IR 05000416/1986026
| ML20215L998 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 10/15/1986 |
| From: | Butcher R, Dance H, Will Smith NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20215L801 | List: |
| References | |
| 50-416-86-26, 50-417-86-04, 50-417-86-4, NUDOCS 8610290429 | |
| Download: ML20215L998 (14) | |
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UNITED STATES
[p3 f trug'o NUCLEAR REGULATORY COMMISSION y"
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REGION 11
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101 MARIETTA STREET.N.W.
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ATLANTA, GEORGI A 30323
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Report No.:
50-416/86-26 and 50-417/86-04 Licensee:
Mississippi Power and Light Company Jackson, MS 39205 Docket No.:
50-416 and 50-417 License No.: NPF-29 and CPPR-119 Facility Name: Grand Gulf Nuclear Station, Units 1 and 2 Inspection Conducted: August 9 through September 12, 1986
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Inspectors: ;
R. C. Butcher, S,enior Resident Inspector Date Signed
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W. F. Smith,@nt' Inspector Date Signed Accompanying Inspector:
L. P. Modenos, Project Engineer (August 25-28,1986)
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H. C. Dance, Section Chief Date Si'gned Division of Reactor Projects SUMMARY Scope:
This routine inspection was conducted by the resident inspectors at the site and on Unit 1 in the areas of Licensee Action on Previous Enforcement Matters, Operational Safety Verification, Maintenance Observation, Surveillance Observation, ESF System Walkdown, Reportable Occurrences, Operating Reactor Events, Inspector Followup and Unresolved Items, Preparation for Refueling and Refueling Activities, and for Unit 2, Review of Quality Assurance for Extended Construction Delay.
Results: One deviation was identified: Failure to initiate quarterly Engineered Safety Features electrical switchgear room cooler testing as committed.
8610290429 861020 DR ADOCK 05000416 PDR
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i REPORT DETAILS 1.
Licensee Employees Contacted
- J. E. Cross, GGNS Site Director
- C. R. Hutchinson, GGNS General Manager
- R. F. Rogers, Manager, Unit 1 Projects A. S. McCurdy, Manager, Plant Operations
- J. D. Bailey, Compliance Coordinator M. J. Wright, Manager, Plant Support
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- L. F. Daughtery, Compliance Superintendent D. G. Cupstid, Start-up Supervisor R. H. McAnulty, Electrical Superintendent
- R. V. Moomaw, Manager, Plant Maintenance W. P. Harris, Compliance Coordinator
- J. L. Robertson, Licensing Superintendent L. G. Temple, I & C Superintendent J. H. Mueller, Mechanical Superintendent L. B. Moulder, Operations Superintendent J. V. Parrish, Chemistry / Radiation Control Superintendent Other licensee employees contacted included technicians, operators, security force members, and office personnel.
- Attended exit interview 2.
Exit Interview The inspection scope and findings were summari:ed on September 12, 1986, with those persons indicated in paragraph 1 above.
The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspectors during this inspection.
The licensee had no comment on the following inspection findings:
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a.
416/86-26-01, Inspector Followup Item.
Discrepancies between System
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Operating Instructions, Piping and Instrument Drawings (P& ids) and the actual plant configuration.
Plant cleanliness observations were
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discussed.
(paragraph 7)
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416/86-26-02, Deviation.
Failure to initiate quarterly Engineered Safety Features electrical switchgear rcom cooler testing as committed.
(paragraph 9)
c.
416/86-26-03, Inspector Followup Item.
Commitment to incorporate all
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FX series valves on P& ids. (paragraph 7)
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d.
416/86-26-04, Inspector Followup Item.
Potential for loss of Unit 1 safety systems due to control room water leak.
(paragraph 4)
3.
Licensee Action on Previous Enforcement Matters (92702)
(Closed) Violation 416/86-08-03.
The licensee has revised Surveillance Procedure 06-0P-1P41-Q-0005, Standby Service Water System Valve and Pump
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Operability Test, System Operating Instruction 04-1-01-P41-1, and Standby
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Service Water System to require the standby service water B loop
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recirculation isolation valve P41-F002B to be approximately 48% open (flow of 9000,+500,-0 gpm) and locked in position. Also, the licensee verified the minimum flow rate to be 9,500 gpm.
