Text

Insp Rept 50-289/87-09 on 870306-0417.Violations Noted: Failure to Perform Adequate Radiological Surveys Prior to Working in Prefilter Room Resulting in Unplanned Intake of Radioactive Matl by Personnel
	ML20216J581
Person / Time
Site:	Three Mile Island
Issue date:	06/22/1987
From:	Blough A, Conte R; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML20216J568	List: ML20216J563; ML20216J571; ML20216J581;
References
	50-289-87-09, 50-289-87-9, NUDOCS 8707060201
	Download: ML20216J581 (80)
	v • d • e

U. S. NUCLEAR REGULATORY COMMISSION

REGION I

Docket / Report No. 50-289/87-09 License: DPR-50

Licensee: GPU Nuclear Corporation

P. O. Box 480

Middletown, Pennsylvania 17057

Facility: Three Mile Island Nuclear Station, Unit 1

Location: Middletown, Pennsylvania j

Dates: March 6 - April 17,1987 l

Inspectors: L. Briggs, Lead Reactor Engineer, Region I (RI)

D. Coe, Reactor Engineer (Examiner), RI

R. Conte, Senior Resident Inspector (TMI-1)

J. Golla, Reactor Engineer, RI

D. Johnson, Resident Inspector (TMI-1)

A. Krasopoulos, Reactor Engi.eer, RI i

K. Murphy, Technical Assistant, RI l

B. Norris, Reactor Engineer (Examiner), RI j

S. Peleschak, Reactor Engineer, RI !

S. Pullani, Fire Protection Engineer, RI l

J. Rogers, Resident Inspector (TMI-1) '

P. Wen, Reactor Engineer, RI

A. Weadock, Radiation Specialist, RI

F. Young, Resident Inspector (TMI-1)

Accompanied By: D. Kubicki, Technical Reviewer, NRR

n ec r: ~

p,

R. Conte, Senior Resident Inspector

6//9//)

'Date

Approved by: k

A. Bloup(, Chief, Reactor Section No. lA, DRP

0 F/

Date

Inspection Summary:

Region I staff conducted safety inspections (800 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months ) of licensee post-

outage, heatup/startup, and power operation activities which included 24-hour

inspection coverage of licensee transition from cold shutdown to power opera-

tion. Shift inspectors focused on overall plant and personnel performance;

startup readiness, along with material condition of the plant; and, technical

8707060201 870623

PDR ADOCK 05000289

G PDR

1

Inspection Summary 2

support for the Cycle 6 startup. As a result of normal resident inspector

coverage and 24-hour monitoring, follow-up reviews occurred on: contamination / l

uptake events; feedwater injection event; condenser offgas monitoring operabil-

ity; condensate storage tank level oscillations; pressurizer insulation instal-

lation; steam generator level calibration problems; remote shutdown relay

deficiency; reactor coolant system (RCS) leakrate; reactor water level instru-

,

I

mentation operability; and, quality assurance involvement in Cycle 6 startup.

Review of the following additional areas occurred: post-outage review of radi-

ation protection; post-modification / refueling testing; Heat Sink Protection

System (HSPS); remote shutdown panels; low power and startup physics testing;

special and periodic reports, including the containment integrated leak rate

test report; semi-annual release and effluent data report; licensee event re-

ports; and, annual reports. A special review of related surveillances for the

RCS pressure isolation valves occurred. Substantial follow-up occurred on ,

licensee action on previous inspection findings, especially in the fire protec- "

tion areas with the inspectors assisted by a technical reviewer from the Office

of Nuclear Reactor Regulation (NRR).

Inspection Results:

In general, the licensee made personnel, equipment, and building spaces ready l

to adequately support operations. A number of licensee initiatives contributed !

to this overall readiness. However, the most significant problem was the

alignment / calibration error for the steam generator (SG) level instrumentation

channels of the Heat Sink Protection System (HSPS). This instrumentation was

not fully ready to support operation; and it, in part, caused operator confus-

ion during precritical operations. Several factors contributed to the failure

to identify the SG level instrument misalignment sooner in the prerequisite

process. Compensatory measures were established for decay heat removal con-

siderations with the reactor critical at low power and while the licensee

investigated the SG level indication problem.

