet 1, 6/5/86.
SP-103 reduced (from 125%) and verified the mechanical turbine overspeed trip device at 111% plus/minus 1% rated speed. With rated speed at 4,350 rpm, the net effect was to reduce the trip setpoint from about 5,438 rpm to 4785-4872 rpm.
Per the pump manufacturer, maximum turbine and pump speed is 4,900 rpm.
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SP-116 reset and verified both the mechanical and electronic overspeed trip protection following a turbine driver overhaul in June 1986. An overspeed test device supplied by the speed governor manufacturer (Woodward) was used for the first time in testing the mechanical trip device. The mechanical trip setting acceptance criteria (4785-4872 rpm) were the same as those for the Unit 1 turbine driver (SP-103, above), however, the " caution" statements repeatedly state 4872 rpm as the limit, rather than 4900 rpm as the Unit 1 procedure did. The acceptance criteria for the electronic overspeed trip are 106%-110%, or 4611-4785 rpm.
IMP.074 verifies that speed indications on three instruments (at the pump, in the Unit control room and at the Unit hot shutdown
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panel opposite the Unit control room) are accurate plus/minus 100 rpm (approximately 1.5% full scale) and checks / calibrates the trip setpoint against an acceptance criteria of 4740 plus/minus 100 rpm; i.e. 4640-4840 rpm.
This is slightly inconsistent with SP-116, for which IMP.074 was a prerequisite.
As evident from the above, consistent TDAFP overspeed trip settings were not established for a period in late 1985 through mid 1986.
The inspector verified proper prior safety review and approval of each procedure. As noted in Paragraph 3 of this report, plant operators had not been informed ar.d alarm response procedures revised (as of February 1986) on the revised trip setpoints.
Further, the Operations Department controlled copy of the Auxiliary
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Feedwater System Description was outdated with respect to both the electronic and the mechanical overspeed trip setpoints.
This may be indicative of a gap in the licensee's administrative processes for handling setpoint changes, and further review of these processes is planned.
This is considered an unresolved item pending such review (Unresolved Item 315/87004-02; 316/87004-02).
One unresolved item, and no violations, deviations or open items were identified.
6.
NRC Region III Requests The Region III office requested that "Open Item" tracking numbers be assigned to two questions raised by an NRC-sponsored survey for the control room ventilation systems, and that the items be further reviewed.
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Numbers were assigned and the items were reviewed and closed as described below:
a.
(Closed) Open Item 315/87004-03; 316/87004-03: Adequacy of evaluations under 10 CFR 50.59 to support / demonstrate acceptability of adjustments made in control room ventilation system dampers after submittal of analyses to NRC in 1981. This question was raised during an NRC-sponsored survey to evaluate control room ventilation systems'
capabilities conducted at the D. C. Cook plant during September 15-19, 1986. Two items that needed to be addressed were whether the
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licensee performed analyses to determine the acceptability of these adjustments and, second, whether such analyses demonstrated, in a technically correct manner, that compliance to safety requirements were maintained.
The licensee did perform analyses of the effects of specific adjustments in that the two specific cases questioned by the NRC survey team are documented with the licensee's Licensee Event Report (LER) 315/85007, which discusses system adjustments in response to adverse test findings on February 22 and again on March 10, 1985.
Subsequently, also as documented with LER 315/85007, the licensee developed a matrix which can be used (each time control room envelope filtered and unfiltered inleakage are determined by periodic testing) to establish whether this combination of measured values remains in compliance with limits.
This approach is specifically acknowledged in the NRC survey teams consultant's report as a good concept for dealing with these interdependent
" variables."
The technichl adequacy of the licensee's analyses was examined for two attributes. The first involves mathematical assumptions, techniques and accuracy. No attempt was made during this inspection to independently calculate various dose consequences. The inspector verified only that the licensee utilized standard nethodology referenced by NUREG-0737 and submitted the dose envelope matrix to independent review (by Westinghouse Electric Corporation) with acceptable results. The second attribute examined for was whether safety criteria was met as prescribed by 10 CFR 50.59.
This means assurance that previously-analyzed accidents (as documented in the Safety Analysis Report) were not worsened, and that the margin of safety in the basis of the applicable technical specifications were not reduced. The licensee's approach was satisfactory on both counts because the Safety Analysis Report, the Technical Specifications basis, and the specific and generic (matrixed) analyses previously discussed all uniformly utilized 10 CFR 50 Appendix A General Design Criterion 19 as the safety limit.
