IR 05000416/1987022
| ML20235U866 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 10/05/1987 |
| From: | Butcher R, Dance H, Mathis J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20235U848 | List: |
| References | |
| TASK-1.G.1, TASK-TM 50-416-87-22, NUDOCS 8710140289 | |
| Download: ML20235U866 (10) | |
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION-
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101 MARIETTA STREET,N.W.
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l Report No..: '50-456/87-22
' Licensee:. System. Energy.Resou'rces, Inc.
fJackson, MS 39205.
Docket'No;:.50-416'-
License.No.:- NPF-29-Facility Name:
Grand Gulf-Inspection Cohducted: August.22 - September 18, 1987-Inspect rs:
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R. C But'cher, Senior Resident Inspector.
Date Signed.
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A J. L athis,- esident Inspector Date. Signed Approved by:-
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H.;C., Dance,.Section Chief Date signed.
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Division of Reactor Projects l
SUMMARY Scope':
This routine inspection was conducted by the resident ins'pectors at '
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the site in.' the areas of Licensee Action on Previous Enforcement Matters,
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Operational Safety Verification,. Maintenance Observation ~,
Surveillance
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Observation, ESF System Walkdown, Reportable Occurrences, Ope' rating Reactor
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Events,fInspectorFollowupandUnresolvedItems,andManagementMeeting.
i Results: No viola'tions or' deviations were identified.
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al REPORT DETAILS
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' Licensee Employees Contacted J. E. Cross, GGNS Site Director
- C. R. Hutchinson, GGNS General Manager
+J. G. Cesare, Director, Licensing and Safety
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+D. L. Pace, Manager, Nuclear Design
+T. H. Cloninger, Vice President, Nuclear Engineering & Support j
+F. W. Titus, Director, Nuclear Plant Engineering
E R.'F. Rogers, Manager, Unit 1 Projects A. S. McCurdy, Manager, Plant Operations
- J. D. Bailey, Compliance Coordinator
-*M. J. Wright, Manager, Plant Support
- L. F. Daughtery, Compliance Superintendent L
D..G. Cupstid, Start-up Supervisor q
R. H. McAnuity, Electrical Superintendent
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J. P. Dimmette, Manager, Plant Maintenance W. P. Harris,' Compliance Coordinator J. L. Robertson, Licensing Superintendent L. 'G. Temple, I & C Superintendent
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J. H. Mueller, Mechanical Superintendent
L. B. Moulder, Operations Superintendent J. V.. Parrish, Chemistry / Radiation Control Superintendent S. M. Feith, Director, QA
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- S. F. Tanner, Manager,; Quality Services Other licensee employees contacted included technicians, operators, security force members, and of.fice personnel.
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NRC Representatives.
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+L.P.
Modenos, ProjectL Engineer, Division of Reactor Projects,
+L. L. Kintner, Licensing Project Manager, Division of Reactor Projects, Nuclear l
Reactor Regulation.
+0.M. Verre111, Branch Chief, Division of Reactor Projects,
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Region II.
+F. Jape, Section Chief, Division of Reactor Safety, Region II
- Attended exit interview on September 18, 1987 L
+ Attended management meeting on September 15, 1987.
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2.
Exit Interview (30703)
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l The inspection scope and findings were summarized on September 18, 1987, I
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with those persons indicated in paragraph I above. The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspecters during this inspection.
The licensee had no comment on the following inspection findings:
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.416/87-22-01, Unresolved Item.
Review of Standby Liquid Control System
tests for acceptability.
(paragraph 7)
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416/87-22-02, Inspector Followup Item. Correction of discrepancies noted l
during walkdown of the Low Pressure Core Spray System.
(paragraph 8)
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416/87-22-03, ' Inspector Followup Item.
Clarification of LER 87-009 to
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. caution operators on valve lineups. (paragraph 9)
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Licensee Action on Previous Enforcement Matters (92702)
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(Closed) Violation 416/86-32-15. The licensee denied this violation based g
on their contention that tne Standby Service Water System should not be
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considered an Engineered Safety Feature.
NRC Region II concurred in the licensee's position and withdrew the violation.
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Unresolved Item *
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One new unresolved item was identified.
See paragragh 7.
