ML20132E225

Insp Rept 50-416/85-20 on 850518-0615.Violations Noted:Maint & Plant Administrative Procedures Inadequate & Two Valves Required to Be Closed Found Open

ML20132E225
Person / Time
Site:	Grand Gulf
Issue date:	07/02/1985
From:	Butcher R, Caldwell J, Panciera V NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20132E194	List: IR 05000518/2006015 ML20132E189 ML20132E225
References
50-416-85-20, NUDOCS 8508010743
Download: ML20132E225 (7)

See also: IR 05000416/1985020

Text

.

.

3 Mog

[

[5 -

Do

n

UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION ll

j 101 MARIETTA STREET,N.W.

' * '- 'C ATLANTA. GEORGI A 30323

\...+/+

Report No.: 50-416/85-20

Licensee: Mississippi Power and Light Company

Jackson, MS 39205

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

Facility Name: Grand Gulf

Inspection Cond c 4 May 18 - June 15, 1985

Inspectors:

R. C.

1 h .

'er, Senior Residtint Inspector

Oh &

,

Date Signed

LeLAJ~Resident I#spector

'ha Isr

Date Signed

J.L.Calfell

'

Approved by: I fmm w y [,./f [

V'. V:' Tantfeia', Sectron Chief D/tF61'gned

Division of Reactor Projects

SUMMARY

Scope: This routine inspection entailed 150 resident inspector-hours at the site

, in the areas of Operational Safety Verification, Maintenance Observation,

Surveillance Observation, ESF System Walkdown, Reportable Occurrences, Operating

Reactor Events, Design, Design Changes and Modifications, Startup Testing, and

Independent Inspection.

Results: Of the eight areas inspected, no apparent violations or deviations were

identified in six areas; two apparent violations were found in two areas.

t

8500010743 850711

PDR ADOCK 05000416

G PDR

- .-- - -- -- - .-

. _ . _ _ _

.

'

.

REPORT DETAILS

1. Licensee Employees Contacted

'

• J. E. Cross, General Manager

• C. R. Hutchinson, Manager, Plant Maintenance

• R. F. Rogers, Technical Assistant

!

• J. D. Bailey, Compliance Coordinator

M. J. Wright, Manager, Plant Operations

• L. F. Daughtery, Compliance Superintendent

D. Cupstid, Start-up Supervisor

'

R. H. McAnulty, Electrical Superintendent

R. V. Moomaw, I&C Superintendent

• B. Harris, Compliance Coordinator

• W. Russell, Assistant, Operations Superintendent

!

• L. G. Temple, Assistant, I&C Superintendent

Other licensee employees contacted included technicians, operators, security

force members, and office personnel.

NRC Inspector

• W. K. Poertner

• Attended exit interview

2. Exit Interview

l

The inspection scope and findings were summarized on June 14, 1985, 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

j inspectors during this inspection. The licensee had no comment on the

following inspection findings:

a. Violation (50-416/85-20-01), two examples (1) inadequate procedure

resulting in a reactor scram; (2) inadequate procedure resulting in

failure to perform a safety evaluation paragraph 10.

.

b. Violation (50-416/85-20-02), failure to follow procedures for valve

lineups paragraph 11.

c. Inspector Followup Item (50-416/85-20-03), revise startup procedure to

,

reflect Technical Specification limits paragraph 12.

3. Licensee Action un Previous Enforcement Matters

This subject was not addressed in the inspection.

4. Unresolved Items

Unresolved items were not identified during this inspection.

,

___ __ . - _ _ . _ -. . -

.

j .

,

'

2

i

5. Operational Safety-Verification (71707)

1

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. Observation 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 activity included numerous informal

discussions with operators and their supervisors.

i

Weekly, when onsite, a selected ESF system is 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.

r

General plant tours were ccqducted on at least a biweekly basis. Portions

of the control building, turbine building, auxiliary building and outside

areas were visited. Observations included safety related tagout verifica-

tions; shift turnover; sampling program; housekeeping and general plant

conditions; fire protection equipment; control of activities in progress;

, radiation protection controls; physical security; problem identification

systems; and containment isolation.

l

4

In the areas inspected, no violations or deviations were identified.

) 6 Maintenance Observation (62703)

i

During the report period, the inspector observed selected maintenance

activities. 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 of the

actual work and/or retesting in progress, specified retest requirements, and

-

adherence to the appropriate quality controls.

'

In the areas inspected, no violations or deviations were identified.

,

7. Surveillance Testing Observation (61726)

i i

The inspector observed the performance of selected surveillances. The

,

observation included a 3 view of the procedure for technical adequacy,

i conformance to Technical Specifications, verification of test instrument-

j calibration, observation of all or part of the actual surveillances, removal

'

>

from service and return to service of the system or components affected, and

review of the data for acceptability based upon the acceptance criteria.