Also, as reflected by the number of unresolved items in the design and engi-

neering area, the inspectors continued to note signs of weak technical support.

The open issues centered around: procurement problems; vendor interface prob-

lems; and, weak design due, in part, to apparent lack of consideration for cer- l

a

tain human factors. An apparent violation of regulatory requirements was noted

in that sufficient thermal growth was not factored into the design of the pres-

surizer platform (see paragraph 4.2.6.3). Also, the inspectors continued to

note a reliance on the test program to identify wiring errors as a result of

maintenance / modification work.

I

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1

Inspection Summary 3 i

1

,

The inspectors concluded that the operators conducted themselves in a competent l

and safe manner. Extensive training provided to the operators was apparently ;

conducive to their adequate response to off-normal situations, especially for

problems noted on HSPS SG level. In general, evolutions and testing were con-

ducted in accordance with applicable station procedures. Minor nonadherences

were noted, especially when procedural steps were conducted from memory. As

equipment problems arose appropriate corrective actions, in general, were

initiated.

Overall, test procedures were adequate to perform their intended functions. In

general, the test procedures were properly implemented. For major tests, there i

was reasonably good interface between test personnel and plant staff, espec-

ially in the reactor engineering area. Minor communication deficiencies were

noted between startup and test personnel and other groups, such as operators,

maintenance, and NRC staff. The most significant event as a result of these

communication deficiencies was the inadvertent pressurization of the "B" steam

generator due to three evolutions being conducted simultaneously. The pace of

activities was a factor in this event. There was reasonably good design pre-

diction of measured reactor core parameters.

Based on this limited review, the inspectors verified proper completion by the

licensee of Cycle 6 startup and other related NRC requirements and licensee

commitments. The areas covered were: reactor water level indication system; 1

emergency feedwater system; fire protection upgrades; and, reactor coolant sys-

tem pressure isolation valve surveillance. Additional detailed NRC staff re-

view will be warranted. A large number of past inspection findings in these

areas were satisfactorily resolved by the licensee. For these past findings,

the licensee was responsive, in general, to NRC staff concerns.

Required reports were made in a timely manner. In general, they appeared to be

adequate, but more detailed NRC staff review is planned as time and resources

permit.

One violation involving failure to perform adequate radiological surveys prior

to working in a prefilter room (see paragraph 7.3.2) was identified. This

failure resulted in an unplanned intake of radioactive material by personnel.

Licensee corrective a:tions for the letdown prefilter room events extend over a.

lengthy schedule; however, a number of those events are design related and need

to be well thought out before effective solutions are implemented. The licen-

see appears to need to pursue letdown prefilter cubicle design problems. Over-

all, based on this review, the radiation protection program is considered to be

generally sound. In general, the radiation protection program appears to be

properly implemented. Since the main focus of this inspection was an event

follow-up, some negative findings were expected. A noteworthy characteristic

identified by the inspector was the quality of the licensee's self-review

process in this area.

Overall, the plant was returned to power operation in a relatively smooth

manner.

/

TABLE OF CONTENTS

Page j

i

1. Introduction and Overview . . . . . . .... ...... 4

2. Cycle 6 Startup Readiness . . . . . . . . ........ 5

3. Shift Inspection Activities . ........ .... .. 9

4. Plant Operations. .......... ... ... ... 15- !

l

5. Post-Modification / Refueling Tests . . . . . . . . . . . . . 41 )

6. Pressure Boundary Isolation Valve Surveillance. . ..... 50

7. Radiation Protection. . . . . . . . . . . . . . . . . . .. 53

8. Regulatory-Required Reports . .... .. ........ 63

9. Previous Inspection Findings in the Fire Protection Area. . 68

10. Licensee Action on Other Previous Inspection Findings . . . 79

11. Inservice Testing Program Relief Request Resolution . . .. 85

12. Exit Interview. .......... ........... 86

Attachment 1 - Activities Review

Attachment 2 - RCS Leak Rate Results (3/21-31/87) j

<

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DETAILS

1.0 Introduction and Overview i

1.1 Background and Purpose

With the shutdown of TMI-1 on October 31, 1986, the licensee com-

pleted the first cycle of operation since the TMI-1 restart and l

entered a scheduled five-month outage for refueling and extensive l

facility modifications. Significant modification work included up- l

grading of the fire protection and emergency feedwater systems. Also,

a number of the restart commitments and TMI Task Action Plan (TAP)- i

items were due to be completed for this startup. In light of the 1

outage length, significant licensee workload, and scope of modifica-

tions, Region I conducted a special readiness assessment team inspec- I

tion (NRC Inspection Report No. 50-289/87-06). The TMI-1 Resident 1

Office continued that assessment, which included 24-hour inspection i

coverage supplemented by Region I-based inspectors for licensee

startup activities.

