ML20246B280

Insp Rept 50-416/89-16 on 890520-0616.Violations Noted.Major Areas Inspected:Operational Safety Verification,Maint Observation,Surveillance Observation,Esf Sys Walkdown, 10CFR21 Procedures & Action on Previous Insp Findings

ML20246B280
Person / Time
Site:	Grand Gulf
Issue date:	06/27/1989
From:	Cantrell F, Christensen H, Mathis J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20246B259	List: ML20246B257 ML20246B267 ML20246B280
References
50-416-89-16, NUDOCS 8907070264
Download: ML20246B280 (11)

See also: IR 05000416/1989016

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NUCLEAR REGULATORY COMMISSION

REGION 11

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101 MARIETTA ST, N.W.

.,,,,. ATLANTA, GEORGIA 30323

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. Report No.: 50-416/89-16

Licensee: ' System Energy Resources, Inc.

Jackson, MS 39205

Docket No.: 50-416 License No.: NPF-29

Facility Name: Grand Gulf Nuclear Station

Inspection' Conducted: May 20 - June 16, 1989

Inspectors: A A 4 If9

e

K 0. Christensen, Ffe f6r#esid' nt inspector Da'te Signed

Yb45 kA

YL Mathis, Residenyryfpsct'Or

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D6te Siigned

Approved.by:

F. 3. 'Cahlrell, Sectidrif#jtbf

dk/d

Dste 5'igned

Division of of Reactor +rojects

SUMMARY

Scopp:

The resident inspectors conducted a routine inspection in the following areas:

operational- safety verification, maintenance observation, surveillance

. observation, engineering safety features (ESF) system walkdown, 10 CFR Part 21

. procedures, action on previous-inspection findings, and reportable occurrences.

The inspectors conducted backshift inspections on May 20, 21, 23, 24, 29 and

June 13, 1939.

Results:

Within the areas inspected two violations were identified: Failure to take

adequate corrective action to prevent RCIC system isolations, paragraph 9, and

failure to maintain post-accident sample system design control, adequate

training, and perform required samples, paragraph 9.

The post accident sample system problems appear to be related to the licensee

treating the system as a non-safety, non-technical specification system (no

= limit conditions for. operations). Controls placed on safety-related systems

are more inclusive and the plant staff appears to focus more attention to the

safety related system status as compared to non-safety related system.

The recirculation pump shaft replacement outage was well managed.

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8907070264 890627' >

{DR ADOCK 05000416

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REPORT DETAILS

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1. Persons Contacted

Licensee Employees

• J. G. Cesare, Director, Nuclear Licensing

W. T. Cottle, Vice President of Nuclear Operations

D. G. Cupstid, Superintendent, Technical Support

• L. F. Daughtery, Compliance Supervisor

• J.'P. Dimmette, Manager, Plant Maintenance

S. M. Feith, Director,' Quality Programs

• C. R. Hutchinson, GGNS General Manager

F. K. Mangan, Director, Plant Projects and Support

R. H. McAnuity, Electrical Superintendent

A. S. McCurdy, Technical Asst., Plant Operations Manager

• L. B. Moulder, Operations Superintendent

J. H. Mueller, Mechanical Superintendent

S. F. Tanner, Manager, Quality Services

L. G. Temple, I & C Superintendent

F. W. Titus, Director, Nuclear Plant Engineering

• M. J. Wright, Manager, Plant Support i

J. W. Yelverton, Manager, Plant Operations

G. Zinke, Superintendent, Plant Licensing

Other licensee employees contacted included technicians, operators,

security force members, and office personnel.

• Attended exit interview

F. S. Cantrell, Section Chief, Division of Reactor Projects, was on site

May 30 and 31, 1989, to conduct a plant tour and hold discussions with the

resident inspectors.

2. Plant Status

Unit 1 began the inspection period in a recirculation pump maintenance

outage. On May 31, 1989, Unit i restarted and returned to power

operations.

3. Operational Safety, (71707)

The inspectors were cognizant of the overall plant status, and of 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.

Observations included the verification of 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

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inspection activity included numerous informal discussions with operators

and their supervisors.

On a weekly bases selected engineered safety feature systems were

confirmed operable. The confirmation was made by verifying that

accessible valve flow path alignment was correct, power supply breaker and

fuse status was correct, and instrumentation was operational. The

following. systems were verified operable: Suppression pool makeup,

control rod drive system, HPCS and LPCS.

