ML20140C564

Insp Repts 50-245/97-01,50-336/97-01 & 50-423/97-01 on 970101-0310.Violations Noted.Major Areas Inspected:Maint, Engineering & Plant Support

ML20140C564
Person / Time
Site:	Millstone
Issue date:	04/11/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20140C552	List: ML20140C546 ML20140C555 ML20140C564
References
50-245-97-01, 50-245-97-1, 50-336-97-01, 50-336-97-1, 50-423-97-01, 50-423-97-1, NUDOCS 9704170127
Download: ML20140C564 (50)

See also: IR 05000245/1997001

Text

.

--

.

.

-

._. _-- -

.

.

,

l

'.

U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Docket Nos.:

50-245

50-336

50-423

1

l

Report Nos.:

97-01

97-01

97-01

,

License Nos.:

DPR-21

DPR-65

NPF-49

i

Licensee:

Northeast Nuclear Energy Company

.

P. O. Box 128

i

Waterford, CT 06385

.

Facility:

Millstone Nuclear Power Station, Units 1,2, and 3

Inspection at:

Waterford, CT

i

Dates:

January 1,1997 - March 10,1997

inspectors:

T. A. Easlick, Senior Resident inspector Unit 1

D. P. Beaulieu, Senior Resident inspector, Unit 2

I

A. C. Cerne, Senior Resident inspector, Unit 3

j

A. L. Burritt, Resident inspector, Unit 1

2

R. J. Arrighi, Resident inspector, Unit 3

'

L. L. Scholl, Reactor Engineer, Region l

N. J. Blumberg, Project Engineer, Region i

R. J. Urban, Project Engineer, Region I

3

,

'

J. T. Furia, Senior Radiation Specialist, Region I, DRS

a

J. E. Carrasco, Reactor Engineer, Region I, DRS

i

Approved by:

Jacque P. Durr, Chief

Inspection Branch

Special Projects Office, NRR

I

1

7

9704170127 970411

PDR

ADOCK 05000245

Q

PDR

f

e

_. -

_ _ .

- -_

_ . - . . _ _ __

___

.___

_

,

,1

,

TABLE OF CONTENTS

EX EC UTIVE S U M M AR Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii

.

4'

'

U1.1 Operations

.................................................. 1

U101

Cond uct of O perations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

U102

Operational Status of. Facilities and Equipment . . . . . . . . . .

1

,

...

U103

Operations Procedures and Documentation

3

,

................

U105

Operator Training Qualification . . . . . . . . . . . . . . . . . . . . . . . . . 3

U108

Miscellaneous Operations issues . . . . . . . . . . . . . . . . . . . . . . . . 5

1

i

U 1.ll M ainte na n c e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6

U1 M2

Maintenance and Material Condition of Facilities and

Equipment

6

.......................................

'

U 1.lli Enginee ring . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7

'

U1 E1

Conduct of Engineering

7

..............................

U1 E2

Engineering Support of Facilities and Equipment . . . . . . . . . . . . . 9

i

U1 E8

Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . 11

S2.1 Operations

12

.................................................

U2 01

Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12

U2 O2

Operational Status of Facilities and Equipment . . . . . . . . . . . . . 12

)

,

U2 08

Miscellaneous Operations issues . . . . . . . . . . . . . . . . . . . . . . . 14

U 2.ll M ainte n a n c e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16

j

U2 M8

Miscellaneous Maintenance issues

16

.....................

U 2.lli Engine e ring . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17

U2 E8

Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . 17

i

4

U3.1 Operations

20

.................................................

U3 01

Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20

U~s 03

Operations Procedures and Documentation

23

...............

U3 07

Quality Assurance in Operations . . . . . . . . . . . . . . . . . . . . . . . 26

i

U 3.11 M ain te n a n c e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27

U3 M1

Conduct of Maintenance

27

............................

U3 M8

Miscellaneous Maintenance issues

29

.....................

U 3.Ill Engine e ring . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31

U3 E8

Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . 31

IV Plant Support

35

.................................................

R1

Radiological Protection and Chemistry Controls

35

............

R8

Miscellaneous Radiological Protection and Chemistry issues

39

...

V. M anage ment M eetings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39

X1

Exit Meeting Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39

il

_

_

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

._ _

,

'l

,

.

'

!

4

!

s

i

EXECUTIVE SUMMARY L

-

!

Millstone Nuclear Power Station'

.

Combined inspection 245/97-01; 336/97-01; 423/97-01

!

!

Operations

• .

Numerous inaccurate Personal Qualification Statements (Form 398) were identified

'

at all four Connecticut plants following NRC questions on two recent adverse

condition reports. Approximately two thirds of the Personal Qualification

!

Statements submitted for recent license applicants were inaccurate. These

applications resulted in the conduct of NRC license examinations and the issuance -

j

of licenses. In a significant number of cases, the licenses were issued without the.

.

candidates fully completing the licensee's training and qualification program, and in

j

a few cases the reactivity manipulations, specifically required by 10 CFR 55, were

i

also not complete. This issue is unresolved for'each Millstone unit pending the

completion of the licensee investigation, resolution of allidentified deficiencies and

.

implementation of programmatic corrective actions. (Section U1.05.1)

!

!

The degraded conditions found in the Unit 1 spent fuel pool are representative of a

'

icng standing disregard for foreign material exclusion (FME) during the conduct of

.i

refuelmg fioor activities. Past low standards for FME control allowed the

,

accumulation of a large amount of debris, which could potentially have a significant

l

-impact on the fuel assemblies stored in the pool. Once the recovery organization

!

became aware of the extent of the problem, by reviewing video tapes, the

i

inspectors noted a good response, including clear direction as to what needed to be

,

8

done in the short term. Based on this information, the acceptability of the degraded

conditions in the spent fuel pool will be unresolved pending NRC review of the

,

issues. (Section U1.02.1)

1

At Unit 1, the licensee failed to evaluate and address a violation concerning the

• -

,

failure to prevent work which had the potential for draining the reactor vessel during

refueling. Further, the licensee failed to provide a comprehensive closure package -

of a quality consistent =with the process committed to in December 1996. The

' manner in which this issue was addressed provides evidence of the licensees ability

to perform effective reviews and to implement appropriate corrective' actions. As a

result of the inspector's concerns with the quality of the completion packages, the

licensee withdrew the NRC completion package schedule. A revised schedule was

still in developmcat at the end of the inspection period. (Section U1.08.1)

Although the Unit 2 backlog of 798 adverse condition reports (ACRs) that are

greater than 120 days old indicates that timeliness for completing corrective actions

remains a concern, the reduction in the backlog of older ACRs from 940 to 798

since the last inspection period is a positive trend which reflects the licensee's

increased level of effort in this area. Timeliness and effectiveness of corrective

actions are areas '... which the licensee must demonstrate sustained improved

performance. (Section U2.02.1)-

l

iii

J

l

i

j

.

A Unit 2 licensee event report discussed that while recovering from a loss of a

direct current (dc) bus, an operator failed to enter an action statement when three

channels of wide range nuclear instrumentation were rendered inoperable. This was

characterized as a non-cited violation. The primary concern associated with this

event was the fact that there was minimal procedural guidance provided to

operators to recover the loss of a de bus. The licensee is in the process of

preparing 12 abnormal operating procedures for recovering various de buses and

distribution panels. This concern is being tracked by an unresolved item. (Section

U2.02.2)

In addition to the physical plant design controls, a longstanding NRC concern at Unit

2 is that operating procedures do not reflect the Final Safety Analysis Report

(FSAR), and an NRC open item has existed since 1993 to address this concern.

This inspection report closes the old open item npqt because adequate corrective

actions have been taken, but because this concern is being addressed and tracked

by more recent items. The issue includes an evaluation of the procedure change

process, as well as the design control process, to ensure future operation is

conducted in accordance with the FSAR. (Section U2.08.1)

The procedure upgrade program has been effective in standardizing procedure

formats across the site. Because of the number of individuals involved in the

procedure upgrades and the long period of time to complete the task, the quality of

procedures vary substantially. As an adjunct to the Demand For Information

process [10 CFR 50.54(f)], which incorporates a verification of the design and

licensing bases, procedure accuracy will be verified. (Section U3 03)

The licensee's root cause investigation and corrective action plan for control of high

energy line break (HELB) doors were determined to be good. However, the

requirement to label the required HELB doors with a minimum number of turns to

ensure proper latching should have been included in the adverse condition report

corrective action plan if it was deemed necessary to prevent recurrence. (Section

U3.01.2)

Good' contingency planning and appropriate consideration of the applicable standard

and regulations were in evidence for both planned operational evolutions and

emergent shutdown conditions. Where necessary to improve shutdown risk

margins, temporary modifications or special system lineup were considered and well

controlled. The licensee development of a standardized approach for disseminating

operations policy for interpreting the language and action statement applicability of

the Unit 3 technical specifications appeared warranted. The examples noted during

this period will be reviewed further as an inspector follow item. (Section U3.01.1)

Maintenance

At Unit 1, the preparation and conduct of work associated with the entry and video

survey of the reactor water cleanup (RWCU) demineralizer room were well

controlled. The inspector noted that good radiological practices were used.

iv

J

.