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Operational Safety, Radiological Protection and Physical Security Verification (71797, )1709 and 71881)
i The inspectors kept themselves informed on a daily basis of the overall plant status and any significant safety matters related to plant ope-rations. ' 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 on equipment controls and switches, annunciator alarms, adherence to limiting conditions for operation, temporary alterations in effect, daily journals and data sheet entries, control room manning, and access controls.
This inspection
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activity included numerous informal discussions with operators and their supervisors.
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Weekly, when the inspectors were onsite, selected Engineered Safety Feature (ESF) systems were confirmed operable.
The confirmation is made by verifying the following: Accessible valve flow path alignment, power supply breaker and fuse status, major component leakage, lubrication, cooling.and general condition, and instrumentation. General plant tours were conducted on at least a biweekly basis.
Portions of the control
- An unresolved Item is a matter about which more information is required to l
determine whether it is acceptable or may involve a violation or deviation.
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' buildirig, c t'urbineL building, auxiliary building and ' outside ' areas were
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visited.
Observations included safety 'related tagout verifications,
, shift ' turnover, -sampling ~ program, housekeeping and general plant
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conditions, fire protection equipment,Lcontrol of activities in progress,
- problem identification systems, and 1 containment isolation.
At least-monthly, the licensee's onsite emergency response : facilities -were toured l
to determine facility: readiness.
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h Monthly, the inspectors reviewed at.least one' Radiation Work Permit (RWP),
W observed health ~ physics management: involvement and awareness of signi-
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' ficant plant. activities, and observed plant radiation controls. At leaste
' quarterly the. inspectors reviewed' the licensee's program to - limit'-
'e personnel. radiation exposure As Low As Reasonably Achievable (ALARA).
Monthly,-the. inspectors. verified licensee compliance with physical
- security manning and access' control. requirements. At least' quarterly the inspectors verified the adequacy' of physical security detection and
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On August 25, 1987' the inspector noticed on panel.P870 that both hand-switches L(IT51-HS-M607A and B) for the Fuel Pool Cooling and Cleanup (FPCCU) pump' room coolers were.in the AUTO' position. The licensee could-
,notofind.a' manual valve-line up checksheet for the handswitches but System <
,0perating Instruction (S0I) 04-1-01-G41-1, Fuel Pool' Cooling and Cleanup.
System,' paragraph 4.1.1.h, prerequisites, called for both handswitches to be in' the AUTO position.
The licensee ' issued. a change to ' S0I
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L 04-1-01-G41-l adding -the handswitches to the manual valve' lineup : check-
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sheet.
The: inspectors' determined that both switches should be 'in AUTO.
The system design is such that high FPCCU pump room temperatures would Y
- automatically start the FPCCU pump room cooler fans. :This' would notL
.provideocooling since Standby Service Water (SSW) is the source of cooling water 'and the SSW system must be manually initiated.
License Condition (LC):2.C.(21) addressed paragraph 9.4.2 of the Safety Evaluation Report (SER).(NUREG-0831) which required the backup Fuel. Pool Cooling pump room coolers be operable prior to placing irradiated fuel in the spent fuel pool.. Also,- the SER states that each fan coil unit is powered from a separate essential bus and is controlled by room temperature.
Cooling water to the. fan coil units is provided by the redundant trains of the safety-related SSW system.
It was not clear that the reviewer was aware
'that the FPCCU purup room cooler fans would be controlled by room tempera-
-ture but not the required cooling water. The Licensing Project Manager verified that manual initiation of the SSW system was acceptable for this system.
No further action is necessary.
No violations or deviations were identified.
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6.
Maintenance Observation (62703)
t During the report period,.the inspectors observed portions of the maintenance activities listed below.
The observations included a review of the Maintenance Work Orders (MW0s) and other related documents for adequacy, adherence to procedure, proper tagouts, adherence to technical ~ specifications, radiological controls, observation of all or
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part of the actual work and/or retesting in progress, specified-retest
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requirements,-and adherence to the appropriate quality controls.
MWO EL4804, change oil in low pressure core spray pump motor per 07-S-12-43.
MWO M73910, Removal of SSW piping.to inspect for corrosion per 07-S-14-281.
MWO EL-1008, Calibration check of relays associated with control red drive pump A and meters.
MWO 74572, Install temperature thermocouple devices inside D main steam
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line radiation monitor drawer to monitor temperature.