General plant tours were conducted on a weekly basis. Portions of the

control building, turbine building, auxiliary building and 'outside areas

were visited. The observations included safety related tagout

verifications, shift turnovers, sampling programs, housekeeping and

general plant conditions, the status of fire protection equipment, control

of activities in progress, problem identification systems, and the

readiness of the onsite emergency response facilities.

The inspectors observed health physics management involvement and

awareness of significant plant activities, and observed plant radiation

controls. Additionally the inspectors verified 'the adequacy of physical

security controls.

The inspector reviewed safety related tagout 892996 (Equipment drain sump

pump). The review ensured that the tagout was properly prepared, and

performed. Additionally, the inspectors verified that the tagged

components were in the required position.

The inspectors verified that the following containment isolation valves

were in' there correct lineup; E22-F035, E22-F022 and E12-F339.

The ' inspectors noted that senior plant management makes routine tours to

the plant and the control room.

The inspectors reviewed activities associated with the failure of the B

recirculation pump shaft. On May 15, 1989, the plant was taken to cold

shutdown due to high vibration on the B Recirculation Pump. Upon

disassembly, a 300 degree crack was discovered on the lower pump shaft.

Both A and B recirculating pump rotating elements were replaced. The

details of the shaft failure are documented in NRC inspection report

50-416/89-15. The recirculation pump maintenance outage was well planned,

scheduled and managed. The plant conducted an orderly restart on May 31,

1989.

No v1alations or deviaticns were identified.

3. Maintenance Observation (62703)

During the report period, the inspectors observed portions of the

maintenance ' Stivities listed oelow. The observations included a review

of the i'WE .nd other related documents for adequacy; adherence to

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L . procedure, proper 'tagouts, technical specifications, quality controls, and

radiological controls; observation of work and/or retesting; and specified

retest requirements.

MWO DESCRIPTION 1

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DCP 86/0085 Extension.of the upper containment pool weir wall

'M 93729 SSW B basin fan D repair

M 93647 Recirculation pump internal inspection

M 93480 Recirculation pump internal removal and replacement

M 934E2 Recirculation pump assemble of spare rotating assembly

EL 11r3 Megger Motor (SBLC) from the breaker

t EL .'1.0 Inspect MOV C41F001B

ME 34m Unit 1 Instrument Air Dryer Desiccant Change Out

No violations or deviations were identified.

5. Surveillance Observation (61726)

The inspectors observed the performance of portions of the surveillance

listed below. The observation included a review of the procedure for

technical adequacy, conformance to technical specifications and LCOs,

verification of test instrument calibration; observation of all or part of

the actual surveillance; removal and return to service of the system or

component; and review of the data for acceptability based upon the

acceptance criteria.

06-0P 1P75-M-0001, Standby Diesel Generator (SDG) II Functional Test,

Attachment 11

06-EL-1E31-M-0001, RCIC Main Steam Tunnel Isolation Delay Timer Channel

A Functional Test and Calibration, Attachment 1

06-IC-1E31-M-0022, Drywell Air Cooler Condensate Flow Rate Monitoring

Functional Test

06-IC-1E32-M-1002, MSIV Leakage Control System Functional Flow Test,

Attachment 1

06-IC-1C34-M-0001, Reactor Vessel Water Level (Level B) MT/RFPT Trip

Function Test 1

06-IC-1D17-A-0012, Fuel Handling Area Ventilation Radiation Monitar

Calibration

06-IC-1E31-M-2003, Main Steam Line "C" High Flow Functional Test

No violations or deviations were identified.

6.. Engineered Safety Features System Walkdown (71710)

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The inspectors conducted a complete walkdown on the accessible portions of

the standby gas treatment system. The walkdown consisted of the

following: confirm that the system lineup procedure matches the plant

drawing and. the as-built configuration; identify equipment condition and

items that might degrade plant performance; verify that valves in the flow

path are in correct positions as required by procedure and that local and

remote position indications are functional; verify the proper breaker

position-at local electrical boards and indications on control boards; and

verify.that instrument calibration dates are current.

The inspectors walked down the system using system operating instruction

04-1-01-T48-1, Revision 19, SBGT and P&ID M-1102 A and B, SBGT system.

The monthly operability test for SBGT systems A and B were performed

satisfactory and the 18 month system logic and vacuum test was

successfully performed. Additionally, the 18 month calibrations for ,

drywell high pressure and reactor vessel water level were performed. The i

annual fuel handling area ventilation exhaust radiation monitor f

calibrations were also performed successfully.