Personnel involved in reviewing design change packages to assure affected procedures are changed were cautioned on the importance of ensuring all requirements are met.
(Closed)
Violation 416/86-04-03.
Failure to promptly correct the installation of non-seismic qualified relay.
Plant Procurement Procedure 01-S-09-1, has been revised to include a review by the Material Engineering group of all purchase requisitions involving components used in safety-related systems. Design change package 84/3516 replaced the non-seismic relays with qualified Agastat ETR relays. Corrective Action Reports 764 and 771 involving equipment procurement of 600 purchase orders have been reviewed and completed to assure that procurement of properly qualified equipment had been made.
4.
Operational Safety Ve.ification (71707)
The inspectors kept themselves informed on a daily basis of the overall plant status and any significant safety matters related to plant operations.
Daily discussions were held with plant management and various members of the plant operating staff.
The inspectors made frequent visits to the control room such that it was visited at least daily when an inspector was on site. Observations included instrument readings, setpoints and recordings status of operating systems, tags and clearances cn equipment controls and switches, annunciator alarms, adherence to limiting conditions for operation, temporary alterations in
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effect, daily journals and data sheet entries, control room manning, and
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access controls.
This inspection activity included numerous informal discussions with operators and their supervisors.
General plant tours were conducted on ct least a biweekly basis. Portions of the control building, turbine building, auxiliary building and outside i
areas were visited.
Observations included safety related tagout verifications, shift turnover, sampling program, housekeeping and general plarit conditions, fire protection equipment, control of activities in progress, radiation protection controls, physical security, problem identification systems, and containment isolation.
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The following comments were noted:
a.
On August 26, 1986, the inspector noted a liquid that appeared to be light oil dripping on Division I cable run IAATMG19 on the 93 foot elevation in the auxiliary building (area 7). The liquid collected in a puddle on an insulated duct below the floor grating.
The shift supervisor was notified; however, through September 11, 1986, the same
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condition existed.
This was discussed with plant management.
b.
On September 4, 1986, door 1A401, entry to the auxiliary building from the turbine building on elevation 166, was found not in the shut position.
The latch mechanism would not permit the door to fully close.
The Shift Supervicor (SS) was notified and since door 1A401 is part of secondary containment, the SS secured the door in the shut position and made the door inoperable for entry / exit of the auxiliary building, c.
On September 4, 1986, the licensee discovered a water leak in the sub-floor within the control room envelope. The apparent cause was a water leak from a temporary room cooler located at elevation 189' above the Unit 2 control room.
The water reportedly seeped from the Unit 2 control room side to the Unit I control room. The water also leaked from the control room sub-floor down into the computer room below and damaged the Balance of Plant Computer.
The inspectors expressed concern that such a potential leakage interface between Unit 2 and Unit I could have safety implications and should be thoroughly investigated and corrected as appropriate.
Also, the inspectors questioned the licensee if this same problem exists on the Unit I control room side.
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The licensee is investigating.
This shall be Inspector Followup Item 416/86-26-04,
l No violations or deviations were identified.
5.
Maintenance Observation (62703)
During the report period, the inspector observed portions of the maintenance activities listed belcw. The observations included a review of the work documents for adequacy, adherence to procedure, proper tagouts, adherence to technical specifications, radiological controls, observation of all or part
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of the actual work and/or retesting in progress, specified fetest requirements, I
and adherence to the appropriate quality controls.
MWO M64859, Install submersible pump and pump down spent fuel pool.
MWO I65189, Troubleshoot and clean standby service water system cooling lines to ESF equipment room fan-coil units.
(see paragraph 9).
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4 MWO M63296, Functional Testing of Mechanical Snubber 10702.
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No violations or deviations were identified.
6.
Surveillance Observation (61726)
The inspector observed the performance of portions of the surveillances listed below.
The observation included a review of the procedure for technical adequacy, conformance to tecnnical specifications, verification of test instrument calibration, observation of all or part of the actual surveillances, removal from service and return to service of the systen or components affected, and review of the data for acceptability based upon the acceptance criteria.