The inspector has no further questions on this matter at this time, b.
(Closed) Open Item (315/87004-04; 316/87004-04): Training manuals /
instructions contained out-of-date information (March 1986) concerning the control room heating, ventilation and air conditioning (HVAC)
system. This item was also identified during the September 1986 NRC-sponsored control room HVAC survey. The inspector reviewed the licensee's mechanisms for updating training materials, and reviewed the current materials used for providing training on the control room HVAC systems. The licensee utilizes " lesson plans" for systems (and other) training. The lesson plans for the control room HVAC training are scheduled and performed on an as-needed basis.
Lesson Plan R0-C-AS9 " Auxiliary Building and Control Room Ventilation" dated March 1986, was reviewed. One Change Sheet (consisting of 14 individual change items) was processed into the lesson plan in
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October 1986. Among the items addressed are: clarification that certain stated flow values for makeup and recirculation are approximations, not absolutes; reference to the appropriate dose assessment calculation procedure concerning operation in the recirculation mode; additions to some descriptions and corrections to some details (e.g. location of chl'orine detectors, function of toilet exhaust damper, position indication and control); and miscellaneous enhancements.
Responsibility is clearly assigned to the lead instructor to assure all known, needed updates are incorporated into as-needed lesson plans prior to each use. The inspector has no further questions on this matter at this time.
c.
By memorandum dated January 30, 1987, NRC Region III requested that the resident inspectors review the licensee's actions taken for I.E.
Information Notice 86-106 "Feedwater Line Break" and complete a questionnaire on the licensee's program. The majority of the questions were addressed in the write-up for Inspection Report 050-315/87003(DRP); 050-316/87003(DRP). The additional questions pertained to the attributes used to determine which piping sections would be examined for wall thinning. The attributes were obtained from the licensee and transmitted to Region III on February 3,1987.
Two open items were identified, reviewed and closed.
No violations, deviations or unresolved items were identified.
7.
Safety Committee Activities During this inspection, a review was performed of licensee activities pursuant to licensee procedure PMI-1040 " Plant Nuclear Safety Review Committee," Rev. 4, 4/18/86 through Change Sheet 3, 11/26/86.
Several sections, including at least 4.5.2, 4.5.3 and 4.9.2, reflected membership responsibilities via titles which became obsolete with the site reorganization of July 1,1986. On further discussion with the licensee, the inspector found that Revision 5, which the licensee had processed but not yet issued, updated titles as appropriate.
Compliance to selected procedural stipulations was verified. This included annual training of PNSRC members in committee responsibilities and authorities and the requirements of Federal Regulations concerning safety evaluations pursuant to 10 CFR 50.59.
Committee receipt and evaluation of an annual summary report on overall facility training status was verified for continued need and validity, as was the annual
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j review of committee-originated technical specification " clarifications."
The procedure establishes several subcommittees and designates their operational controls and responsibilities.
Requirements associated with subcommittee reviews, recommendations to the full committee, and l
documentation and distribution of subcommittee meeting minutes, were l
selectively verified.
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The inspection did not encompass a programmatic review of Committee activities against basic quality assurance requirements.
Such a review may be performed by NRC quality-specialist inspectors at a future date.
?!o violations, deviations, open or unresolved items were identified.
8.
Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, violations, or deviations. An unresolved item disclosed during the inspection is discussed in Paragraph 5.k.
9.
Open Items Open items are matters which have been discussed with the licensee, which will be reviewed further by the inspector, and which involve some action
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on the part of the NRC or licensee or both. Open items disclosed auring the inspection are discussed in Paragraphs 3.c and 4.a.
10. Management Interview
The inspectors met with licensee representatives (denoted in Paragraph 1)
on February 25, 1987 to discuss the scope and findings of the inspection.
In addition, the inspector asked those in attendance whether they considered any of the items discussed to contain information exempt from disclosure as proprietary.
No items were identified.
a.
An apparent violation of security requirements was discussed (Paragraph 3.e).
b.
Each open and unresolved item identified during the inspection was specifically identified to the licensee representatives (Paragraphs 3.c, 4.a, 5.k, and 6.).
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