The purpose of the inspection was to continue to assess the licen-

see's overall readiness for startup. Also, the purpose was to assess

overall licensee control of activities during the transition period j

from cold shutdown to power operations. The main focus of the in- !

spection was on operating staff performance as a result of perform-

ance-based training and on the adequacy of licensee management reso- l

I

lution of technical and safety issues prior to and during plant

startup.

1.2 Licensee Cycle 6 Startup Activities

During the period of March 19-27, 1987, the significant TMI-1 opera-

tional milestones included: (1) plant heatup; (2) partial cooldown

from outside the control room; (3) taking the reactor critical; (4)

completing low power physics testing; and, (5) completing initial

main turbine generator testing and electric power generation up to

80-85 percent of rated power. The chronological summa ry of plant ,

'

operations during this period is listed below.

Date Time Description

3/20/87 11:51 a.m. Containment integrity set in prepara-

tion for start of plant heatup later

that day

3/22/87 9:00 a.m.- Remote Shutdown Panel Test

12:00 Noon

3/23/87 5:50 p.m. Reactor critical

,

E_____'_____________- '- ' '

5

!

Date Time Description

1

3/26/87 5:15 a.m. Main turbine generator tied to the {

regional electrical grid 1

!

3/28/87 4:15 a.m. Achieved 75% reactor power I

I

3/30/87 5:35 p.m. Achieved maximum achievable reactor

power (84%)

1.3 Operational Events

There were no significant operational events during this period.

Minor equipment actuations and problems were addressed in paragraphs

3 and 4.

2.0 Cycle 6 Startup Readiness

During the inspection, the inspectors continued the review of plant equip-

ment readiness and licensee preparations for Cycle 6 ttartup to assess the

readiness of the plant for startup. The specific areas of focus included

safety-related building spaces; outstanding licensee-identified items in

the surveillance, maintenance, and modification areas; outstanding NRC

inspection findings; and selected valve lineups. The objective was to

identify equipment operability problems that could adversely affect safe

operation of the facility. l

The results of this review are documented below.

2.1 Safety-Related Building Spaces

Periodically, the inspectors observed safety-related building spaces

to identify any loose equipment, scaffolding, or other problems such

as fire hazards / housekeeping that could adversely affect the opera-

bility of s&fety related equipment in adjacent areas. The inspection

also included a review of the related procedures which provide admin-

istrative control of such miscellaneous equipment stored in safety-

related areas.

Selected areas of the following safety-related buildings were inspec-

ted: reactor building; auxiliary building; fuel handling building;

intermediate building; diesel generator building; river water screen-

house; and, ;ontrol building.

6 i

l

i

In general, equipment storage was satisfactory and in accordance'with l

established procedures. Unrestrained equipment without rollers was l

located and positioned on the floor at a safe distance to not ad- i

versely affect safety-related equipment if it fell over. Equipment

on rollers was restrained with wire rope. Scaffolding was minimal

and, where it was installed, it was rigid and restrained to the !

building walls using wire rope and cement anchors. Loose equipment ]

on top of platforms or brackets was minimized and were assumed to '

fall and, therefore, kept at a safe distance from the safety-related

equipment. 1

During plant walkthroughs on Sunday, March 22,1987, the inspector ,

noted certain instances where housekeeping practices and storage of 1

scaffolding (taken down) were poor. Of particular note was the aux- I

iliary building (AB) ventilation room. An excessive amount of re-

moved scaffolding was temporarily placed between the ventilation

filter trains making operator access difficult. In addition, the

inspector noted minor storage problems in other areas, but they did

not adversely affect safe operations of the plant. After the dis-

crepancies were brought to the attention of licensee management, the

concerns were eventually corrected. The storage problem in the AB

ventilation room was corrected later in the week.