!

, In the areas inspected, no violations or deviations were identified.

I

,

n. .---n. - - ,- - . , . ~ , - .

--m -

,--.---,.-v.,--r -,, -, ,~

.

.

3

8. ESF System Walkdown (71710)

A complete walkdown was conducted on the accessible portions of the Standby

Service Water system in the basin and standby diesel generator areas. 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.

In the areas inspected, no violations or deviations were identified.

9. Reportable Occurrence (90712 and 92700)

The below listed Licensee Event Reports (LERs) were reviewed to determine if

the information provided met 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 LERs were reviewed using the guidance of the general

policy and procedure for NRC enforcement actions. The following LERs are

closed.

LER No. Report Date Event

83-80 10-11-83 Problems with Control Room

HVAC Systems

83-82 08-01-83 Valid Failure of Division I

Standby Diesel Generator

83-103 08-23-83 Shutdown Cooling Isolation

Due to Spurious Isolation Trip

83-156 11-02-83 Capscrew Securing Division I

Diesel Generator Starting Air

Manifold to is Support Plate

Found Broken

• 83-179 12-06-83 Division II Diesel Generator

Shutdown Due to Fuel Line Leak

• 83-191 01-11-84 Standby Fresh Air Unit A

Inadvertently Secured When Being

Used To Meet Action Statement of

Technical Specification 3.7.2

• 83-192 01-23-84 Failure To Adequately Perform a

12 Hour Channel Check

u 1

.

.

4

• 83-53 03-10-83 Electrical Penetrations Opened

For Planned Maintenance

Activity

• 83-162 11-14-83 Drywell Pressure Instrumentation

Readings Exceed Allowable Values

of Technical Specifications

LER 83-179 is associated with violation 83-56-02 which was reviewed and

closed in Inspection Report 84-16.

In the areas inspected, no violations or deviations were identified.

10. Operating Reactor Events (93702)

The inspectors reviewed activities associated with the below listed reactor

scrams. 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 and

engineering support personnel as appropriate.

Scram No. 23 occurred on April 14, 1985, with the reactor operating at 73%

of rated core thermal power. At the time of the scram, Instrumentation and

Control (I&C) Technicians were attempting to calibrate main steam line flow

instruments per Maintenance Procedure (MP) 07-S-53-C34-4 as a prerequisite

to Startup Test Procedure SU-25-3. The MP had been recently changed to

require the reactor feed pump control to be placed in manual and the lifting

of an input lead from the steam flow instrumentation to the vessel level

instrumentation to prevent level changes during the calibration. The

lifting of this input lead resulted in a large drop in sensed vessel level

at the recirculation pump controls causing the recirculation pumps to

transfer back to the Low Frequency Motor Generator (LFMG) Set. The

resulting reduction in recirculation flow, power level and steam flow

combined with the feed pump controls in manual caused vessel level to

increase above the high level scram setpoint, automatically scramming the

reactor. The engineering review of the procedure change noted above was

inadequate in that it failed to realize the magnitude of the sensed level

change to the recirculation pump controls. This review resulted in the

performance of an inadequate procedure which caused a reactor scram.

10 CFR 50, Appendix B, Criterion V states that activities affecting quality.

shall be prescribed by documented instructions, procedures, or drawings of a

type appropriate to the circumstances and shall be accomplished in accordance

with these instructions, procedures or drawings. Failure to provide an adequate

procedure will be identified as violation (50-416/85-20-01).

Also during the review of the procedure change the inspector discovered that

the safety evaluation applicability screening required by Plant Administrative

Procedure 01-S-06-24 was incorrectly performed.' This screening is performed

to determine if a safety evaluation is required. The screening procedure

asks four questions and, if any of the four questions-are answered yes, then

a safety evaluation form (Attachment I of Procedure 01-S-06-24) would be

L

. . . . . . - .

.

.

!

i

required to be filled out. The first question, " changes to the facility as

described in the FSAR", was incorrectly answered. Discussions with the

personnel involved indicated that per their interpretation, procedure

01-S-06-24 did not require a safety evaluation for the lifting of the steam

flow input to the reactor vessel water level controller since it was being

lifted for calibration purposes only. The failure to perform a safety

evaluation was due to an inadequate procedure and will be identified as the

second example of violation (50-416/85-20-01).