.

The material condition in the room was acceptable and no deficiencies were

identified. (Section U1.M2.1)

Maintenance and surveillance activities were performed professionally and

thoroughly. All observed maintenance activities were performed with the work

l

package or surveillance procedure present at the job site and personnel were noted

to be closely following the procedures. Review of the surveillance procedures

revealed that the requirements of the applicable technical specifications were

'

appropriately incorporated into the implementing procedure. (Section U3.M1.1)

The licensee developed a Fix-It-Now (FIN) multi-discipline work team approach to

augment the way maintenance is performed at the unit. This process is in addition

to the normal work control process. The FIN work process is being implemented in

a conservative manner. Any work required on protected train equipment was not

being assigned to the FIN team. All monitored work activities performed by the FIN

team were performed in accordance with the unit and station procedures. FIN team

members appeared to be well qualified. (Section U3.M1.1)

Plant inspection-tourt revealed improvements in the Unit 3 areas of housekeeping,

material conditions, and work controls. Field observations raised no new

unresolved safety issues, but did highlight the need for additional management

attention to a previously identified concern regarding the control of temporary

equipment with the potential to adversely impact safety-related components. This

" Seismic II/l" issue will be tracked as an inspector follow item and will receive

further evaluation as a "significant item" in the NRC Restart Assessment Plan.

(U3.M8.1 )

Engineering

A review was performed at Unit 1 of the licensee's progress in resolution of the

Unresolved Safety issue (USI) A-46 outliers documented in the Licensee Event

Report (LER)96-003, Rev. 2. These deficiencies involved inadequate anchorage of

the emergency diesel generator day tank and the turbine building secondary closed

cooling water air coolers. The regulatory requirements for reportability were met

and the corrective action prescribed in the LER were adequate in general, based on

the detailed walkdown of the A-46 modifications, the inspector concluded that the

licensee performed a substantial number of field modifications to accommodate the

seismic loading on mechanical and electrical equipment identified in the USl A-46

scope and documented in LER 96-003; this LER is closed. The followup of the

licensee's commitments to resolve the A-46 program outliers prior to startup for

cycle 16 operation and the assessment of the implication of this event on the

operation of Unit 1 will be unresolved pending further NRC review. (Section

U 1.E1.1 )

The corrective actions taken by the Unit 1 Cnmponent Engineering

Services / Nondestructive Test Engineering (CES/NTE) concerning the use of test

equipment was acceptable. The corrective actions appeared to be broad-based. A

majority of the short term corrective actions were complete, and the long term

v

._

- -

.--

.

-

.

-.

.-

- _.

.

'

4

.

corrective actions were being tracked for closure., The significance of the UT

-

instruments being past their calibration due dates was minimal because they were

. subsequently found to be within tolerance. (Section U1.E2.1)

The NRC is concerned about the operation of the new containment isolation check

a

valve 1-CU-29 at Unit 1, which is operating with lower than expected flow rates

during this extended shutdown. While the use of non-intrusive check valve testing

has verified that the check valve is backseated, this is a short term indication. The

long term effects of the low flow operation have yet to be determined. This issue is

unresolved pending the NRC review of the licensee's final determination of the

operability of the valve prior to plant startup. (Section U1.E2.2)

The inspectors found that overlap test reviews performed in 1993 were not

adequate. The licensee failed to identify this deficiency and improperly concluded

that the 1993 reviews accomplished the actions requested by NRC Generic Letter 96-01, " Testing Of Safety-Related Logic Circuits." (Section U3.E8.1)

t

Plant Support

The licensee has demonstrated a significant increase in management attention

towards work control and maintaining occupational exposures as low as is

reasonably achievable. However, two of the licensee's activities were determined

not to be in compliance with NRC regulations. The Unit 1 violation involves a long-

standing situation (since initial plant start-up), concerning an unmonitored release

pathway in the ventilation system for the radwaste storage building. The violation

at Unit 2 involves a failure to adhere to the licensee's radiation protection program

concerning proper use of electronic dosimeters. (Section R1)

vi

-

-

-

.

Report Details

Summarv of Plant Status

Unit 1 remained in an extended outage for the duration of the inspection period. The

licensee continues to implement configuration management program activities, engineering

reviews, and docketed correspondence assessments to verify compliance with the

established design and licensing basis of the unit. The successful completion of these

activities is required by NRC order prior to restart of the unit. During this period, the

.

licensee implemented a major revision to the corrective action procedure. The goal was to

simplify the process, and at the same time make it more responsive towards restart and

needed organizational improvements. Under the new process, " condition reports" have

replaced " adverse condition reports" to capture both the regulatory defined adverse

condition, as well as other conditions that do not meet managements expectations.

The licensee has recently made two changes to the organizational structure at Unit 1. A

project management group was established to facilitate the implementation of plant

j

modifications prior to startup. Additionally, a restart manager was selected to oversee

j

completion of the Operational Readiness Plan, which will be used to identify and control

i

the actions necessary to achieve and maintain improved performance. The restart manager

will also be responsible for the review of corrective action completion packages, which will

provide objective evidence of corrective action completion. Section U1.08.1 & U1.E8.1 of

this report provides an assessment of the licensee's progress in the area of completion

-

package development. The effectiveness of these changes, as well as the new corrective

action program, will be assessed as part of future NRC inspections.

U1.1 Operations

U101

Conduct of Operations

01.1 General Comments (71707)

Using Inspection Procedure 71707, the inspectors conducted frequent reviews of ongoing

plant operations. The inspectors reviewed operability determinations, availability

determinations, and witnessed the conduct of management review team discussions

regarding the disposition and closure of condition reports (CRs). During a routine tour of

the Unit 1 intake structure, the inspector found the material condition of systems and

components to be adequate. The licensee initiated a material condition improvement

project in the Spring of 1996. The inspector observed some material improvement work

on-going. Specific events and noteworthy observations are detailed in the sections below.

U102

Operational Status of Facilities and Equipment

O 2.1 Soent Fuel Pool Cleanliness

a.

Inspection Scope (71707)

NRC inspection report 245/96-08, dated December 3,1996, discussed the continued

identification of discrepant conditions in the spent fuel pool, indicating a need to accelerate

the evaluation portion of the spent fuel pool cleanup / recovery plan. At that time, the

_m

.

.

___

...

_

. _ . __

._

_.

1

-

,

2

1

-inspectors concluded that all discrepant conditions warrant identification and evaluation in

,

the short term to ensure the collective impact of these issues were addressed. On January

10,1997, the inspectors reviewed a video tape surveillance of the spent fuel pool

conducted on January 3, using an under water camera. The video tapes were reviewed to

J

evaluate the conditions in the spent fuel pool including the fuel storage racks and stored

fuel bundles.

b.

Observations and Findinas

As discussed in NRC report 245/96-08, the earlier video tapes identified improperly seated

)

fuel bundles. In light of the recent video surveys, the licensee determined that the

improperly seated bundles were caused by one of three conditions: 1) There are 55 fuel

'

bundles elevated as a result of their channel fasteners being caught on the spent fue!

,

racks; 2) 14 fuel bundles are elevated due to unknown reasons, although it is suspected

)

that debris is in the fuel rack preventing proper seating; 3) One additional fuel bundle was

resting on a 1/4 inch metal tube (suspected to be a boron tube from a control rod blade

]

segment) that is lying on the floor liner and bends upward into the bottom of the fuel

1

storage cell. An evaluation was performed to address all relevant issues including: the

J

'

effect of a bundle drop on the fuel bundle itself, the fuel rack, and spent fuel pool liner;

,

i

seismic response of the fuel racks; the criticality margin; fuel assembly cooling; and water

i

shielding. The licensee concluded that the storage racks that contained elevated fuel

assemblies were operable, but were not full qualified, since the fuel assemblies were not

fully seated. In response to this concern, procedural controls were put in place to ensure

'

that no fuel assemblies are transferred within the spent fuel pool until all fuel is fully

.

seated.

The video tapes also revealed a significant amount of debris on the fuel bundles, fuel

{

racks, and the floor of the fuel pool. The debris included rope, cable, boron tubes, a broom

f

4

head, filter hoses, nuts, and unidentifiable objects. In addition, the bottom of the fuel pool

was covered by a layer of sediment. Additionally, the viden ^) view identified that the

velocity limiter portion cf '.our control rod blade assemblies vere stored vertically on top of

one another without support. The velocity limiter sections were located on the spent fuel

i

pool floor in the space between a spent fuel rack and the control rod blade storage rack.

An engineering evaluation of the velocity limiter storage configuration concluded that the

structural integrity of the spent fuel rack, the control rod blade storage rack, and the pool

liner would be maintained in the event of an impact caused by the velocity limiters falling

over.