In general, the licensee was very successful in removing scaffolding

(with exception as noted above) and cleaning up the plant in a short

period of time. Subsequent tours later in the startup process found

housekeeping to be adequate.

The inspector also noted that the licensee's program to decrease the

number of areas that had slowly become posted radiation or contami-

nation areas produced positive results by an aggressive radiological

cleanup program. The purpose of this program was to allow access to

a large percentage of the plant without requiring substantial radio-

logical controls.

2.2 Outstanding Licensee-Identified Items

The inspector reviewed selected portions of the licensee's applica-

ble corrective action tracking systems to di:termine if any adverse

condition for safety-related equipment operability existed. The

inspector's review included tracking systems for open maintenance job

tickets, open exceptions and deficiencies (E&D's) associated with

technical specification surveillances, and open plant modification

incomplete work list items.

The inspector reviewed the list of open job tickets and discussed all

selected outstanding work items with cognizant personnel. The in-

spector determined that the work required by these jobs would not

have an adverse effect on the plant safety if not performed prior to

returning the plant to operations.

.

. .

... . . .

.. . .

7

The inspector reviewed the list of E&D's noted by the licensee for

all current surveillances. From a sampling review, the inspector

determined that the licensee was properly conducting required sur-

veillances and none of the noted deficiencies or exceptions would

adversely affect plant safety.

The modification incomplete work items list (IWL) was also reviewed.

The inspector found no condition that would adversely affect plant

safety in this area.

2.3 Outstanding Inspection Findings

During the readiness assessment team inspection, the inspectors .re-

viewed the Region I file of outstanding inspection findings to iden-

tify any equipment operability problems adversely affecting sa fe

operation of the facility. No conditions adverse to nuclear safety j

were identified. i

?

2.4 Valve Lineup Verifications

The inspectors independently verified the position of safety-related

valves with the aid of auxiliary and licensed operators. The follow-

ing operating procedures were reviewed:

--

Operating Procedure (OP) 1104-1, Revision 17, dated December 8,1986,

" Core Flooding System;"

--

OP 1104-11, Revision 25, dated December 3,1986, " Nuclear Services

Closed Cooling Water System;" and,

!

--

OP 1104-30, Revision 25, dated January 19, 1987, " Nuclear River ]

Water." I

Additionally, the inspectors independently verified that containment

integrity had been established by witnessing a supervisory lineup of

containment isolation valves and verifying that proper conditions

existed inside the containment. This was accomplished in accordance

with Operating Procedure (OP) 1101-3, Revision 43, dated January 14,

1987, " Containment Integrity and Access." In addition to verifying

valve lineups and proper capping of containment penetrations, other

items were verified to be acceptable on a sampling basis.

--

Proper attachment of removable gratings

--

Missile shields in place

--

Spoo'. pieces removed where applicable

a

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.

..

.

.. .

. . . .

.. . .

. . . . .. .

. .

. . . . . . . .

.- .

. . . .- .

. . . . -

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

8

--

Verification of 30 degree maximum opening for AH-V-1A, B, C, and

D containment purge isolation valves

i

--

Containment sump accessible and free of debris

Several minor problems, such as loose gratings, leaking test connec-

tion caps, and a main steam line snubber which appeared to be blocked,

were noted which were corrected during the lineup or were noted for

later correction or evaluation by licensee personnel. Subsequent 1

inspections of the containment verified that these items were cor- I

rected. The inspectors concluded that proper containment integrity ,

had been established for the components inspected or action initiated !

to correct documented discrepancies.

Also, the inspectors observed the licensee's administratively-con-

trolled valve alignment check for the high pressure injection (HPI)

and low pressure injection (LPI) systems. No unacceptable conditions l

were identified.

1

In general, the valve lists were determined to be accurate and valves '

checked were in their proper position. However, system lineup for

reactor coolant system (RCS) did not reflect the current condition of

several RCS instrument isolation valves and the operator performing

the valve lineup was unaware of the cause of the misalignment. Sub-

,

j

sequent discussions with the Instrument and Control (I&C) lead fore-

man indicated that instrument testing was being performed at that

time and the valves were in the position desired by plant staff.

These valves were later repositioned to support startup.