The SBGT electrical lineup was verified by using attachment III to the

system operating instruction. All electrical breakers were in the

required position. The instructions component description differed from

the breakers label name for all breakers. A labeling program has been

implemented to correct all labeling deficiencies.

All annunciators and valve positions were in accordance with the system

operating instruction.

No violations or deviations were identified.

7. 10 CFR Part 21 Inspection (36100)

The inspectors verified that procedures and controls were established and

-implemented for 10 CFR Part 21 requirements. The initiating document for

Part 21 is through a Material Nonconformance Report (MNCR). MNCRs are

used to document discrepancies concerning material-related documentation,

i.e., test results, certification, and etc. Administrative Procedure

01-5-03-3, Material Nonconformance Reports designates Nuclear Plant ]'

Engineering (NPE) as the organization responsible for evaluating whether

deficiency or nonconformances constitute deportability pursuant to l') CFR

Part 21. Additionally, Quality Programs screens all nonconformance

reports for potential Part 21. Quality Assurance Procedure (QAP) 6.40,

Potential Reportable Deficiency Screening, provides guidance to be used by

the screening teviewer for Part 21.

The inspectors reviewed procedure 01-S-09-1, Revision 24, Procurement of )

Materials and Services, which requires 10 CFR 21 be required on all  ;

Quality Level 1 and 2 material item procurement and on all Quality Level

1, 2 and 3 service contracts. The inspector performed a random sample of

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purchase orders written after January 6,1987. All four purchase order

. contained the 10 CFR 21 applicability statement. In addition the inspectors

selected two evaluated deviations or nonconformances not resulting in a

report to the commission to verify the following:

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The item was identified for evaluation consistent with established

procedures.

- The information and data used in the evaluation appear to be factual

and complete.

- The nonconformance was evaluated, or forwarded to the purchaser for

evaluation consistent with established procedures.

Overall it appears that the licensee has an effective program in place for

evaluating 10 CFR 21 requirements.

8. 10 CFR 50.59 Safety Evaluation

In a May 22, 1989, NRC letter to SERI, NRR documented the NRC staff's

safety evaluation results for five SERI 10 CFR 50.59 safety evaluations.

Corrective actions were recommended in three areas:

- Revision of the SERI safety evaluations NPE-86-279 and PLS-86-123 to

include adequate bases to support a determination that the changes do

not involve unreviewed safety questions.

- Revision of the UFSAR to include information to show how safety

significant cranes meet NUREG-0612, as discussed in NRC evaluation of

NPE-86-279.

- Revision of the surveillance procedure for TS 4.6.6.1.b to require

that drawdown test of secondary containment be run with the primary

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containment hatch open as discussed in the NRC evaluation of

PLS-86-136.

Tne resolution of the above recommendation will be an inspector followup

item 89-16-01.

9. Reportable Occurrences (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, regarding licensee identified violations.

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.a. (Closed) LER 89-005, Reactor Core Isolation Cooling System Isolations ,

on Indicated High Steam Line Flow. The RCIC isolations of April 29,

and May 8,1989, were documented in NRC inspection report 89-14.

Upon completion of the investigation, it was determined that the

isolations were caused by a spurious high flow signal produced by

pressure oscillations in a sensing. line for a RCIC steam line

differential pressure transmitter. The sensing lines were backfilled

with demineralized water to reduce the amount and amplitude of the

oscillation and the damping pots were increased to give a stable

signal to the transmitters. Similar isolations occurred in December

1984, as reported under LERs 84-56 and 84-57. The 1984 correction

was to adjust the damping of the transmitters to depress the level of

oscillations. The adjustments were completed by early 1985. During j

refueling outage three, I&C technicians replaced the electronic ,

circuit boards for the transmitters. The work package required the

damping pots to be set to the as found setting. The old pots were

found to be set at minimum damping, therefore the technicians set the

new circuit board pots to minimum. A records search of work

performed on the transmitters did not reveal when or how the old pots

were readjusted. The failure to adequately control the damping

values for the RCIC transmitters is a violation of 10 CFR 50

Appendix B, Criterion XVI Corrective Actions. The LER 89-005 will be

administratively closed and corrective actions will be tracked under

violation 89-16-02.