A dent was identified in the spent fuel pool floor liner from an impact of an unknown

object. The dent was approximately 4 inches in diameter and was relatively uniform and

smooth, with no obvious nicks or gouges. A final operability determination and safety

evaluation concluded that there were no operability or safety issues, and that the dent did

not challenge the leak tightness or structural integrity of the fuel pool.

j

A previously known condition concerning the storage of a damaged, irradiated fuel

i

assemble stored in a " damaged fuel container," was roviewed to consider the collective

impact of this issue and the other fuel pool discrepan:les. The fuel assemble was

damaged in 1974, placed in a storage container in 1976, and moved to its current location

. ~-

..

-

-

.

.

.

- .

,

..

i

i

3

in 1989. The licensee performed a safety evaluation, as part of an operability

"i

determination, that addressed the storage configuration of the damaged fuel container and

its location in the control rod storage rack. Similar to the unseated bundles, criticality,

<

seismic response, water shielding, and decay heat removal were evaluated. Based on that

'

evaluation, the licensee concluded that the storage configuration is safe and does not

constitute an unreviewed safety question. In addition, a procedural restriction was placed

in procedure EN 1067, Supplemental Procedure for Inventory and Control of Special

Nuclear Material, to prevent storing fuelin locations adjacent to the damaged fuel

assembly. This was required since the criticality margin for storage of fuel assemblies in

adjacent rack locations has not been fully evaluated and full qualification has not been

verified,

c.

Conclusions

The inspectors concluded that the degraded conditions in the spent fuel pool are

,

representative of a long standing lack of concern for fore:gn material exclusion (FME)

'

during the conduct of refueling floor activities. Past low standards for FME control allowed

.

the accumulation of a large amount of debris, which could potentially have a significant

4

i '

impact en the fuel assemblies stored in the pool. Once the recovery organization became

aware of the extent of the problem, by reviewing the video tapes, the inspectors noted a

,

good resoonse, including clear direction as to what needed to be done in the short term.

,

The appropriate operability determinations and safety evaluations were prepared, and

adverse condition reports were initiated to document the findings.

In a letter to the NRC dated February 21,1997, the licensee documented the current

conditions in the spent fuel pool and their future plans for correcting the adverse

, conditions. Prior to core reload all unseated fuel assemblies will be properly seated and all

'

-

new and reload fuel bundles will be visually inspected from below to check for foreign -

material before placement in the core. The licensee committed to cleaning up the pool,

including removal of debris and various used components, prior to Refueling Outage 16.

Based on this information, the acceptability of the degraded conditions in the spent fuel

_

pool will be unresolved (URI 245/97-01-01) pending NRC review of the issues and the

l

completion of the licensee's root cause analysis.

U103

Operations Procedures and Documentation

03.1 Operations Procedures

During May and June,1996, two Unit 1 Operations department staff engineers performed

a self assessment of the Procedure Upgrade Program (PUP) for Unit 1 Operations

procedures. The Unit 1 operations self-assessment contained a significant number of

negative findings concerning the Unit 1 PUP process for the Operations Department and for

the quality of the upgraded procedures produced. The inspector discussed this assessment

with one of the staff engineers and the Unit 1 Operations Manager. Although the

assessment applied to Unit 1 only, the Unit 1 Operations Manager stated that he would

share the self assessment results with other Unit 1 Departments and with the operations

managers of the other Millstone units. The licensee's resolution of the problems identified

_. _

_

_

.

. _ _

_

_ _ . _ .

,

>

'

)

i

4

i

.

)

in the Unit 1 self assessment are unresolved (URI 97-01-02) pending the NRC's review of

the associated corrective actions.

,

U105

Operator Training Qualification

e

!

05.1 Inaccuracies in Personal Qualification Statements Certifications

)

a.

Inspection Scope

!

Two adverse condition reports (ACRs) were initiated to address operator license training -

related deficiencies. The ACRs document the failure of license candidates to complete all

classes, on the job training (OJT), and on shift watch standing time, along with the failure

to comply with procedures resulting in weaknesses in the systematic approach to training.

>

i

The issues were identified as a result of preliminary findings and insights gained from an

j

independent root cause investigation to address poor candidate performance during a

f

recent Millstone Unit 1 initial license examination. The inspectors reviewed the short term

1

!

actions taken in response to these two ACRs. The reviews focused on the accuracy of

Personal Qualification Statements (Form 398) submitted to the NRC staff as an application

for an operators license. The Form 398 contains assertions by the applicant, the training

coordinator and senior management representative on site, that among other things, the

,

}

applicant completed the licensee's requirements to be a licensed operator.

]

i

~

b.

Observations and Findinas

i

-

During the review subsequent to the initiation of the ACRs, the licensee identified

numerous discrepancies which resulted in inaccurate Personal Qualification Statements

-

(Form 398). In some cases, the errors resulted in candidates not meeting the licensee's.

minimum program requirements prior to signing of the 398 forms; however, the necessary

e. .

j

training was completed prior to the license examination. In other cases, the candidates

were issued licenses without the program requirements being met. The discrepancies

include failure to complete the required on shift watchstanding time, OJT, and the required

i

number of reactivity manipulations, in addition, several candidates failed to meet the

program prerequisites such as technical degree or additional experience requirements.

At Millstone Unit 1, the four most recent license classes were reviewed by the licensee, in

the two most recent classes,12 of 13 candidates submitted inaccurate 398 forms. In the

i

two prior classes, only 1 of 9 candidate's 398 form was inaccurate. Most of the

}

discrepancies involved the failure to complete required under-instruction watches, but also

included the failure to complete the required OJT. In the worst case, the candidate

completed little more than 3 of the required 13 weeks of OJT specified by the training

program description.

At Millstone Unit 2, a review of the two most recent license classes revealed 14 out of 16

candidates submitted inaccurate 398 forms. These discrepancies generally consisted of

insufficient hours of under-instruction watchstanding, but also included one case of

.

insufficient reactivity manipulations and two cases in which OJT records appear to be lost.

4

e

.

l

-

5

At Millstone Unit 3, the review of the most recent license class revealed 3 of 10 of the

candidates submitted inaccurate 398 forms. These discrepancies included one missed

under-instruction watch, one case of insufficient reactivity manipulations and the failure to

meet the program prerequisites, and one case in which OJT requirements were

accomplished after the assertion that all training program requirements were completed on

the 398 form.

At Connecticut Yankee, the review of the most recent license class revealed 10 out 12

candidates submitted inaccurate 398 forms. These discrepancies included insufficient

hours of under-instruction watchstanding, insufficient reactivity manipulations in two

cases, and program prerequisites not met in two cases. Additionally, OJT records were

lost or signed after the 398 was completed.

On March 3,1997, the licensee issued a letter to the NRC staff to discussing these issues.

Subsequently, the NRC staff issued a confirmatory action letter. The reviews are being

'

expanded on Millstone Unit 3 and Connecticut Yankee. Millstone Unit 2 is still evaluating

if review scope expansion is warranted and Millstone Unit 1 is preparing a position that

'

additional expansion is not necessary. The licensee has removed numerous individuals

from watchstanding duties and requested the withdrawal of two licenses. However, in the

case of Millstone Unit 2, some licensed operators removed frorn watchstanding duties have

been restored to an active status following the completion of missed requirements. The

licensee believes the majority of the discrepancies can be attributed to unclear

expectations on program requirements, the failure to maintain the programs currer t using

the systems approach to training, and poor record keeping practices. All of the fctors are

~!

the result of inadequate management oversight.

c.

Conclusion

The licensee identified numerous inaccurate Personal Qualification Statements (Form 398)

following NRC questions on two recent ACRs. Approximately two thirds of the Personal

Qualification Statements submitted for recent license applicants were inaccurate. These

applications resulted in the conduct of NRC license examinations and the issuance of

licensees. In a significant number of cases, the licenses were issued without the

candidates completing the licensee's training and qualification program, and in a few cases

the reactivity manipulations, specifically required by 10 CFR 55, were also not complete.

This issue is unresolved (URI 245,336,423/97-01-03) pending the completion of the

licensee's investigation, resolution of allidentified deficiencies and implementation of

programmatic corrective actions.

U108

Miscellaneous Operations issues (92700)

08.1 (Undate) Violation 50-245/95-42-01: Failure to Prevent Work Which Had the

Potential for Drainina the Reactor Vessel Durina Fuel Movements

This violation concerned the failure to prevent work which had the potential for draining

the reactor vessel while fuel removal was in progress. In addition, the licensee does not

have a formal process to ensure all applicable technical specifications are properly

implemented during refueling. Further, based on the inspector's review of the licensing

,

.

.. ,_ __

_ - - . . _

_

_ __

___

_

_

_

___

. _ _ _

.

.

d

-

6

.

bases for the current Technical Specification 3.5.F 7, it did not appear that the conditions

j

.

. initially established and reviewed by the NRC were appropriately maintained during

s"

subsequent amendments.