2.5 Engineering Analysis to Support Startuy

The readiness assessment inspection report identified that a number I

of licensee engineering analysis for the Heat Sink Protection /Emerg-

ency Feedwater (HSPS/EFW) systems remained to be completed and :

issued. The licensee committed to complete this action before Cycle l

6 startup (Unresolved Item Nos. 289/87-06-08 and 09).

The licensee completed these analyses and issued appropriate internal

documentation. The inspector performed a cursory review of these

documents to assure that the licensee completed the stated action and

to assure that there were no obvious safety issues.

The inspector noted last minute modification actions occurred as a

result of these analyses. For example, a single failure suscepti-

bility was identified at the common instrument air supply line for

.

'

the pressure regulator for the turbine-driven emergency feedwater

pump. This was corrected before Cycle 6 startup. Overall, the

analyses substantiated the design criteria for the systems and no

major modification work was needed.

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9

The reports varied in length, detail, and thoroughness. The NRC

staff will use additional specialist personnel to complete a more

thorough review of the HSPS/EFW analyses. Accordingly, the adequacy i

of these analyses remains unresolved pending further NRC Region I '

review. l

i

During the course of this inspection, the inspector became concerned i

that the main feedwater isolation on high steam generator level func-

tion of HSPS might be susceptible to a single failure causing main i

feedwater (MFW) isolation to a steam generator. After discussions I

with licensee engineering personnel and review of design drawings,

the inspector concluded that the function was not subject to a single i

failure. In a feedwater path, HSPS trains "A" and "B" control each l

of the associated regulator and isolation valves. Accordingly, both '

Trains "A" and "B" must actuate in order for the valve to close. A !

single failure in either train would not cause the valve to inadver- )

tently close. l

Also, on April 14, 1987, prior to startup, Region I personnel met

with licensee personnel on the deficiencies identified by NRC staff

in the environmental qualification area (NRC Inspection No. 50-289/

87-01). The licensee reported that these deficiencies were reviewed

and none warrart hardware changeouts. Region I personnel were satis-

fied - with the licensee's presentation and a confirmation follow-up I

inspection will occur after Cycle 6 startup. l

l

2.6 Readiness Summary

i

No conditions were identified that would preclude safe operation of !

the facility. Additional issues were identified, in retrospect, con- l

cerning readiness of equipment to support startup and these were

discussed in Section 4.

3.0 Shift Inspection Activities

3.1 Scope of Review and Observations

The NRC's TMI-I Resident Office monitored licensee heatup, startup, i

and power ascension activities on a 24-hour basis from 3:00 p.m. ,

March 19,1987, to 6:00 p.m. March 27, 1987. The NRC shift inspec-

tors assessed the adequacy and effectiveness of operating personnel

performance based on the inspectors' observations of preoperational

and startup activities to determine that:

--

operators are attentive and responsive to plant parameters and

conditions;

--

plant evolutions and testing are planned and properly authorized;

-

_ - . .

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--

procedures are used and followed as required by plant policy;

--

equipment status changes are appropriately documented and com-

municated to appropriate shift personnel;

--

the operating conditions of plant equipment are effectively

monitored and appropriate corrective action is initiated when

required;

--

backup instrumentation, measurements, and readings are used as

appropriate when normal instrumentation is found to be defective

or out of tolerance;

--

logkeeping is timely, accurate, and adequately reflects plant

activities and status;

--

operators follow good operating practices in conducting plant

operations; and,

--

operator actions are consistent with performance-oriented !

training. l

Shift inspectors also assisted in the overall restart readiness re-

view (described in Section 2) by conducting selected valve and

breaker lineup verifications of safety-related systems.

1

The shift inspectors' observations included, but were not limited to,

those shutdown facility operations, reactor plant startup and testing

operations, periodic surveillance activities, and preventive and

corrective maintenance activities listed in Attachment 1.

The shift -inspector assured that any potential safety concern or

regulatory finding was promptly identified to the licensee's shift

supervisor. Those items requiring additional staff review or

follow-up are described in Section 4 of this report.

3.2 Shift Inspection Assessment

i

3.2.1 Plant Operations

The plant operations department continued to maintain good

overall command and control of functions in the control

room. Operators conducted business in a professional man-

ner and the control room layout remained conducive toward

this goal with the segregation of. where work was performed.