06-IC-IC51-V-0003, Revision 25, Source Range Monitor Channel Calibration 06-EL-1R21-M-0001, Revision 22, 4.16 KV Degraded Voltage Functional Test and
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06-ME-1M61-V-0001, Revision 27, Local Leak Rate Test of Valves G41-F004 and G41-F201 MWO EL3407, Check emergency lights in the control building No violations or deviations were identified.
7.
Engineered Safety Features System Walkdown (71710)
A complete walkdown was conducted on the accessible portions of the Division II Emergency Diesel Generat.r. The walkdown consisted of an inspection and verification, where possible, of the required system valve alignment, including valve power available and valve locking where required, instrumentation valved in and functioning; electrical and instrumentation cabinets free from debris, loose materials, jumpers and evidence of rodents, and system free from other degrading conditions.
The results of the above inspection were satisfactory in terms of the apparent operability of the unit; however, the inspector identified the following discrepancies.
Each item was discussed with a member of the i
licensee's staff.
a.
F061A is an instrument root valve for pressure switch PSL N122B; however, it is improperly labeled as a drain.
The valve does not appear on Piping and Instrument Drawing (P&ID) M-1070B, Revision 16, nor does it appear in the valve lineup attachments in System Operating i
Instruction (SOI) 04-1-01-P75-1, Revision 2.
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b.
On P&ID M1070D, valves F033E and F033F, starting air drip leg drains, appear to be in reverse of actual installation.
F033E is installed where the P&ID shows F033F, and F033F is installed where the P&ID shows F033E.
P&ID M1070D shows valve FG73B, discharge header drain, installed upstream of air inlet valve F072G when in fact it is installed upstream of air inlet valve F072H. The licensee's representative indicated that this is t robably a case of improper labeling, rather than P&ID errors.
c.
The differential pressure instrument piping for the fuel oil booster pump suction strainer has a root valve on each leg; however, P&ID M-10708 does not show both valves. One valve, FX136 appears on the P&ID, and on the S0I valve lineup checksheet.
The other valve has no label, it does not appear on the P&ID nor does a valve meeting this description appear in the SOI valve lineup checksheet.
The licensee has previously identified this on material nonconformance report 0540-86 for correction.
The inspector will follow up.
d.
The following valves appear in the SOI valve lineup checksheets, but do not appear on the P&ID: FX201, FX203, FX205, FX211, FX214, FX215, and FX216.
e.
Attachment III (electrical lineup checksheet) of S0I 04-1-01-P75-1 contains errors with respect to component descriptions.
For example, on Attachment III K, breakers 72-11825 and CB1-2 refer to train A when it should be train B; and panel P400 when it should be P401. The inspector also identified to the licensee the differences between the component descriptions listed in the electrical lineup checksheets and the descriptions appearing on the actual breaker label.
For exa.nple, breaker 52-1P66123 is labeled, Control Room PGCC Panel 1H13-P872, but attachment III K in the SGI lists the component description as, Diesel Generator Room Outside Air Temperature Switch.
This could cause confusion during the performance of switch lineups, though both are technically correct.
f.
The general cleanliness and condition of equipment and the equipment spaces appeared to have deteriorated.
For example, there was a piece of wood and a considerable amount of oil-laden dirt and dust between the diesel foundation and the jacket water heat exchange, and other areas not within easy access. A metal bucket was stowed on top of one of the terminal boxes.
The diesel-driven air compressor exhaust insulation was falling apart.
Absorbent material was spread out in various locations to catch oil leaks.
When soaked with oil, the material is a fire hazard. The panel door seal on the back of local control panel IH22-P401 was deteriorated. An unused, energized drop light was left hanging on the side of the generator casing. The handle on valve F0600 was being used as a hanger for a pair of hose adapters tied together with a rag. The bottom of safety related 120 volt AC panels were being used as a repository for Tyraps and other debri.
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There have been previous discussions between the NRC Resident Inspectors and plant management in this regard.
Panel 14P12 had two screws lying in the bottom (along with some Tyraps), which could easily fall into the power transformer breaker area below 3/4".
Division II conduit IDRM2YE had only one support in its approximately 30 feet length, and it was loose. The conduit contacted the "T" handle on an instrument root valve labeled F061A which is a small tubing valve connected to starting air storage tank A001D.
g.