The licensee developed a process for preparation of corrective action completion packages

and a schedule for providing them to the NRC, as a result of a previous NRC request. The

inspector reviewed the first corrective action completion package prepared to address the

three issues discussed above. The documentation package contained a root cause analysis

-(RCA), license event report, and a violation response, which were developed to address the

3

issues. However, these documents were not consistent with each other and generally did

not address the cited violation. The identified causes and many of the corrective actions,

address maintenance and planning issues that led to the unplanned draining of a small

amount of reactor water during maintenance on a recirculation discharge valve. The RCA

i

appears to have been performed prior to the licensee's acknowledgment of the technical

'

,

specification compliance issue. The majority of the corrective actions specified, involve

I

improvements to the shutdown risk program; however, the inspector determined that these

actions would not preclude a recurrence of the technical specification non-compliance.

'

The violation response discussed the development of mode change checklists and

j-

enhanced logs; however, these actions, which may address technical specification

'

compliance, were not implemented by the end of the inspection period.

The licensee did not address the adequacy of Technical Specification 3.5.F.7, nor verify

that the conditions initially established and reviewed by the NRC were appropriately

maintained during subsequent amendments. The license event report was submitted 3

months after the event without a detailed reason for the delay and no corrective actions

i

4

'

were specified for the late reporting.

2

This item will remain open pending resolution of this item. The licensee failed to

0

appropriately evaluate and address these issues for more than a year since the event.

!

Further, the licensee failed to provide a closure package consistent with the process

committed to in December 1996. The manner in which this issue was addressed provides

evidence of the licensees ability to perform effective reviews and to implement appropriate

corrective actions. As a result of the inspector's concerns with the quality of the

completion packages, the licensee withdrew the NRC completion package schedule. A

..

revised schedule was stillin development at the end of the inspection period.

U1.ll Maintenance

U1 M2

Maintenance and Material Condition of Facilities and Equipment

i

i

M2.1 RWCU Demineralizer Room Material Condition

a.

insoection Scope (71750)

'

The inspector observed activities associated with the entry into, and video survey of, the

'

reactor water cleanup (RWCU) demineralizer room. The purpose of the entry was to

-i

i

determine the material condition of the infrequently accessed room. A remote controlled

robot was used, which supported both a video camera and radiation detection equipment.

.

.

7

b.

Observations and Findinas

The health physics (HP) department was well prepared for this activity since preparations

'

and staging were completed the day before. This allowed potential problems to be

identified and corrected prior to the start of work. In particular, it was identified in

advance that the robot would need to be lifted into the room through the access in the

wall, and preparations were made to account for this. Positive control over personnel

access, was observed with only people that were needed for the activity permitted in the

area. The workers' awareness of radiological hazards was evident. HP supervision and

the system engineer provided oversight of this activity. The video survey indicated that

the room was in good condition and the structural integrity of the piping and three

'

domineralizer tanks was intact. There was no indication of any system leakage and

radiation levels in the general area were normal.

c.

Conclusions

Based on the above review, the inspector determined that the preparation and conduct of

work associated with the entry and video survey of the RWCU demineralizer room was

well controlled. The inspector noted that good radiological practices were used. The

material condition in the room was acceptable and no deficiencies were identified.

U1.Ill Enaineerina

U1 E1

Conduct of Engineering

E1.1

Unresolved Safety issue USl A-46 " Seismic Qualification of Eauipment in Operatina

Plants."

a.

Insoection Scope (37550)

The scope of this inspection was to review the licensee's progress in resolving the outliers

identified during the implementation of the Unresolved Safety issue (USI) A-46 " Seismic

Qualification of Equipment in Operating Plants."

b.

Observations

The inspector reviewed the Licensee Event Report (LER)96-003, Rev. 2 that documented

deficiencies involving inadequate anchorage of the emergency diesel generator (EDG) day

tank and the turbine building secondary closed cooling water (TBSCCW) air coolers.

Backaround

in December 1980, The NRC staff initiated an Unresolved Safety issue, (USI) A-46,

" Seismic Qualification of Equipment in Operating Plants," related to seismic adequacy of

mechanical and electrical equipment in older nuclear plants. After technical research by

the Seismic Qualification Utility Group (SQUG) and the NRC regarding this issue, the NRC

.,__ _

_

_

- _ _

.-

.

_ . _ . _

_

. _ _ _ . . .

. _

_

.re i

,

.

'

7

b.

Observations and Findinas

i

The health physics (HP) department was well prepared for this activity since preparations

and staging were completed the day before. This allowed potential problems to be

identified and corrected prior to the start of work. In particular, it was identified in

"

advance that the robot would need to be lifted into the room through the access in the

wall, and preparations were made to account for this. Positive control over personnel-

,

access, was observed with only people that were needed for the activity permitted in the

j

]

area. The workers' awareness of radiological hazards was evident. HP supervision and

'

the system engineer provided oversight of this activity. The video survey indicated that

l

the room was in good condition and the structural integrity of the piping and three

,

l

demineralizer tanks was intact. There was no indication of any system leakage and

l

radiation levels in the general area were normal.

c.

Conclusions

t.

Based on the above review, the inspector determined that the preparation and conduct of

j

work associated with the entry and video survey of the RWCU demineralizer room was

well controlled. -The inspector noted that good radiological practices were used. - The

material condition in the room was acceptable and no deficiencies were identified.

,

I

U1.Ill Enaineerina

'

I

U1 El

Conduct of Engineering

E1.1

Unresolved Safety issue USI A-46 " Seismic Qualification of Eauioment in Ooeratina

,

Plants."

I

a.

Inspection Scope (37550)

l

The scope of this inspection was to review the licensee's progress in resolving the outliers

identified during the implementation of the Unresolved Safety issue (USl) A-46 " Seismic

i

Qualification of Equipment in Operating Plants."

b.

Observations

The inspector reviewed the Licensee Event Report (LER)96-003, Rev. 2 that documented

deficiencies involving inadequate anchorage of the emergency diesel generator (EDG) day

tank and the turbine building secondary closed cooling water (TBSCCW) air coolers,

i

Beckaround

in December 1980, The NRC staff initiated an Unresolved Safety issue, (USI) A-46,

" Seismic Qualification of Equipment in Operating Plants," related to seismic adequacy of

mechanical and electrical equipment in older nuclear plants. After technical research by

the Seismic Qualification Utility. Group (SQUG) and the NRC regarding this issue, the NRC

.

-

. .

_

.

.

. - _ . .

.

-

-

. -

- - -

.

1

8

4

>

published a detailed approach for resolving USl A-46, in Generic Letter 87-02, "Verif!:ation

of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors, USl A-

-

4 6. "

The Generic Letter procedure set forth an approach for verifying seismic adequacy of

i

equipment using earthquake experience data supplemented by test results and analyses, as

,

necessary. Licensees subject to USl A-46 were encouraged to participate in the generic

'

'

program to accomplish seismic verification of equipment. As a result, SQUG developed the

" Generic Implementation Procedure (GIP) for seismic verification of Nuclear Plant

equipment."

l

USl A-46 Proaram at MS1

At Millstone Unit 1, the USl A-46 program was conducted to address the concerns

expressed in GL 87-02 regarding the seismic adequacy of safety related electrical and

d

mechanical equipment. The resolution of the seismic adequacy issue appeared to be

conducted in accordance with the SOUG approach, using the generic implementation plan

(GlP) as approved by the NRC in Supplemental Safety Evaluation Report (SSER) No.2.

I

Supplement 1 to GL 87-02, transmitted May 22,1992 includes SSER-2 which reviews the

l

GIP, requires the licensee to identify within 120 days a schedule for implementation and

'

any anticipated deviation from the GIP methods. The inspector verified that the licensee

met this requirement by reviewing the licensee's letter dated September 21,1992. In this

letter the licensee identified a schedule for MS1, which stated that the submittal of the

final report will take place six months after refueling outage 15, which is stillin progress.

The inspector noted that the implementation of the USI A-46 review program has resulted

'in the identification of outlier conditions which challenged the operability of the plant

,

components. These outliers were reported in an LER. The LER described the proposed

corrective action, which includes resolution of all outlier conditions prior to start up from

Refueling Outage (RFO) 15.

The LER selected for this inspection addresses operability concerns involving inadequate

anchorage of the EDG day tank and the TBSCCW air coolers. These operability concerns

were properly documented as LER,96-003, Rev. 2. The inspector reviewed the LER 96-

003, to ensure that regulatory requirements for reportability were met. The licensee has

properly identified this design deficiency as a USl A-46 program outlier, and has properly

characterized it as being reportable in accordance with 10 CFR 50.72 and 10 CFR 50.73.

The inspector found the LER's event described in a chronological sequence and the

prescribed corrective action appeared to be appropriate.

Conclusion

in terms of reportability, the proper characterizations were given to the outliers of the USl

A-46 program. The licensee prepared the LERs documenting these outliers in accordance

with established regulatory requirements. Based on these inspection results, LER 50-

.