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b. .0n June 7, 1989, the shift superintendent made a one hour report per

10 CFR 50.72.(b).(v) on the loss of post-accident sampling assessment

capability. While performing a post modification retest, for DCP

87-4018, PAS system atmospheric and liquid panel modification, it was

determined that two valves, P33F719 and P33F720, which were required

to be deleted by t.he design change were installed and isolated,

preventing the capability to take a post-accident liquid radionuclides

sample. Further investigation determined other design package ,

deficiencies, which include: Inadequate post modification system '

walkdowns, several valves and piping not in accordance with as-built

drawing, inadequate post modification test'ing, performed system

retest approximately one month after completion of modification and

at 100% reactor power; and inadequate design package review and

implementation. The package contained vendor equipment numbers

which were not converted to SERI system numbers and the package was

not translated into adequate work instructions, that is valves

P33F719 and P33F720 were left installed contrary to design intent.

It was also determined that the system had not previously been

operated in accordance with the plants license condition, TS, or

emergency plan. Licensing Condition 2.C.(33).(c), requires SERI to

incorporate the requirements of Safety Evaluation Report, Supplement

4 (SSER 4), into procedures. SSER 4. Section 22.2,II.B.3, requires a

post-accident sampling program be performed on a semiannual basis and

consists of obtaining and analyzing reactor coolant, suppression .

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s' responsible: for post-accident procedures. SERIs Emergency Plan,

' Section 7.6.4, states that systems are installed to obtain samples

. from the following . locations: " ...RHR A and B; drywell atmosphere;-

and suppression pool." TS 6.8.3.c . requires a post-accident sampling

program be established, implemented and maintained which will ensure'

the capability to obtain and analyze a sample under accident-

conditions,- training of personnel, procedures for sampling and -

analysis',and. provisions for maintenance of the equipnent.

A review of past sampling data, January 1986 to June 1989, indicated

that semiannual samples were not conducted on RHR or suppression pool

and the RHR L B and drywell sample paths _ were never ' tested.

Additionally, discussions with the systeri engineer and plant-

chemistry indicated that they believed a sample could only be taken

during full' system pressure. Chemistry personnel attempted to take

samples' on April 24 and 27,and May 17,18, and 19,1989, when the

plant wasl sh'ut down and depressurized. Only small amounts of water

were obtained,- this l'ed them to believe that a sample path was

available. The system engineer and plant chemist decided to sample

at pressurized conditions due to the sample difficulties at

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, depressurized conditions. The PAS system was declared operable on

April' 27, 1989. The . plant operated at various power levels for

approximately 20 days prior to conducting a post modification sample

test on June.6, 1989. Chemistry was unable to obtain a sample and on

June 7,1989, the valves P33F719 and P33F720 were discovered in the

syste'n. The licensee performed a dose rate assessment calculation to

determine _'the radiation ~ 1evel that would be present during accident

conditions at the area of the two isolated valves. The calculation

determined an exposure field of 800 rem per hour on contact. (Other

locations _are available that could be used to obtain a sample

although the exposure maybe higher than would be expected from an

operable PAS system.)-

The. failure to maintain the PAS system ooerable, to' conduct

semiannual samples of the RHR and suppre a pool, to demonstrate

the drywell and RHR B sample oaths, and to adequately train the

system engineer and plant chemistry in PAS system operations is a

violation of TS 6.8.3.c, and 10 CFR 50, Appendix B, Criterion Ill.

This will be documented as violation 89-16-03.

c. On June 7, and 8, ~1989 during a hotline phone check, none of the

offsite- emergency agencies could be contacted by the TSC or the

control room emergency notification system. The licensee immediately

notified the telephone company to correct the problem. This incident

was also called into the NRC as a one hour reportable event per  ;

10 CFR 50.72(b)(1)(v).

d. On June 8, 1989 a power outage in Claiborne County led to 22 a rens

being without power, the licensee reported the incident to NRC per

50.72(b)(1)(v).

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10. Action on Previous Inspection Findings (92701,92702)

(Closed) IFI 86-02-03, Inspector Followup Item, NRR conducted review of

BWR technical specifications (TS) governing requirements'for safety relief

valve -(SRV) operability and if changes to TS should be made. NRR

concluded that they do not have a safety concern for those BWR reactors

which have TS allowing unlimited operation with more than one SRV out of

service because: (1) the number of SRVs required to be operabl.e by TS are

based on the overpressure protection analysis as described in the Standard

Review Plan 5.2.2, (2) the number of SRVs out of service up to the maximum

allowed by TS is unlikely to affect fuel cladding integrity during

transients and accidents described in the FSAR Chapter 15, and (3) there

are no regulatory requirements for an ATWS analysis. However, the

discrepancy between the TS limiting conditions for operation with SRVs out

of service - and the GE recommended administrative procedure shou'ld be

resolved.