The inspector noted that the motor driven air start compressor was running about 25% of the time during this inspection (18 minutes on, 57 minutes off), although there were no apparent air demands.
This appeared to be an excessive duty cycle indicative of leaks which should be repaired.
Correction of the above deficiencies shall be tracked as Inspector Followup Item 416/86-26-01.
The question of whether FX series valves should appear on the P&ID drawings was raised by the inspectors.
The licensee has stated that their policy will be to include all FX series valves on the P& ids. The licensee will commit to a date to have all FX valves shown on P& ids following the end of this refueling outage. This will be Inspector Followup Item 416/86-26-03.
No violations or deviations were identified.
8.
Reportable Occurrenpes (90712 & 92700)
The below listed event reports were reviewed to determine if the information provided met the NRC reporting requirements.
The determination included adequacy of event description and corrective action taken or planned, existence of potential generic problems and the relative safety significance of each event.
Additional inplant reviews and discussions with plant personnel as appropriate were conducted for the reports indicated by an asterisk. The event reports were reviewed using the guidance of the general policy and procedure for NRC enforcement actions.
The following License Event Reports (LERs) are closed.
LER No.
Event Date Event l
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- 86-013 April 21, 1986 Deficiencies cause unmonitored control room air influent.
- 86-020 June 3, 1986 Nonqualified relay could cause loss of SGTS function.
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- 86-025 July 25, 1986 Inadvertent opening of breaker causes reactor scram.
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The event of LER 86-013 was discussed in IE Report 50-416/86-11 and violation 416/86-11-02.
The event of LER 86-020 was discussed in IE Report 50-416/86-17 and is inspector follow item 416/86-17-05.
The event of LER 86-025 was discussed in IE Report 50-416/86-21.
(Closed) 10 CFR Part 21, P2184-03,(PRD 84-05) Main Feedwater Line Moment Guide Temperature.
The licensee issued and implemented Design Change Package 83/4106 to insure a steel-concrete interface temperature below 200 F for both the main steam and main feedwater line moment guides. This item is closed.
No violations or deviations were identified.
9.
Operating Reactor Events (93702)
The inspectors reviewed activities associated with the below listed reactor events. The review included determination of cause, safety significance, performance of personnel and systems, and corrective action. The inspectors examined instrument recordings, computer printouts, operations journal entries, scram reports and had discussions with operations maintenance ard engineering support personnel as appropriate.
Scram No. 42 On August 25, 1986, the plant was at 78.8% reactor thermal power when at 3:18 p.m. the reactor tripped on a turbine control valve fast closure due to a load reject signal.
All systems operated normally and no engineered safety features actuated. Investigation ty the licensee indicated a Nuclear Plant Engineering (NPE) engineer had opened the load reject relay (R1W80A)
cover plate to attempt to remove an electric schematic that was taped to the inner surface. The schematic was located near the normally open contacts in relay R1W80A which sends a fast closure signal to the turbine control valves when the contacts close. These lock out relays normally actuate approximately 15 seconds after a generator reverse power is sensed.
The cause of the reactor trip is concluded to be the inadvertent closure of the load reject relay by personnel entry into the relay cabinet.
On August 27, 1986, while the unit was still in hot shutdown (condition 3)
following the scram on August 25, 1985, (Scram No.42) the licensee determined that the flow coefficient used to calculate Standby Service Water (SSW) flow to the ESF switchgear room coolers was in error. The flow coefficient used was 0.9779 versus the required flow coefficient of 0.5710. This error was in the nonconservative direction and would yield higher than actuel SSW flow values to the ESF switchgear room coolers. On August 28, 1986, the licensee determined by actual flow measurement that the A Train SSW flow was inadequate
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to maintain certain ESF switchgear rooms below the temperature limits of 140 F which is close to the maximum temperature qualification of some of the switchgear room equipment.
Final Safety Analysis Report (FSAR) para-graph 9.4.5.2.5, ESF Electrical Switchgear Rooms, states that the fan coil units will maintain the room temperatures at less than 104 F during all modes of emergency plant operation.