245/96-003 is closed. The followup of the licensee's commitments to resolve the A-46

program outliers prior to startup for cycle 16 operation and the assessment of the

,

_

_

.--

_ __ _ _

._

.

9

,

implication of this event on the operation of Unit 1.will be unresolved pending further.NRC.

.

. review (URI 50-245/97-01-04)

Review of the Corrective Action for the EDG Day Tank

,

Since the EDG system is an emergency ac power system and the EDG day tank is a safety

,

related component, the inspector focused his review on the licensee's corrective action

- package. In the package, the Design Change Notice (DCN) designed and installed the

seismic restraint that consists of a box frame around the tank supported by the block

wall's structural reinforcements.

f

The design modification package was found to be complete, and the frame was designed

l

in accordance with the American Institute of Steel Construction (AISC) Manual for Steel

Construction, 9th Edition. However, key design parameters in the calculation were not

properly referenced making it difficult for an independent auditor to determine whether or

l

l

not these parameters are correct. These key design parameters questioned by the

inspector included acceleration values, friction values between the day tank and the

concrete base pad; the calculated reaction of the block wall; and the shear capacity of the

block wall. All the inspector's questions and observations were properly resolved by thea

.

licensee.

d

1

Conclusion

.

The inspector determined that,the modification package was complete and the frame

properly designed.

Walkdown of the Modifications

The inspector and the licensee design engineer walked down the modifications for the EDG

day tank and the TBSCCW air coolers, with the following details.

With regard to the EDG day tank, the general area was inspected and the framing to

distribute the tank's load appeared to be structurally sound. The tank was inspected and it

was noted that the impact on the block wall (T-27B) was minimal and limited to shear in

the plane of the block wall. The inspector also noted that these block walls had been

previously upgraded in response to NRC Bulletin 80-11. Outside the EDG day tank along

the hallways the inspector noted that proper seismic bracing and anchorage was evident

on the following:

Several modifications to vital electrical equipment (switchgear, load centers and

motor control centers) were installed.

Modification to vital batteries consisted of shim material which was installed to

address A-46 outlier conditions.

Modifications to prevent seismic interaction between lighting fixtures and vital

equipment were installed.

--

-

-.

_

'

i

10

)

. At the turbine building, elevation 14'-6" the inspector noted that the TBSCCW air

. coolers identified in the USl A-46 program scope were modified to accommodate

-

seismic bracing and anchorage. in the EDG room, more air coolers and other

i

components (air-start tanks, motor control center, and control panel) were

j

seismically anchored or braced to address USl A-46 outliers.

Conclusion

Based on the detailed walkdown of the A-46 modifications, the inspector concluded that

the licensee has performed a substantial number of field modifications to accommodate the

seismic loading on mechanical and electrical equipment identified in the USl A 46 scope.

U1 E2

Engineering Support of Facilities and Equipment

'

E2.1

Adverse Condition Report (ACR) Review

i

a.

Inspection Scope (37550)

i

1

The inspector reviewed an ACR issued on August 16,1996, to assess whether appropriate

)

'

corrective actions were identified and implemented to prevent recurrence of the adverse

condition. The ACR (M1-96-0427) reviewed by the inspector concerned Component

l

Engineering Services / Nondestructive Test Engineering (CES/NTE) ultrasonic test (UT)

instruments that had exceeded their calibration due dates.

]

b.

Observations and Findinos

Six CES/NTE UT instruments were found to have exceeded their yearly calibration due

dates, with four of them having been possibly used during examinations at Unit 1. As

"

' stated on the ACR,'the person who previously handled the CES/NTE material and test

equipment (M&TE) program no longer workea for the company, and none of the job

functions were replaced or reassigned. As a result, there was no ownership of the

CES/NTE M&TE program.

The six instruments in question were sent offsite to a vendor for calibration. All six were

found to be within tolerance when they were received. Additionally, before and after each

examination was performed, the instruments were calibrated in accordance with procedure

NU-UT-1, using a step wedge or calibration blocx The missed yearly calibrations are

j

performed to verify instrument operability only, and do not represent a quality related

i

calibration. In other words, the calibrations done during examinations might be sufficient

to preclude sending these instruments offsite for yearly calibrations. Since these yearly

calibrations are not required by the ASME Code, the licensee will determine whether to

suspend them.

The inspector reviewed Quality Assurance Audit Package A-60607, " Measuring & Test

i

Equipment," an audit that was conducted from August 26,1996 through September 18,

1996. This audit evaluated the key elements and processes of the M&TE program and

determined that the program at Millstone and Connecticut Yankee was ineffective in

_

_.

,

1

.

l

11

fulfilling its mission and did not fully comply with 10 CFR 50 Appendix B criteria. in

- < .

response to this audit, ACR M1-96-0614 was written to address the adverse condition.'

1

i

i

To address the ACRs and audit report, CES/NTE developed a corrective action plan and

dedicated an individual to implement the plan and take ownership of the CES/NTE M&TE

program. Additionally, the licensee ensured that all CES/NTE quality related equipment

currently being used at Millstone and Connecticut Yankee was in calibration.

c.

Conclusion

The inspector concluded that the corrective actions taken by CES/NTE associated with the

ACRs and the QA audit was acceptable. The corrective actions appeared to be broad-

based. A majority of the short term corrective actions were complete, and the long term

corrective actions were being tracked for closure. The safety significance of the UT

j

instruments being past their calibration due dates was minimal because they were

subsequently found to be within tolerance.

E2.2 Containment isolation Check Valve.1-CU-29

a.

Inspection Scope (37551)

'

The inspector reviewed adverse condition report (ACR) 96-0539, which documents an

issue concerning the design specifications of the new containment isolation check valve 1-

1

CU-29. The valve was replaced with a smaller size valve for better flow characteristics,

and to allow testing and maintenance during this outage.

]

b.

Observations and Findinas

i

!

The replacement valve was a specially designed 6" check valve with a 4.5" disc. The disc

'

size was selected based on two flow conditions other than normal operations: 1) shutdown

flow conditions occurring approximately 10% of the time (70 days per operating cycle)

with a minimum system flow of 300 gpm; 2) and startup flow condition, occurring about

1

0.2% of the time with a system flow rate of 100-200 gpm. Due to the current extended

outage, it is not known if the shutdown flow rate will affect the valve. In fact, the RWCU

system is unable to produce the 300 gpm used in selecting the disc size due to the flow

restriction created by the pressure control valve 1-CU-10. During the followup of this

issue, the inspector noted that the actual system flow rate is approximately 190-200 gpm

using the auxiliary cleanup pump, it is not known what effects, if any, will be created by

i

operating the valve at these lower flow rates for this extended period.

In a effort to determine if the valve disc was fluttering or banging into the backseat due to

the low flow conditions, the licensee employed Liberty Technologies to perform non-

intrusive check valve testing. Liberty Technologies performed both acoustic and magnetic

testing on CU-29. Their report stated that the test results provide positive indication of full

opening during the flow initiation test. Acoustic data recorded during steady state flow

indicated no anomalous behavior (such as excessive trim wear or rattling). The report also

'

noted that due to large valve body signal strength from the magnetic sensors during the

--

,

_

_

.

12

valve opening and steady state tests were insufficient to provide useful results. However, .

they determined the acoustic data was a sufficient basis for the conclusions documented.

-

A member of the engineering staff informed the inspector that ACR 96-0539 was currently

open and under review. Engineering needs to determine a method for ascertaining the

actual impact of the lower flow operation on the valve. Localleak rate testing is being

considered as a possible method to determine if any valve degradation has occurred,

c.

Conclusions

The NRC is concerned about the operation of CU-29 with lower than expected flow rates

during this extended shutdown. While the use of non-intrusive check valve testing has

verified that the check valve is backseated, this is a short term indication. The long term

effects of the lower flow operation have yet to be determined. This issue is unresolved

-(URI 245/97-01-05) pending the NRC review of the licensee's finial determination of the

operability of the valve prior to plant startup.

U1 E8

Miscellaneous Engineering issues

E8.1

Closecut Documentation Packaae Review

The inspector reviewed the contents of a corrective action documentation package for NRC

unresolved item 96-04-07 (safety relief valve electric lift modification). The inspector

noted that the package included a draft licensee response to an NRC request for additional

information associated with the license amendment. Per previous arrangements the

licensee documentation packages were to be complete and contain only approved

documents. The package was returned to the licensee and no inspection was performed at

this time.

.

.

13

Report Details

Summary of Unit 2 Status

Unit 2 entered the inspection period with the core off-loaded. The unit was initially shut

down on February 20,1996, to address containment sump screen concerns and has

remained shut down to address an NRC Demand for Information [10 CFR 50.54(f)] letter

requiring an assertion by the licensee that future operations are conducted in accordance

with the regulations, the license, and the Final Safety Analysis Report.

U2.1 Operations

U201

Conduct of Operations

01.1 General Comments (71707)

Using Inspection Procedure 71707, the inspectors conducted frequent reviews of ongoing

plant operations to ensure that licensee's controls were effective in achieving continued

safe operation of the faciiity while shut down. The inspectors observed that proper control

room staffing was maintained, access to the control room was properly controlled, and

operator behavior was commensurate with the plant configuration and plant activities in

progress. In general, the conduct of operations was professional and safety-conscious.