Operations department management insisted and, in most,

cases achieved a relatively quiet control room despite a

relatively large number of personnel present at times (up

to two dozen during significant evolutions ur major

testing).

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In general, operators imparted a cautious approach to major

evolutions and testing. Ample test briefings were conduc-

ted. Additional crews were used to support complex test or

evolutions. During test delays, operations department took

advantage of that time to run standby equipment as a final

check out of readiness.

Overall, procedures were properly implemented. For major

evolutions or tests, strict adherence was noted. For ad-

ministrative procedures and other technical procedures con-

ducted from memory, errors were made as noted in other sec-

tions of this report. The quality assurance department was

very active in verifying proper adherence to procedures. ,

There appeared to be a respect for the use of procedures.

i

3.2.2 Operator Performance / Training l

Overall, operators were knowledgeable of plant design

(including new modifications) and frequently changing plant

status. However, the inspectors noted a weakness or a lack

of full understanding by the operators associated with Heat

Sink Protection System (HSPS). During the steam generator

(SG) level drain down on March 23, 1987, the "B" SG startup

control valve controlled at 30 inches on the startup range;

yet, operators apparently did not notice (in that the evo-

lution was allowed to continue) that the control should

have been 18 inches (see paragraph 4.2.7 on why 30 inches

was set). Also, a number of alarms came in that appeared

to be subctantially off required settings and, again, there

was no stopping of the evolution. The evolution for the

"A" SG was eventually stopped at 18 inches on the startup

range when the operators thought the SG was beginning to

boil dry (see also paragraph 4.2.7). The "A" and "B" SG

levels were restored to 50 inches on the startup range to

be conservative until additional support arrived during the

daylight hours.

Further, on the night of March 23, 1987, with the reactor

critical, due to technician error, two-of-four operating

range level transmitters were inadvertently equalized, in-

dicating high level; and this resulted in isolation of the

'

startup and main feedwater valves for the "B" SG. Also,

during the above-noted inadvertent feedwater isolation,

i

there appeared to be some initial confusion on the part of

the control room operators on how to defeat both HSPS actu-

ations logics for the hign level isolation function in

order to restore main feedwater. With management guidance,

operators quickly restored main feedwater to the "B" SG,

preventing a plant trip.

12

Overall, the inspectors found the plant staff to be appro-

priately trained and prepared for the plant startup. The

above discrepancies can be expected because of operator

working knowledge unfamiliarity compounded by last minute

engineering changes reflecting poor technical support. In

general, operators were conservative in their response

actions to off-normal events.

The NRC's readiness assessment report noted extensive !

classroom and in plant training for the HSPS and fire pro-

tection modifications. Discussions with plant operators I

about plant modifications and evolutions demonstrated that

the operators were cognizant about these various modifica-

tions along with last minute changes. That training proved

to be fruitful based on operator performance during the

remote shutdown panel test and in response to various pro-

blems with steam generator level indications (addressed in l

other sections of this report). The discrepancies as noted !

above could be expected.

3.2.3 Maintenance and Material Condition

Several problems occurred during the heatup and startup

sequence that required involvement by station maintenance.

The inspectors noted on a tour of the remote shutdown

panels that one RCS pressure meter was pegged high. Also

the same pressure instrument recorder in the control room

was noted to be very erratic. Licensee Instrument and

Control (I&C) personnel were aware of this condition and

the instrument was successfully adjusted to eliminate the

anomaly. This was verified by the inspectors on subsequent

tours.

Further, a tour of the reactor building (RB) identified a

small packing leak on the Pressurized Operating Relief-

Valve' (PORV) pilot valve. This was evaluated by plant

maintenance personnel as acceptable. Prior to startup, the

emergency feedwater flow indication read 80-100 gallons per

minute (gpm) with no flow. The licensee corrected this

problem during final startup preparations.

Also noted was a problem with the indication of flow

through the high pressure injection (HPI) lines to the

reactor coolant system (RCS) with no actual HPI flow. These

flow transmitters and their inaccurate indication were the

subject of a previous inspection findings (289/86-17-04).