MWO 74581, Main steam line high radiation monitor setpoint drift recali-bration performed in accordance with 06-1C-1D17-R-1002.
No violations or deviations were identified.
7.
Surveillance Observation (61726)
The inspectors observed the performance of portions of the surveillance listed below.
The' observation included a review of the procedure for
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technical adequacy, conformance to technical specifications, verification of test instrument calibration, observation of all or part of the actual surveillance, removal from service and return to service of the system or components affected, and review of the data for acceptability based upon
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the acceptance criteria.
l 06-0P-1023-Q-0001, Revision 24, Drywell Monitoring Valve Operability Test.
06-IC-1E31-R-0023, Revision 26, Reactor Core Isolation Cooling (RCIC)
Steam Line High Flow.
06-0P-1C41-M-0001, Revision 27, Standby Liquid Control Operability.
06-ME-1C41-R-0001, Revision 21, Standby Liquid Control System Relief Valve Functional Test.
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On? September 310, 1987, the inspector' witnessed Surveillance Procedure
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f 06-0P-1C41-M-0001, Revision' 27, Standby Liquid Control, Operability, test -
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J on Standby ' Liquid Control ' (SLC). system A.
While performing = step 5.2.16 the operators - stabilized the discharge pressure at 1220 - psig and. while =
step 5.2.17 was being accomplished, the' relief valve inadvertently lifted.
Operators reduced SLC_ system pressure to allow the relief valve to reset !
and then reperformed steps 5.2.16 and 5.2.17.
The relief did not lift on the, second try. The inspectors did not witness the test on SLC system B~
.butlwas informed by the licensee that the relief. valve did lift ' once -
during that test.
On September 11, 1987, the inspector. witnessed -
. Surveillance Procedure 06-ME-1041-R-0001, Levision 21, Standby Liquid
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Control System Relief Valve Functional Test, for the SLC system B relief
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valve.
The relief valve POP pressure (for tests with acceptable
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prig /second ramp rates)' were 1349.2 psig,1321.8 psig and 1313.1 psig.
The relief valves in the SLC system'have a + 3% tolerance. The licensee's procedure 06-ME-1C41-R-0001 sets the, relief valve at. 1379 + 0,-13 psig prior <to reinsta11ation. Therefore, the relief valve should lift at 1379
+ 41.4 psig'or 1337.6 to 1420.4 psig. The ASME Code,Section XI, article TwV-3500, Inservice Tests, Category C valves, requires safety valve and-relief valve set points be tested in accordance with ASME PTC 25.3-1976.
It appears from the test data that'the SLC system B relief. valve exceeded the allowable tolerance.
TS 4.1.5.d.2. requires demonstrating that the relief valve does not actuate during, recirculation to the test tank.
y During performance of the system tests noted. above, - the relief valve
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lifted at least' once during each test.-
The lifting of the; SLC system
.' relief valve during simulated operation brings into question the opera-bility of the SLC system. 10 CFR 50, Appendix B, Criterion XVI as incor-
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.porated by the licensee's Operational Quality Assurance manual MPL-TOP-1, Chapter 16, states' in part. that procedures shall be established and implemented to pr' ovide for -the evaluation of conditions such as nonconformances, failures, malfunctions, deficiencies, etc, to determine the'need for corrective action and to identify possible adverse quality trends.
Administrative Procedure (AP) 01-S-03-3, Revision 20, Material Nonconformance Reports (MNCRs), paragraph 1.2 states "This procedure shall be used to document' discrepancies concerning material related documentation, i.e.
test results, certification, etc, which leave the continued acceptability of installed hardware indeterminate." Neither the procedure for conducting SLC system functional test, 06-0P-1C41-M-0001, nor the procedure for conducting relief valve functional test, 06-ME-1C41-R-001, require writing nonconformance reports when the relief
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valve inadvertently lifts during system testing or the relief valve POPS outside the design tolerance.
Also, the licensee did not initiate a nonconformance report on the noted discrepancies and no formal evaluation of acceptability of test performance /results was performed. The inspector l
has requested the Region II ASME code specialists review the noted testing J
for acceptability.
Until this review is accomplished this will be l
Unresolved Item 416/87-22-01.
No violations or deviations were identified.