Operations Management has recently placed greater attention on improving performance

associated with operator response to control room alarms with a focus on communications

and use of alarm response procedures. The NRC has noted the improvements in this area,

i

particularly regarding control room operators communicating to the unit supervisors what

alarms were received and ensuring a mutual understanding of why the alarm was received.

The licensee discovery of potential discrepancies in the personal qualification statements

(NRC Form 398) of certain Unit 2 licensed operators has been assessed for immediate

impact and determined to require further evaluation. This is considered an unresolved

,

issue as described in Section U1.05.1 of this inspection report.

The inspector toured the Unit 2 intake structure and found the material condition of

systems and components to be acceptable, in the Spring of 1996, the licensee instituted a

corrective action plan to address material deficiencies. A number of items have been

corrected, which has improved the material condition of the intake structure.

U2 O2

Operational Status of Facilities and Equipment

02.1 Adverse Condition Report Backloa

a.

Insocction Scope

The NRC evaluated the timeliness in which the licensee completed corrective actions

associated with Unit 2 adverse condition reports (ACRs).

.

.

14

b.

Observetions and Findinas

Timeliness for completion of corrective actions has been a longstanding concern at

'

Millstone. Having an ACR backlog in itself is not a reflection of poor performance because

as the threshold for writing ACRs decreases, the ACR backlog willincrease accordingly.

The concern is the number of ACRs that are not closed in a timely manner. To help

provide the NRC some sense of the licensee's progress in addressing the timeliness

concern, the licensee was asked to provide the number of ACRs having outstanding

corrective actions that are greater than 120 days old. Although the NRC does not consider

120 days a level of excellence nor is it acceptable when addressing immediate safety

concerns, it does provide some understanding of licensee management effectiveness in

addressing the corrective action timeliness issue.

Several months ago, the NRC raised a concern that the licensee's ACR database did not

allow them to determine the number of ACRs having outstanding corrective actions. The

1

licensee's previous understanding, as documented in NRC Inspection Report (IR) 50-

336/96-09, was that the ACR data entries had been corrected to provide reliable ACR

'

backlog numbers. However, additional licensee reviews of the ACR database indicate that

the number of ACRs greater than 120 days old as of December 31',1997, was 940 ACRs,

i

no_t 732 ACRs as stated in IR 50-336/96-09. The increase of 208 ACRS is based on a

licensee review of previously closed ACRs that they decided to reopen based on

incomplete closure documentation. At the end of the current inspection period (February

24,1997), there were 798 ACRs greater than 120 days old that have not been closed.

DEPARTMENT

ACRs OLDER

THAN 120 DAYS

Operations

56

Design Engineering

211

Technical Support (System Engineering)

254

Work Planning

28

Maintenance

55

_

l&C

42

Safety / Licensing

25

Other

127

TOTAL

798

c.

Conclusion

Although the backlog of 798 adverse condition reports (ACRs) that are greater than 120

days old indicates that timeliness for completing corrective actions continues to be a

,

>

15

concern, the reduction in this backlog of older ACRs from 940 to 798 since the last.

inspection period is a positive trend which reflects the licensee's increased level of effort in

this area. As discussed in NRC Inspection Report 50-336/96-04, timeliness and

effectiveness of corrective actions is an area in which the licensee must demonstrate

sustained improved performance.

U2 08

Miscellaneous Operations issues (92700)

08.1 (Closed) Violation 50-336/94-17-10: O_peration Outside System Desian Parameters

a.

inspection Scope

The sc-ope of this inspection included a review of Violation 50-336/94-17-10.

b.

Observations and Findinas

This violation involved the failure to correctly translate design basis temperature limits of

the service water (SW) and reactor building closed cooling water (RBCCW) 2ystems into

operating procedures. As a result, on May 24,1993, a reactor trip occurred when the SW

and RBCCW system temperature limits were exceeded during a main condenser thermal

j

backwashing evolution. This violation was previously reviewed in NRC Inspection Report

50-336/96-05 which concluded that the violation could not be closed because although

'

the specific procedures regarding thermal backweshing were adequately aabemd, the

corrective actions were too narrow in that they failed to address the possibility eat other

plant procedures did not insure operation was in accordance with the plant's oesign basis.

c.

Conclusion

Violation 50-336/94-17-10, which resulted from a 1993 unresolved item, reflects that the

failure to operate the plant in accordance with the design basis had been a longstanding

NRC concern. The licensee failure to address this type of concern eventually culminated in

their current extended shutdown and 10 CFR 50.54(f) effort which is intended to ensure

i

the plant is designed and operated in accordance with the licensing and design basis.

There are several outstanding violations including Escalated Enforcement items 50-336/96-

'

06-05 & 96-08-06 which also address plant operation that is inconsistent with the

licensing basis. Therefore, Violation 50-336/94-17-10 is being closed nql because

adequate corrective actions have been taken but because this concern is being addressed

and tracked by more recent items.

'

08.2 (Closed) LER 50-336/96-15, (Open) Unresolved item 50-336/96-01-04: Failure to

Enter Action Statement Reaardina the Number of Operable Nuclear instrument

Channels

a.

inspection Scooe

The scope of this inspection included a review of Licensee Event Report 50-336/96-15.

.

.

16

b.

Observations and Findinas

On March 12,1996, while the unit was shut down, the "B" train vital de bus was

inadvertently deenergized due to operator error. The resultant loss of various vital and non

vital power supplies overflowed the reactor building closed cooling water surge tank

through a failed open make-up valve, and challenged the operators to recover from this

complex event. The event was complicated by the fact that there was minimal procedural

guidance for operators to use to recover the bus. One train of shutdown cooling remained

in operation throughout the event and normal power was restored within four hours.

During the recovery, operators were required by procedure to deenergize bus VA-40, a

vital 120 Volt ac instrument panel which supplies channel "D" of the reactor protection

system (RPS). Operators were aware that a loss of RPS channel "D" would cause the

channel "D" wide range nuclear instrument to be inoperable. Prior to the loss of the "B"

train vital de bus, in an unrelated situation, channel "B" and "C" wide range nuclear

instruments had been declared inoperable. The action statement for Technical Specification (TS) 3.3.1.1 is applicable when less than two channels of wide range nuclear

instruments are operable. The action statement requires the immediate verification of

shutdown margin and at least once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter. The operators did not

~ recognize the need to erter the TS 3.3.1.1 limiting condition for operation when bus VA .

40 was deenergized.

LER 50-336/96-15 stated that a contributing cause for failing to enter the TS 3.3.1.1

action statement was that the shift was occupied with the restoration of the deenergized

dc bus and was considering the effects of deenergizing loads on plant operatico. The

licensee's corrective actions included training of operators to not only consider the effects

of their actions on plani operations, but they must also assess TS requirements, in

addition, operating procedures were changed to remind operators to determine if TSs are

affected when deenergizing an electrical bus.

c.

Conclusion

The failure of operators to enter the action statement for TS 3.3.1.1 when three channels

of wide range nuclear instrumentation were inoperable is considered a violation. This

licensee-identified and corrected violation is being treated as a Non-Cited Violation,

consistent with Section Vll.B.1 of the NRC Enforcement Policy. The primary concern

associated with tnis event was the fact that there was minimal procedural guidance

provided to operators to recover the loss of de bus. The licensee is in the process of

preparing 12 abnormal operating procedures for recovering various dc buses and

distribution panels. This concern is being tracked by Unresolved Item 50-336/96-01-04.

I

,

.

17

U2.ll Maintenance

i

U2 M8

Miscellaneous Maintenance issues (92903)

M8.1 (Closed) Unresolved item 50-336/96-04-09: Troubleshootina Controls

'

i

a,

lnspection Scope

'

in April 1996, the NRC resident inspectors identified programmatic concerns regarding the

U

conduct of troubleshooting. The concerns centered around the practice of performing

" troubleshooting" under the guise of " investigating" to avoid implementing the

administrative requirements for the performance of troubleshooting that are contained in

procedure WC-1, " Work Control Process." This issue was unresolved pending licensee

changes to WC-1.

b.

Observations and Findinas

in July 1996, the licensee issued Attachment 5.2 to procedure WC-1. This attachment

provides guidelines to be used in conjunction with a work order when a formal

.

troubleshooting plan is not required by Attachment 5 of WC-1.

'

The inspector reviewed the instructions contained in WC-1 and reviewed several work

I

orders that performed troubleshooting since the issuance of the change to WC-1.

c.

Conclusions

The inspector found that Attachment 5.2 of WC-1 contains appropriate directions to

ensure that all troubleshooting work is documented, supervision is consulted prior to

.

performing repair or replacement of components and retest requirements are determined

following the completion of the troubleshooting. No problems were identified during the

review of the troubleshooting work orders. This item is closed.