13

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The licensee adjusted the subject flow transmitters during :

the outage, but the instrument continued to indicate posi- l

tive flow in two of the lines when no flow existed. A j

Permanent Change Request (PCR) was issued to accept this j

condition but the 10 CFR 50.59 did not explain why it i

existed. When the RCS system was pressurized for heatup {

and startup, the erroneous flow indication was reduced, but j

a positive flow indication still existed. It appeared to I

the inspector that these flow transmitters still require

licensee action to resolve the improper indication. This ,

anomaly is one part of the overall concern for delta-P j

transmitter calibration problems subject to static pressure l

shift as discussed later in this report for the Once- l

Through Steam Generator (OTSG) level transmitters (para- i

graph 4.2.7).

The other maintenance activities during the startup were

completed in an orderly manner. Examples of these were

recalibration of NI-1 (source range), which had exhibited

erratic behavior and repair of leaking fitting on a RCS

flow transmitter.

Generally, maintenance activities during the startup were

accomplished in an acceptable manner, except as noted in I

paragraph 4.2.7.

3.2.4 Surveillance Activities

The inspectors witnessed various surveillance activities

throughout the heatup and startup process. Generally, the

surveillances were accomplished in accordance with accept-

able procedures in a safe manner.

The inspector witnessed portions of the control rod drive

mechanism (CRDM) " Drop Testing" accomplished in accordance

with Surveillance Procedure (SP) 1303-11.1, Revision 14,

dated February 27,1987, " Control Rod Drop Time." During

this evolution, the "B" loop RCS flow transmitter indica-

tion in the controi room was out of service for repairs to

the transmitter. Part of SP 1303-11.1 required recording

.

of RCS flow just prior to insertion of the particular rod

group being tested. This data could not be obtained and an

Exception and Deficiency (E&D) sheet was generated. T51s

was corrected by deleting the requirement to record the

flow, as operators could verify that full flow through the

core existed as evidenced by the operation of all four

reactor coolant pumps. The inspector verified that the E&D

was properly dispositioned. The inspector noted that the

procedure prerequisites could have been more complete by

including a requirement that instrumentation for required

data be in service prior to the start of the test.

I

14

!

The inspector also observed portions of the HSPS surveil-

lance procedure which were being implemented via the ;

Special Temporary Procedure (STP) method. The licensee had 1

generated four SP's for quarterly testing of the HSPS logic. ;

These procedures, SP 1303-11.26, 27, 28, and 29 in draft

form, were accomplished as STP's. Changes to the draft

SP's were made, as required, to ensure that later perform- l

ance when the plant was at power would not result in any

adverse transients from inadvertent emergency feedwater

(EFW) automatic initiation.

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During one portion of SP 1303-11.27, emergency feed pumps ;

EF-P-2A and EF-P-1 automatically started due to technician !

error (see also paragraph 4.2.7). ~

Although in normal practice the STP's would require an

approved change when a discrepancy was noted, these STP's

were specially written to allow changes to be made during 1

performance. The shif t supervisor was required to approve I

any changes that could affect plant operations. Except for '

the automatic start of the emergency feedwater pumps

(EFP's), no major problems were noted. The Plant Review

Group (PRG) reviewed the completed STP's and verified that

the surveillances for the HSPS logic were completed satis-

factorily prior to plant criticality. The inspectors re- i

viewed the completed STP's to compare them against the new l

HSPS technical specifications and determined that the HSPS !

technical specification surveillances were adequately com- '

pleted. l

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The inspectors also witnessed portions of the Engineered

Safeguards Actuation System (ESAS) logic testing per SP

1303-5.2. During the performance of this test, a timer

relay for ESAS block RC-2A would not reset. This delayed

testing for a shift while troubleshooting was in progress.

The licensee successfully identified the faulty component -

and replaced the relay and completed the _ test. The inspec-

tors witnessed the remaining portions of the test to verify

that repairs were made correctly.

Generally, surveillance testing during the plant heatup and

startup period was conducted adequately. The licensee used

extra crews of operations personnel to accomplish complex

surveillances while the normal crews maintained plant para-

meters.

-

_ _ _ _ _ _ _ _ _ _ _ _ _ _

_

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3.2.5 Technical Support

Overall, shift inspectors noted a substantial amount of

support from corporate and site engineering personnel for

the startup. The project engineer for the Heat Sink Pro-

tection System (HSPS) was on site on a daily basis and was

available to resolve last minute problems and questions.