11. Independent Inspection (92706)

,

On June 6,1985, while performing a routine startup, operators noticed an

increasing temperature in the steam tunnel followed by an alarm indicating

the steam tunnel blow out panels had opened. An investigation revealed the

'

blow out panels to be closed but there appeared to be a steam leak in the

tunnel. The subsequent shutting of valve ES1-F063, the RCIC/RHR Steam

Supply inboard isolation valve isolated the leak. The licensee discovered

two 3/4" test connection isolation valves Q1E51F207 and Q1E51F208 open. An

investigation by the licensee disclosed that these valves had been last

4

-operated in support of a Local Leak Rate Test (LLRT) 06-ME-1M61-V-0001 on

valve Q1E51F076. The LLRT Valve Lineup Procedure 09-S-08-2 Attachment XIV

'

had been completed subsequent to the LLRT indicating that these valves had

been independently verified in the closed position prior to the reactor

i '

startup. In interviews conducted by the licensee the operators responsible

for closing .these valves stated that they attempted to close these valves

but found the valves already in the closed position. The licensee suggests

1 that these valves were backseated open while they were still hot and had

cooled off causing them to stick on their backseat when the attempt to close

,

them was made. The licensee is looking into procedures for ensuring that

valves are verified in their required positions. Technical Specification

.

6.8.1 requires that procedures shall be established, implemented and

maintained. The failure of the licensee to ensure valve lineup procedure

09-S-08-2 Attachment XIV was correctly completed will be identified as a

l violation of T.S.6.8.1 (50-416/85-20-02).

12. Startup Testing (72530C and 72528C)

The inspector observed all or part of the conduct, or preparation for

conduct, of the below listed startup procedures and operations. The

observation included ' a review of the procedure for meeting all test

,

prerequisites, initial conditions, test equipment and calibration require-

ments. The overall crew performance was observed to ensure that minimum

crew requirements were being met,' that appropriate revised procedures were

in use, that crew actions appeared to be correct and timely, that all data

was collected by the proper personnel for final analysis, and that quick

2 summary analysis showed proper plant response to the test. Where test

i results were available, in preliminary or final form, they were verified to

, be consistent with observations or that overall test acceptance criteria had

j been met.

3

1-000-SU-27-6- GENERATOR LOAD REJECTION

i 1-821-SU-25-6 MAIN STEAM ISOLATION VALVE

i

_ . , _ _ -

--- . _ _ _ - , . _ . ._. .-_ __ ,.

P

.

6

The inspector reviewed the Reactor Startup of June 6,1985. The licensee

pulled critical while RHR Loop A was in the shutdown cooling mode of

operation. Grand Gulf Technical Specifications allows one loop of RHR to be

aligned for shutdown cooling for training startups provided the reactor

vessel is not pressurized, thermal power is less than or equal to 1% of

rated thermal power and reactor coolant temperature is less than 200 F.

The licensee placed RHR Loop A in shutdown cooling after placing the mode

switch in startup to reduce coolant temperature so that the Rod Criticality

data could be obtained at a coolant temperature of less than 150 F in order

to determine core thermal margins as requested by General Electric.

The inspector reviewed Integrated Operating Instruction (101) 03-1-01-1,

Cold Shutdown to Generator Carrying Minimum Load, interviewed management and

operations personnel and reviewed applicable logs and chart recorders to

determine if the licensee had violated Technical Specifications or applic-

able plant procedures.

The inspector determined that 10I-03-1-01-1 addresses a startup with

shutdown cooling in operation, however the procedure does not address the

action statement of Technical Specification 3.10.5 which requires that the

mode switch be placed in the shutdown position if the reactor vessel is

pressurized, thermal power exceeds 1% or coolant temperature exceeds 200 F.

Interviews with operations personnel determined they were aware of the

Technical Specification requirements and the action required if any of the

above parameters were exceeded. The licensee committed to referencing the

requirements of Technical Specification 3.10.5 in their procedure to ensure

the operators are aware of the requirements. Until incorporated this will

be identified as an Inspector Followup Item (85-20-03).

In the areas inspected, no violations or deviations were identified.

13. Design, Design Changes and Modifications (37700)

During Startup Test 1-000-SU31-2, Loss of T-G and Offsite Power, a

deficiency was noted in that manual valve P41-F175A was closed by an

operator when offsite power was lost, a post test review revealed that

previous experience had shown that whenever Plant Service Water (PSW) was

shut down, the Standby Service Water (SSW) basin experienced a loss of

80 gpm due to siphoning of water back through the PSW supply header. The

licensee incorporated administrative controls to keep valve P41-175A closed,

except when filling the SSW basin, as an interim fix. The licensee has now

incorporated a Design Change Package (83/0316, Revision 0) which provides an

open vent path to ensure a vacuum breaker exists. The inspectors reviewed

the installation of DCP 83/0316 and verified, where possible the installa-

tion was complete, the proper procedural controls were followed, and the

change was appropriately reviewed and approved.

In the areas inspected, no violations or deviations were identified.

m