I

M8.2 (Closed) Unresolved item 50-336/96-06-06: Hioh Pressure Safety Iniectiori Check

Valve Backflow Testino

a.

Insoection Scope (92903)

The inspectors identified that the licensee had unnecessarily relaxed the frequency of

backflow testing of high pressure safety injection system (HPSI) pump discharge check

valves. The licensee agreed to revise the surveillance procedure to perform quarterly

backflow testing. This issue was unresolved pending NRC review of the planned

procedure changes,

b.

Observations and Findinas

The licensee revised procedure SP 21136, " Safety injection and Containment Spray

System Valves Operational Readiness Test," to include quarterly backflow testing of the

HPSI pump discharge check valves.

,

. --

i

'

18

i

c.

Conclusions

'

The NRC concluded that the licensee had appropriately resolved the check valve testing

concern in Revision 10 of procedure SP 21136 and the associated data forms. This item is

closed.

U2.lll Enaineerina

'

U2 E8

Miscellaneous Engineering issues

i

E8.1 LClosed) Unresolved item 50-336/95-07-06: Condensate Storaae Tank Siohon Break

a.

Inspection Scope (92903)

On February 10,1995, the licensee discovered that the condensate storage tank (CST)

level had dropped to approximately 30% due to a heat exchanger tube leak. At the time of

the event the plant was defueled and no minimum tank volume was require by the plant.

technical specifications. During an investigation of the inadvertent loss of CST inventory

the licensee discovered that a siphon break (a 1/2 inch hole) in the tank recirculation piping

was missing. This issue was unresolved pending further review of how the plant design

change process missed the removal of the siphon break,

b.

Observations and Findinas

Plant information Report (PIR) 2-95-174 documented the licensee's investigation of the

missing siphon break. The licensee's review concluded that the section of piping, in which

the siphon break was located, may have been removed and replaced during a modification

performed in 1992 to install a CST nitrogen blanketing system. The modification included

the addition of a stiffener beam on the internal tank wall that required the recirculation loop

suction piping to be modified to provide clearance for the beam.

The interference problem between the stiffener and the recirculation piping was not

identified during the initial modification design. When the problem was identified a design

change notice (DCN) was processed to provide details on how to modify the piping.

However, the drawing provided with this DCN did not depict the existence of the siphon

break hole.

The work description in the work order that modified the piping was to " fabricate and

installinternal piping to the CST per drawing #25203-13006 Sheet 44". Although only

the section of piping at the elevation of the stiffener beam was affected, the work order

did not preclude the replacement of the section of piping that contained the siphon break

hole, nor did the drawing provide the details necessary to drill the hole in the event the

piping was replaced.

The cause of the event appears to be inadequate attention to detail during the preparation

of the DCN and the associated drawing since other drawings (the system piping and

-

_

I

.

19

instrumentation drawing and the original piping isometric drawing) showed the existence of

a siphon break hole.

The licensee briefed the design engineering staff on this event and the need to pay

particular attention to less obvious design attributes such as siphon breakers when

developing design changes. Also, since the time of this event, the licensee has

implemented programmatic improvements to the design control process to improve the

effectiveness of the process.

The inspector noted that the plant technical specifications require the maintenance of a

minimum volume of 150,000 gallons of water in the CST when in modes 1,2 or 3 and the

volume is verified every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. These requirements reduce the possibility of a

significant, undetected loss of CST inventory when the plant is in operation.

c.

Conclusion

l

The inspector concluded that the licensee had taken appropriate actions to resolve this

issue. However, the inspector noted that the licensee evaluation could have been more

thorough in that the investigator did not review the associated work order until questioned

by the inspector. A review of the work order was necessary to determine if the event may

have been a result of poor work controls, which has been a problem in the past at this site.

This item is closed.

E8.2 LClosed) Unresolved item 50-336/95-11-03: 10 CFR 21 Reportability Review

a.

Inspection Scone (92903)

Following the licensee identification of several design problems that affected replacement

components of the engineered safeguards actuation system (ESAS) cabinets, the NRC

inspectors questioned if the findings had been reviewed for reportability under the

requirements of 10 CFR 21, " Reporting of Defects and Noncompliance." At the time of

the inspection in 1995, no formal review had been initiated and the issue was unresolved

pending further actions to be taken by NU, and NRC review of the licensee actions.

b.

Observations and Findinas

in June 1995, the licensee performed an assessment of the design problems experienced

with the ESAS system during the previous two refueling outages. The assessment

identified four significant design problems that were related to defects in the components

provided by the vendor. The four problems were assessed for reportability in engineering

evaluation M2-EV-97-0004, Revision 0, " Evaluation of URI 95-11-03, Reportability of

ESAS Design Deficiencies."

Three of the four issues had been reported to the NRC in Licensee Event Reports (LERs)

94-12-00, 95-18-00 and 95-21-00. The issue associated with LER 94-12-00 had also

been reported by the vendor ~(Eaton Corporation) in accordance 10 CFR 21. The fourth

issue was documented in ACR 00506 and the licensee had determined that the issue was

not reportable.

,

_.

_

.,

i

!

.

20

The engineering evaluation notea that problems reported in an LER did not require

additional reporting in accordance with 10 CFR 21. The bases for an LER fulfilling the

{

licensee's reporting obligations under 10 CFR 21 is contained in 10 CFR 21.2(c).

'

The inspector also noted in discussions with the licensee that the components that were

'the subject of LERs95-018 and 95-021 were designs that were unique to Millstone Unit 2.

c.

Conclusion

The inspector reviewed the licensee evaluation, LERs and ACR and found that the licensee

appropriately evaluated and reported the failures to the NRC. This item is closed.

,

1

l

1

l

j

- .

__ __

_

_

__ __

&

i

n.

21

,

.

Reoort Details

Summary of Unit 3 Status

Unit 3 remained in cold shutdown (mode 5) status throughout the inspection period. The

licensee continued its implementation of the Millstone Unit 3 Recovery Plan and the

'

configuration management program activities in support of the milestones leading to the

readiness for the unit restart. In accordance with commitments made to the NRC with

regard to corrective action progress and documentation of the completed work items, the

licensee provided the first set of corrective action completion packages for NRC review.

To date, the presentation of such packages to the NRC has been timely,. relative to the

scheduled workload. This documentation has also provided evidence of progress in the

resolution of open NRC inspection items, as well as an indication of the licensee efforts to

demonstrate corrective action program effectiveness. NRC review of the available closure

)

packages will continue as an ongoing process, with the individual technical issues

,

discussed, as appropriate, in the following report sections and in future NRC inspection

reports.

On January 22,1997, the appointment of Mr. M. H. Brothers, then the Unit 3 Director, to

the position of Vice President-Millstone Unit 3, was announced, in this new position, Mr.

-

j

Brothers fills the role of Recovery Officer for the unit. On January 28,1997, Mr. Brothers

announced the appointment of Mr. G. D. Hicks, who had been serving on the Carolina

Power & Light Recovery Team for Unit 3, to the position of Unit 3 Director, in an acting

capacity. Both of these managerial changes became effective on February 3,1997. The

inspector noted that Mr. Hicks' qualifications to assume the Unit 3 Director position had

been reviewed by both the licensing department and the Nuclear Safety & Oversight

organization. -The inspector also reviewed section 6.3.1 of the unit technical specifications

and American National Standard, ANSI N18.1-1971, regarding " Plant Managers" and -

?

d

identified no qualification concerns or other questions regarding these licensee

management changes.

U3.1 ODer8tions

U3 01

Conduct of Operations

01.1 General Comments (71707)

Using inspection Procedure 71707, the inspectors conducted frequent reviews of ongoing

plant operations. During a walkdown of the Unit 3 intake structure, the inspector observed

large scale painting and materialimprovement work in progress. Various degrees of work

activity have been on-going in the intake structure since the licensee instituted a material

condition improvement project in the Spring of 1996. The inspector noted that adverse

condition report M3-97-0370, dated February 1,1997, was written to report that station

air and instrument air supply piping in the intake structure was in poor extemal condition.

The licensee is planning to replace or paint the piping. In general, the conduct of

operations was professional and safety-conscious; specific events and noteworthy

i

observations are detailed in the sections below.

e

, _ ,_ . _ _ _ . . -

. _ _ ~

. -. _ _ . _ _.. _ ._ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ . _

'

i

'

.

1

.

j

22

i

!

< Over the course of this inspection, the inspectors witnessed and/or reviewed a number of

b

,. operational activities, and noted the following observations and assessments of operations

!

!

performance:

Good contingency planning (e.g., preparation for field flashing the "B" emergency _

f

diesel generator during the Battery 2 outage) was in evidence.