He was actively involved in the resolution of the steam

generator (SG) level calibration problem. Startup and Test

(SU&T) personnel, along with nuclear engineers, worked well

with the shift crews to assure appropriate test briefings

and overall coordination of test activities. The shift

inspectors noted relatively few interface or communication

problems between technical support personnel and the opera- 1

ting crews.

'

Design weaknesses noted during shift inspection coverage ;

were addressed in other areas of this report. !

3.3 Shift Inspection Summary

In general, there was overall good command and control of evolutions.

Operators and technicians were knowledgeable of plant design and

status. In general, procedures were properly followed. Errors were

made, but they were relatively minor and they had minimal impact on

plant operations.

Operations were appropriately supported by engineering personnel.

As equipment problems occurred, support personnel and management were

active in their resolution of those problems.

Additional insights are addressed in paragraph 4.3.

4.0 Plant Operations

4.1 Routine Review

In addition to the shift coverage noted above, the resident inspec-

tors periodically inspected the facility to determine the licensee's

compliance with the general operating requirements of Section 6 of

the Technical Specifications (TS) in the following areas:

--

review of selected plant parameters for abnormal trends;

--

plant status from a maintenance / modification viewpoint, includ-

ing plant housekeeping and fire protection measures;

--

control of ongoing and special evolutions, including control

room personnel awareness of these evolutions;

_ _ _ _ _ _ _

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16

--

control of documents, including logkeeping practices;

--

implementation of radiological controls; and, >

t

--

implementation of the security plan including access control,

boundary integrity, and badging practices.

The inspectors focused on the following areas:

--

control room operations during regular and backshift hours, in-

cluding frequent observation of activities in progress and per-

iodic reviews of selected sections of the shift foreman's log '

and control room operator's log, and selected sections of other

control room daily logs;

--

areas outside the control room;

--

selected licensee planning meetings; and,

--

various activities listed in Attachment 1.

As a result of this review and shift inspector observations, the

inspectors reviewed - specific events in more detail as noted below.

4.2 Findings / Conclusions J

4.2.1 General

Licensee site and corporate management continued their

detailed attention and involvement in plant operations.

The inspectors noted a number of corporate-based engineer- 1

ing and management personnel on site just prior to and dur- I

ing the startup process to assure last minute problems were

resolved and their respective departments were ready to

support operations. As equipment or test problems arose, 1

site management was active in their resolution.

The most noteworthy startup problem was the calibration / ,

alignment errors associated with the HSPS SG level instru- 1

antation (details in paragraph 4.2.7). It caused a delay

the scheduled criticality time on March 23, 1987, and a

delay in escalating above 5 percent power. That instrumen-

tation was not ready to support startup because of several '

.

factors addressed in that section. Without the root cause

of the problem identified, the licensee made the reactor

critical and simultaneously started a substantial trouble-

shooting effort on SG level instrumentation. However, com-

pensatory measures were. established to assure reactor decay

heat removal capabilities through the steam generators by

conservatively resetting the low level limits function.

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17 !

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Overall, housekeeping and fire protection were adequate to i

support startup. A number of problem areas, as noted in {

paragraph 2.1, were discussed with licensee management. In '

general, . the licensee was responsive to NRC staff concerns

in this area.

As reflected by the number of unresolved items listed be-

low, the inspector continued to note signs of weak tech-

nical support for plant operations. The open issues cen- j

tered around: procurement problems; vendor interface prob- '

lems; and, weak design reflected an apparent lack of con-

sideration for certain human factors.

Overall, there was a relative smooth transition from cold

shutdown to power operations.

'4.2.2 Reactor Coolant System (RCS) Leak Rat _e

The inspector selectively reviewed RCS leak rate data for

the past inspection period. The inspector independently

calculated certain RCS leak rate data reviewed using licen-

see input data and a generic NRC " BASIC" computer program

"RCSLK9" as specified in NUREG 1107. Licensee (L) and NRC

(N) data are tabulated below.

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TABLE 1

RCS LEAK RATE DATA

All Values (GPM)

DATE/ TIME (NUREG 1107) CORRECTED

DURATION Lg Ng Ng N L

U U

.3/21/87 1451 0.1642 0.16 -0.03 0.07 0.0717