In IE Report 85-45 a deviation was issued due to the failure of the licensee to develop and implement a program to inspect and test the ESF switchgear room coolers as committed in para-graph 9.4.5.4 of the FSAR.
Report 85-45 was issued on January 15, 1986.
The licensee committed on February 14, 1986, in their response to Deviation 416/86-45-02 to develop a general maintenance instruction which establi.shes the inspection and testing program for the ESF room coolers. The inspectors documented in IE Report 416/86-08, dated April 23, 1986, that the licensee committed to perform the ESF switchgear room cooler tests on a quarterly basis until such time as sufficient data had been collected to determine the correct periodicity. The inspector found that as of August 29, 1986, no quarterly testing of the ESF switchgear room coolers had been accomplished.
Failure to conduct ESF switchgear room cooler tests as committed is a deviation (416/86-26-02).
On August 30, 1986, the licensee notified the inspectors that cleaning of the SSW piping and the ESF switchgear room coolers had been accomplished and MPE had determined that the system would maintain the ESF switchgear rooms at 104 F or less under accident conditions. Also, the licensee committed to take differential pressure (DP) readings daily across the room coolers and to monitor daily the operating control room air conditioning compressor inlet pressure. At a period of every two weeks annubars would be installed at the flow points for all ESF electrical switchgear room coolers to monitor system flowrates. These commitments were interim measures until September 15, 1986, or until spent fuel movement is required. At that time other measures will be required. Based on the commitments noted the licensee started up on August 31, 1986.
On September 5, 1986, the licensee gave the inspectors an evaluation of low flow to the ESF switchgear room coolers as reported by Material Non-conformance Report (MNCR) 642-86. This report (86-M-003) used the as-found data for SSW loops A and B and found the following ESF switchgear room coolers to be outside the design tolerance of 10%.
The test results were as follows:
ESF ELECTRICAL SWITCHGEAR DESIGN FLOW AS FOUND FLOW ROOM ROOM COOLER (GPM)
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1A208 B001A-A
3.60 1A219 B002A-A
29.26 1A309 B003A-A
19.03 1A410 B004A-A
0.00 1A207 B001B-B
2.81 1A221 BPP28-B
30.94 1A308 B003B-B
28.45 1A320 B005B-B
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The report reviewed the accelerated aging temperature used to calculate the qualified life for all 10 CFR 50.49 qualified components within the above rooms.
The weak link components are considered to be the 7.2 kV Power Vac breakers End of Cycle-Recirculation Pump Trip (E0C-RPT) breakers which provide power to the reactor recirculation pump motors. Bechtel Specification E-009.4, Power Vac Switchgear Breakers, was reviewed and found to contain several components which were qualified separately.
The lowest test temperature used involved 60 C (140 F). The condition resulting in SSW supplying the room coolers in combination with peak heat loading in the Switchgear Rooms is a total loss of of fsite power in conjunction with LOCA loads when operating. This event would cause the RPT breakers to operate immediately.
In this time frame the room temperature would not have risen much above ambient room temperature.
Therefore these breakers will have performed their safety function well before approaching the 140 F qualification temperature. The next weakest link items identified (i.e. Agastat relays)
are installed within the ESF load centers (Bechtel Specification E-017.0)
which used accelerated aging profiles of 160 F and 185 F.
A calculation was performed which demonstrates that the 10 CFR 50.49 equipment installed in the subject areas have sufficient qualified life to operate in a 158 F
environment for 100 days up to the time when the work was done to correct tne problem. The calculated peak room temperature, using the as-found flow conditions, is 147 F.
This is well below the temperature thresholds of these components. Therefore, the equipment would have remained operable throughout the analyzed event.
Per the results of the analysis, all rooms could be maintained at a temperature less than 140 F with exception of the Division I ESF Switchgear Room, 1A207, Elevation 119' and the Division II ESF switchgear room, 1A207, Elevation 119'.
As shown by analysis, Room 1A208 could be maintained less than 141* F and Room 1A207 less than 147* F.
Credit for the coolers capability to remove heat was not taken for these rooms.
Based on the above, the licensee concluded that prior to shut down on August 25, 1986, although the ESF switchgear rooms would have exceeded 104 F room design temperatures in a design basis loss of coolant accident, combined with a loss of off site power, the event would have had no adverse safety
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consequences as safety related components in these rooms would still nave remained operable at this time in plant life.