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Engineered Safety Features System Walkdown (71710)
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A comp 1ete walkdown was conducted on the accessible portions of the Low Pressure Core Spray (LPCS) System.
The walkdown consisted of an inspec-
. tion and verification, where possible, of the required system valve-i alignment, including valve power 'available and valve locking where required, instrumentation valved in and functioning; electrical.and
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instrumentation cabinets free from debris, loose materials, jumpers and evidence of rodents, and system free from other degrading conditions.
There were no significant problems identified during the walkdown.
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However, the following discrepancies were noted by the inspectors relating to the Piping and Instrumentation Diagram (P&ID) M1087, Revision 23.
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Relief valve (PSV F018) specified on the P&ID was not labeled.
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Flow control valves..F225 and F226 shown on the P&ID are numbered
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backward in relation to actual installation'out in the plant.
The correction of the noted minor discrepancies will be tracked as Inspector Followup Item 416/87-22-02.
No violations or deviations were identified.
9.
Reportable Occurrences (90712 & 92700)
. The below listed event reports were reviewed to determine if the infor-mation 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, regarding licensee identified violations, i
The following License Event Reports (LERs) are closed.
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LER No.
Event Date Event
- 87-012 August 6, 1987 Reactor scram due to turbine i
control valve fast closure.
The event of LER 87-012 was discussed in Inspection Report 416/87-18.
I During the event discussed in LER 87-009, the Division 2, group 8 isola-tion valves automatically isolated while operators were placing the
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Reactor Water Cleanup (RWCU) system into the blowdown mode of operation.
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In the'LER the. licensee stated that plant directive 04-1-01-G33-1 would be
. revised to note the possibility of a group 8 isolation during valve line up.
In discussions with the inspector, the licensee stated that the significance of that statement wasLto pre-notice-the potential group 8 isolation and if an isolation did occur, it would not be reportable under 10 CFR 50.73(a)(2)(iv). The inspector does not agree with the licensee's position that notating an ESF actuation may or may not occur relieves,them of the reporting requirements of 10 CFR 50.73.
No provision is made for evaluating an evolution where an isolation does not occur. If'an isola-tion should have taken place but did not occur, then operability of.the ESF is in question.
Specific instructions are given in 10 CFR 50.73 directing the licensee in how to obtain written exemptions to reporting requirements.
If ESF actuations inadvertently occur during licensee controlled evolutions, then the licensee should determine the cause of the-actuation and correct the problem so as to maintain control of the system.
By notating in procedures that an ESF actuation may or may not occur, it
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appears the licensee has lost control of the evolution in progress.
Paragraph 6.0 in NUREG-1022, Supplement No. I clearly shows this event would be reportable regardless of procedural notations.
The licensee should. correct LER 87-009 to reflect the actual changes to directive 04-1-01-G33-1 and define what ESF actuations should or should not occur.
Clarification of LER 87-009 will be Inspector Followup Item 416/87-22-03.
No violations or deviations were identified.
10.
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 inspector.s examined instrument recordings, computer printouts, operations journal entries, scram reports and had discussions with operations, maintenance and engineering support personnel as appropriate.
At 1:00 a.m. on August 30, 1987, switchyard breakers J5248 and J5240, which connect the 500 KV Franklin line to the East and West 500 KV busses, opened. Operators were dispatched to the switchyard and found an alarm on the A phase of the Franklin line. The operators reset the lockout and the dispatcher attempted to close the J5248 and J5240 breakers. Breaker J5248 was closed but while attempting to close breaker J5240, breaker J5236 opened. Breaker J5236 connects the Unit 1 generator to the 500 KV East bus. The Unit 1 generator was still connected to the 500 KV West bus.
The ground fault on the Franklin 500 KV line which had initiated the i
original breaker opening had cleared. A Control and Indication (C&I) card l
in the supervisory unit for breaker J5236 was suspected of causing the inadvertent opening of J5236.
The C & I card was replaced and breakers j
J5236 and J5240 were closed without further problems. Subsequent investi-
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gation could not determine a problem with the removed I & C card.
No violations or deviations were identified.
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11.
Inspector Followup and Unresolved Items (92701)
(Closed) Inspector Followup Item 416/87-18-04.