As a followup to this evaluation, NRR will address the inconsistency

between the TS and the GE recommended procedure with the BWR Owners Group

as a part of the TS Improvement Program.

The Generic Communi~ c ations Branch, NRR has determined that no action is

warranted to inform affected licensees because operation with the present

TS is not a safety significant concern. This item is closed.

(Closed) IFI 89-04-04, Incorporate GE SIL 319 recommendations into PN

program. Maintenance Planning & Scheduling System initiated task card

ME6044 to inspect RCIC Turbine Drive Gear Assembly every 18 months.

Procedure 07-2-14-301, RCIC Turbine Drbre Gear Assembly Inspection, is the

procedure used by mechanical maintenance personnel to inspect the RCIC

gear assembly. This item is closed.

(Closed) IFI 8b-09 n6, Design change on precoat filter isolation causing

scram. The inspectors reviewed DCP 84/3000 and MWO 52146 which provided

an automatic start of the instrument air compressor on low air receiver

pressure and allowed operators to restart the air compressor remotely from

the control room whether it is being operated in the automatic or manual

mode. This item is closed.

(Closed) URI 86-01-02, Review test exceptions to level 1 startup test

acceptance for 10 CFR 50.59 consideration. The inspectors reviewed IPC

86/0296 and IPC 86/363, which reviewed all startup exceptions written to

document level I criteria failures. The findings of t,he review verified

that no further FSAR changes were required. This item is closed.

(Closed) Violation 88-07-03, Violation, Failure to follow procedure for

procurement of materials. The licensee revised the warehouse inventory

computer system to reflect the voltage level and type as part of the model

number for all ASCO solenoid valves in stock. Future ASCO solenoid valves

will be stored according to voltage level and type. This item is closed.

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(Closed) IFI 88-19-02, Inspector Followup Item, Implementation and repair

of 3/4" HCD-73 line. MWO M91252 implemented MCP 89/1011 during RF03 to

repair the vent line for valve Q1833F023A and repair the 4" drain hub.

This MCP modified the vent line and replaced valves B33F025A and B33F026A

with identical valves. This item is closed.

(Closed)IFI 87-16-05, Installation and testing of warning system for high

noise areas of the plant. Visual alarms were installed and tested under

DCP 84/0231. This item is closed.

11. ExitInterview(30703)

'The inspection scope and findings were summarized on June 16, 1989, 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 General flanager stated that

they view the PAS system problem very serious and that it is a result of

multipule breakdowns. .The licensee had no additional comments on the

following inspection findings:

Item Number Description and Reference

89-16-01, IFI Resolution of 10 CFR 50.59 reviews,

paragraph 8.

'89-16-02, V10 Failure to adequately control the damping

values for RCIC transmitters, paragraph 9.

89-16-03 V10 Failure to maintain the PAS system design

control, to conduct adequate training and

perform required samples, paragraph 9.

12. Acronyms and Initialisms

ADHRS- Alternate Decay Heat Removal System

l ADS - Automatic Depressurization System

APRM - Average Power Range Monitor

CRD - Control Rod Drive

DCP - Design Change Package

DG -

Diesel Generator

ECCS - Emergency Core Cooling System

ESF - Engineering Safety Feature

FCV - Flow Control Valve

FSAR - Final Safety Analysis Report

HPCS - High Pressure Core Spray

HPU - Hydraulic Power Unit

1&C - Instrumentation and Control

IFl - Inspector Followup Item

LCO - Limiting Condition for Operation

LER - Licensee Event Report

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LPCI - Low Pressure Core Injection

LPCS - Low Pressure Core Spray

MNCR - Material Nonconformance Report

MWO - Maintenance Work Order

NPE - Nuclear Plant Er.gineering

NRC - Nuclear Regulatory Commission-

PASS - Post Accident Sample System

PDS - Pressure Differential Switch

P&lD - Piping and Instrument Diagram

PSW -

Plant Service Water

QAP - Quality Assurance Procedure

QDR - -Quality Deficiency Report

RCIC - Reactor Core Isolation Cooling

RHR -

Residual Heat Removal

RPS - Reactor Protection System

RWCU - Reactor Water Cleanup

RWP - Radiation Work Permit

SBLC - Standby Liquid Control

SDG - Standby Diesel Generator

SERI - System Energy Resource Incorporation

501 - System Operating Instruction

SSW - Standby Service Water

TCH - Temporary Change Notice

TS - Technical Specification

TSC - Technical Support Center

UFSAR- Updated Final Safety Analysis Report

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