On September 9, 1986, a telecon was held between the licensee, Mr. L. Kintner,
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NRC Licensing Project Manager, Mr. H. Dance, NRC Region II Section Chief and
the residant inspectors to discuss the effect of License Condition 2 C(20),
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which states that no irradiated fuel may be stored in the spent fuel storage pool prior to completion of modifications to either SSW toop and verification that the design flow can be achieved to all essential SSW system comaonents in the modified loop. Loop B of the SSW system had been modified during the
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fall 1985 outage but, due to silt deposits from the plant service water
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system, as-found flow measurements indicate that the flows would not meet the design flow. NPE evaluated the measured flows and determined that under plant condition 4 or 5 the required flow to maintain essential ESF rooms less that 104* F was substantially less than design flow.
The measured flows were as follows:
MEASURED REQUIRED DESIGN COOLER FLOW,GPM FLOW,GPM FLOW,GPM T46B001B ESF B 1191 EAST 13.8 10.1
T468002B ESF B 1191 WEST 36.3 9.7
T468003B ESF B 1391 EAST 28.1 10.8
T46B004B ESF B 1661 11.7 0.8
T46B005B ESF B 139' WEST 11.3 1.6
T510007B FPC & CU PUMP 13.8 12.6
G418001B FUEL POOL Hx 1182 1065 1065 ROOM T518005B RHRC PUMP 23.7 22.5
ROOM
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Based on the licensee's data, on September 9, 1986, NRC Region II Management
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invoked discretionary enforcement and permitted the licensee to commence refueling operations based on the measured SSW flow to the B ESF switchgear room coolers as shown above.
Other essential coolers also exceeded the required flow. The licensee committed to take DP readings daily and evaluate those readings for possible degradation in flow. Annubars will be installed monthly to measure actual flow. Temporary procedures will guide operations on special considerations that might be applicable in the event of an accident where ESF rooms might see increased heat loads. The licensee is currently modifying the A loop of ESF room coolers piping to provide for flushing / cleaning during operation in case of future fouling. Later in the outage, when the A loop is operational, the licensee will modify the B loop to provide for flushing / clearing.
10.
Inspector Followup And Unresolved Items (92701)
(Closed) Inspector Followup Item 416/85-46-02, Licensee to prepare a procedure for placing shutdown cooling in effect when in operational conditional 2.
System Operating Instruction 04-1-01-E12-1, Residual Heat Removal System, was revised to prohibit the use of shutdown cooling without an approved
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written procedure, which addresses in detail the precautions and limitations associated with the unique circumstances that exist at the time of desired usage. Also, the use of the shutdown cooling mode of RHR in operational condition 2 requires the approval of the duty manager prior to use.
(Closed) Inspector Followup Item 416/86-20-05, Review of Technical Section Instruction 09-S-02-300, Revision 4, SNM Movement Control.
The inspectors reviewed the noted procedure and had no comments. Adequate controls for restricted storage locations for spent fuel were incorporated.
No violations of deviations were identified.
11. Preparation for Refueling and Refueling Activities (60705 & 60710)
The inspectors witnessed the movement of new fuel from the new fuel storage vault into the spent fuel pool in preparation for the outage.
The fuel movement was conducted in accordance with an approved Special Nuclear Material (SNM) movement plan and approved procedures. The inspectors review of procedures revealed two minor discrepancies. Technical Section Instruction 09-S-02-300, Revision 3, SNM Movement Control, paragraph 5.6 stated the restrictions of step 5.A may be relaxed if an engineering analysis indicates plans satisfy TS 3.7.9, 3.9.9 and 5.6.
Omitted was TS 3.9.4 which prohibits fuel movement unless the reactor has been subcritical at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Also, attachment 1, Restricted Storage Locations for Spent Fuel, failed to list rack AS, column Z, row 29 which contains the test coupons for the high density storage racks.
The licensae was notified and corrected the discrepancies prior to the refueling outage.