The nuclear fuel vendor,
-Advanced Nuclear Fuel (ANF), found damage in the two returned fuel bundles such that the licensee is scrapping both bundles. However, some pellets from the least damaged fuel bundle will be utilized to make up 3 new fuel pins. The fuel bundles will be ready to be shipped in early September and carry the same warranty as other new fuel bundles.
(Closed) Inspector Followup Itam 416/86-08-01.
As part of license condition 2C (33)(g) the licensee installed pressure gauges in the control room to indicate the. Autcmatic Depressurization System (ADS)
accumulator pressure.
This modification was completed during the first refueling outage and license condition 2.C.(33)(g) was closed in Inspec-tion Report 86-41.
(Closed) TMI Action Item 1.G.1.,
Training during low power testing, was incorporated as a license condition 2..C.(33)(b) into Grand Gulf facility operating license.
The license condition requires prior to restart following the first refueling outage, for SERI to complete the additional testing and training related to TMI Action Item 1.G.1 as described in Section 2.3 of the MP&L submittal dated April 3, 1986.
The MP&L submittal had two tests remaining prior to the first refueling outage.
The first test was the Reactor Core Isolation Cooling (RCIC) system
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with a sustained loss of AC power to the system. This was successfully completed on September 5, 1986, and is documented in NRC report 416/86-26, in paragraph 11.
The second test was the Containment Pressure Instrument Test. This test was witnessed by the resident inspector and documented in NRC report 416/86-39, in paragraph 11.
This particular item was closed out as IFI 416/86-37-05 in report 416/86-39, thus satisfying TMI Action Item I.G.I.
(Closed) Unresolved.' tem 416/87-18-02.
See paragraph 12 for discussion.
12.
Management Meeting (30702)
A management meeting was held in the NRC Region II offices on September 15, 1987.
Attendees are noted in paragraph 1.
The licensee requested this meeting to discuss two recently identified design defi-ciencies.
These deficiencies were:
Redundant isolation valves share a common power source due to a design error.
This item was reported in LER 87-011 and left as a unresolved item in Inspection Report 416/87-18.
Ventilation duct sections not designed to withstand a design basis tornado. This item was reported in LER 87-013 and closed out with no violation issued in Inspection Report 416/87-18.
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-It was noted that GGNS has had several design basis documentation / analysis type deficiencies identified over 'the past fewlyears of. which ithe above noted deficiencies arc' typical.. ;The licensee presented their. findings of'
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each-deficiency, :their-immediate - corrective actions,- supplemental cor-lrective actions, and their safety assessment.
In eachLcase,.although.the design was ' deficient, the safety evaluation concluded,that no significant safety hazards existed due to the' design deficiency.
In addition to those actions described in the LERs the licensee. committed-to revise the LERs tos a:
clarify other areas as discussed in the meeting.
ThoseJareas are:
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. Thel 1icensee stated that an interlock would be installed between;
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isolation valves G33F252 and G33F001-to. prevent opening' both : valves simultaneously;
' Thel statement regarding' electrical power supply redundancy in
- paragraph E.3.(b) of LER 87-011 applied only to the Reactor Water Clean Up (RWCU) system.
The licensee. is to describe the assumptions made in the. accident analysis regarding isolation of.the assumed pipe break.
The licensee is to describe how 'the RWCU. piping' design? meets the requirements of the. Standard Review Plan.
Based on the information presented in the meeting, the significance of the event,. licensee actions taken and. planned, the inspector concluded _that
'the five criteria listed.in Section 5.1 of 10 CFR F, Appendix C, General Statement Policy:and Procedure for NRC Enforcement Actions, have been met and.no notice ~ of violation will be issued. Unresolved Item 416/87-18-02 is closed.
The licensee also presented their Configuration Management Initiatives programs which addressed actions taken or planned in the following areas:
System Design Basis
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Design Basis Calculations Instrument Setpoint/ Control Seismic Qualification Files Environmental Qualification Files Fire Protection Electrical Load Flow Program Station Information Management System (SIMS)
Cable Information System L
As Built Program L
The initiatives that were presented were quite extensive and should result in a greatly improved in-house design capability, improved drawings and improved data storage and retrieval capability.
Based on past history, the configuration management initiatives programs will most probably result in revealing more design basis documentation / analysis deficiencies.
The licensee is to be commended for the extensive improvement
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