The inspectors attended several briefings by plant management to present the refueling outage schedule to Nuclear Plant Engineering, Quality Assurance, Plant staff and the other personnel that will be involved in the outage. During the latter part of this inspection period the inspectors reviewed the var.ous refueling control documents and changes thereto, as the outage commenced.
The inspectors also witnessed new fuel transfer from the spent fuel pool to the containment fuel pool, removal of the drywell head, removal of the reactor vessel head, and other supporting refueling activities.
On September 5, 1986, the inspectors witnessed Technical Special Test Instruction (TSTI) 1E51-86-001-0-5 which had three objectives:
Verify the ability of the Reactor Core Isolation Cooling (RCIC) system
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to start without the aid of AC power as would be experienced during a Station Blackout.
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Verify the ability of the Reactor Core Isolation Cooling (RCIC) system to operate for a sustained period of time without the aid of AC power as would be experienced during a Station Blackout.
Evaluate the effects of limiting parameters on the ability to sustain
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prolonged Reactor Core Isolation Cooling (RCIC) system operation without the aid of AC power as would be experienced during a Station Blackout.
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This test was to simulate the station blackout loss of all AC power and evaluate the operation of RCIC under those conditions.
License Cotdition (LC) 2.C(33)(b), which is to satisfy TMI action Item I.G.1, states that MP&L shall conduct a special test, Simulated Loss of Onsite and Offsite Alternating-Current Power Test, as described in the MP&L letter dated August 18, 1981. At least 4 weeks prior to performing the Special Test, MP&L shall provide a safety analysis for this test and its procedures to NRC for review and approval. Subsequently, the licensee has submitted proposed amendments to the operating license to modify LC 2C(33)(b) to read:
Prior to restart following the first refueling outage, MP&L shall complete the additional training and testing related to TMI action plan I.G.1 as described in Section 2.3 of the MP&L submittal dated April 3, 1986.
Although the licensee has not received NRC approval, they are conducting the tests they have proposed to fulfill the BWR Owners Group Program for TMI Action Plan Item I.G.I.
This would be in agreement with the instructions issued as part of Generic Letter 83-24 dated June 29, 1983. The inspectors verified the test was conducted per the latest approved test instruction, test equipment was calibrated, technical specification requirements were ad5ered to, procedural prerequisites and initial conditions were met and test data was accurately recorded. At 12:01 a.m. on September 6,1986, the licensee shutdown to begin the first refueling outage.
No violations or deviations were identified.
12.
Review of Quality Assurance for Unit 2 Extended Construction Delay (92050)
This inspection was conducted to provide a periodic review of Quality Assurance (QA) activities for a construction site under an extended delay.
The inspection was to determine whether the licensee has established adeauate implementation plans, instructions and procedures which are in conformance with the established QA plan. The construction activities for Unit 2 are essentially stopped. A Preventive Maintenance (PM) program for all Q-listed and balance of plant equipment is the responsibility of Bechtel Power Corporation, the Architect-Engineer for the project. The construction work plan / procedure WP/P-15, Maintenance of Materials and Equipment while in Storage, is the controlling document for the PM program. The Mississippi Power & Light (MP&L) QA organization conducts periodic audits of the Bechtel construction activities as described in their QA manual MPL-Topical-1.
The inspector reviewed the applicable procedures, walked through the Unit 2 power block and the warehouse and verified that the equipment is properly maintained as described in WP/P-15.
The following equipment was verified to have received the proper PM:
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Standby Diesel Engine Generator Stator Assembly Inspected 8/15/86 for space heaters
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HPCS Diesel Generator Control Panel H22P118
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Inspected 7/14/86 for space heaters Inspected 7/16/86 for visual Standby Service Water Pump Motor SFM087.0
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Inspected 7/17/86 for oil change Inspected 8/8/86 to apply rust preventive Recirculation Pump Motors
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Inspected 8/18/86 Megger with 500 VDC Inspected 6/18/86 for oil sample A complete listing of all equipment and frequency of inspections for Unit 2 is tracked by the Instorage Maintenance System Bulk Listing (Job NO. 09645001)
The inspector reviewed the following QA audits conducted by the MP&L QA organization for compliance to tFe QA manual:
MAR 85/2-0010 Unit 2 MAR 85/2-0012 Unit 2 No violations or deviations were identified.
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