-t

'

Appropriate regulatory provisions (e.g', a technical specification [TS] " bases"

~

-

j

change to allow operation of the safety injection pumps in mode 5 to fill an

,

accumulator) were considered and dispositioned to address emergent operational

conditions.

j

m

[

Consideration of how the incore flux mapping results affect the quadrant power tilt

I

• -

ratio (OPTR) was found to be consistent with both the TS definition for QPTR and

f

the Westinghouse position statement on " core tilt".

j

-

Evaluation of the calibration provisions for the range and accuracy of digital

l

-

j-

instrumentation utilized in the performance of operational surveillances was

i

j-

>

determined to meet the NRC guidance discussed.in NUREG-1482.

4

,

i

j

Reductions in the range of reactor coolant system operating temperatures (i.e., T-

avg) to maintain margins relative to nil ductility transition (NDT) considerations were

!'

implemented consistent with both TS 3.4.10 and ASME Code requirements.

!-

The implementation of temporary modifications (e.g., a bypass-jumper for cross-

j

.

i

tying trains in the auxiliary feedwater system flow paths) was determined to be

j,

conservative in providing additional heat sinks (i.e., two steam generators) for

mshutdown risk considerations.

i

.

i

The licensee discovery of potential discrepancies in the personal qualification-

)

l

statements (NRC Form 398) of certain Unit 3 licensed operators has been assessed

for immediate impact and determined to require further evaluation. (NOTE: See

j

unresolved item associated with similar Unit 1 activities - Section U1.05.1 of this

i

h

inspection report)

4

Additionally, during control room inspections and reviews of TS limiting condition for

-

operability (LCO) action statements, the inspector raised a question regarding the

!

applicability of actions for single system / train inoperability when more than one system or

train is determined to be inoperable. Examples where the need for interpretation of the

i

4

LCO actions might be appropriate for multiple system or component unavailability were

$

I

identified in TS 3.7.7 and 3.7.1.2. The inspector also noted recent correspondence (i.e., a

'

memorandum dated January 17,1997) from the Unit 3 licensing staff providing a

,

clarification of TS terminology, e.g., how to interpret "at least once per 7 days". The

j

inspector determined that the licensee needed to further develop its approach to

4'

promulgating such interpretive guidance. While no TS violations or technical concerns

j

were evident in the areas of TS compliance questioned by the inspector, a standardized

method for disseminating Unit 3 policy in the interpretation of TS language and LCO

gj'

actions appears prudent. The inspector discussed this issue with cognizant licensee

.'

\\

i

,

&

"

,

_

,

.

23

personnel and intends to review this matter further as an inspector followup item. (IFl

423/97-01-06)

01.2 Control of Hiah Enerav Line Break (HELB) Doors

a.

Insoection Scooe (71707)

On several occasions during this inspection period the licensee identified that HELB doors

were open or not fully latched. These conditions were reported in accordance with 10 CFR 50.72 as a condition that could have prevented a safety system from functioning as

required. The inspector reviewed the licensee's corrective actions to assess the

effectiveness of the licensee's root cause determination and whether appropriate corrective

actions were identified and implemented to prevent recurrence of the adverse condition.

b.

Observations and Findinas

,

On January 10,1997, the licensee reported that the HELB door to the "A" train 4160 volt

switchgear room was open. The door had been opened to facilitate battery 301 A-1

,

replacement. Within the next few days, four additional occurrences of an open or

improperly latched HELB door were identified, including one by the resident inspector. All

the doors had a HELB sign affixed on both sides of the door.

As a result of these incidents, a level "B" adverse condition report (ACR) was gererated

i

and an event review team assembled. As an immediate corrective action, a work stand

-

down was held within all departments and licensee management briefed employees on the

control of plant doors. During the next several days, several other ACRs regarding HELB

issues were generated; including doors not being appropriately labeled in the field or on

prints.

The licensee's root cause investigation detarmined that the cause of the events was a

failure to develop and implement a HELB door control program, which resulted in a lack of

understanding of the HELB requirements associated with the plant design basis. Corrective

actions included: develop a door control program, insert a door control training module in

general employee training, and simplify and label all HELB doors. Although not specifically

stated in the corrective action plan, the licensee indicated that door labels would specify

the number of turns required to latch the doors.

As part of the investigation, the licensee concluded that the HELB door issue was not a

reportable condition with the plant in modes 5 or 6. Final Safety Analysis Report, Section

3.6.1 states that a high energy system is a fluid system that operates during normal plant

operating conditions. Normal plant conditions are defined as startup, operation at power,

hot standby, or reactor cooldown to cold shutdown conditions. Therefore, those

conditions that were identified while in mode 5 were retracted. However, the licensee

concluded that the status of HELB doors should be considered in the shutdown risk

program since a rupture of the hot water heating line in the service building could result in

the loss of safety-related equipment located in the switchgear room, independent of the

plant mode.

.

.

24

The inspector toured the plant and verified that HELB doors 'nere properly latched.

Applicable HELB doors were labeled to indicated the required number of turns to properly

latch the door. The inspector also verified that a training module was being developed for

inclusion in the general employee training. The other corrective actions are scheduled to

be complete prior to the unit entering mode 4.

c.

Conclusion

The licensee's root cause investigation and corrective action plan for control of HELB doors

were determined to be good. However, the requirement to label the required HELB doors

with a minimum number of turns to ensure prope- U.ching should have been included in

the ACR corrective action plan if it was deemed necessary to prevent recurrence.

U3 03

Operations Proceduies and Documentation

a.

Inspection Scope

i

The purpose of this inspection was to determine the adequacy of the procedure upgrade

program (PUP) as it applies to Unit 3. The licensee started the PUP in 1992 to standardize

procedure format for all units on the site and to improve the technical adequacy of all

j

procedures. The process was a third iteration of previous procedure improvement

'i

programs which were started in the late 1980's. This inspection was performed from

j

August 1996 through February 24,1997.

The onsite inspection included interviews with the site PUP group, Unit 3 procedure

coordinators and procedure writers, station oversight group, station quality assurance, and

Unit 1 operations personnel. Documents reviewed included, but were not limited to,

document control (DC) procedures: DC-1, " Administration of Millstone Procedures and

Forms"; DC-2, " Developing and Revising Millstone Procedures and Forms"; DC-3,

" Verification and Validation of Millstone Procedures and Forms"; and a sample uf Unit 3

procedures which had already been upgraded. in addition, the inspector reviewed the

Millstone Unit 3 PUP self assessment of their Operations Department (conducted March-

'

May,1996); Millstone Unit 1 PUP self assessment of their Operations Department

,

(conducted May-June,1996); Station oversight audit conducted during September 1996;

and various procedure related adverse condition reports (ACRs).

Although the focus of this inspection was Unit 3, the Unit 1 self assessment was reviewed

and discussed with Unit 1 personnel who performed the assessment and the Unit 1

Operations Manager. The inspector considered its potential applicability to Unit 3.

b.

Observations and Findinas

,

At the start of this inspection, there was a station wide Procedure Upgrade Group to

provide overall control of the program. This group developed and maintaineci the station

DC procedures for control of the program, the overall status of upgraded procedures,

coordinators for each Millstone Unit, and the hiring of contractors, as necessary, to write

the procedures. The actual upgrade of procedures was the responsibility of each

department within each unit. Since the licensee's reorganization in October 1996, the

.

.

25

station PUP group has been decentralized. The group now controls the station

administrative procedures including the PUP DC procedures. It has no control of the

production of upgraded procedures. Despite the changes in PUP control, the quality and

quantity of upgraded procedures has depended on the individual technical departments in

each unit.

A review of the program as it applies to Unit 3 noted the following:

The program has been effective in standardizing procedure formats. The document

control procedures are lengthy and cumbersome to use, but appear to be

comprehensive. The quality of the upgraded procedures appears to depend on the

producers of the procedures and the adequacy of the performance verification and

validation (V&V) process rather than on any apparent process deficiency.

1

The V&V process can be by table top review, procedure walkdowns, or by

i

procedure performance. All three methods are used, but the table top review by a

technical peer is the most common form of validation for Operations and

Maintenance Department procedures. Instrument & Control procedure technicians

stated that when possible their V&V process consisted of procedure walkdowns. -

In general, management involvement in the upgrade process seems to be minimal.

Some department managers have more involvement than others. There is heavy

reliance on the procedure coordinators for each department. For example, the Unit

,

3 operations manager authorized his procedure coordinator to act on his behalf for

procedure review and approval.

,

The inspector noted that during the five years that the PUP has been in place, there

had been no Quality Assurance (QA) audits of the program itself. Unit 3 Operations

Department had performed a self assessment of the PUP process in May 1996, but

this assessment was fairly limited. As a result of this NRC inspection, the licensee

performed two further assessments of the PUP program as it relates to Unit 3.

An assessment of the verification and validation process by Nuclear Oversight for

Unit 3 was performed in September 1996. This assessment noted some

weaknesses and problems in the V&V process. Another assessment was performed

on the PUP program for Units 1,2 and 3, November 13-16,1996, by the Region

One Procedere Working Group. This group is composed of persons from other NRC

Region i nuclear utilities. This assessment also noted strengths and weaknesses.

Both assessments noted deficiencies in the PUP program but did not conclude that

the program was seriously flawed.

On February 19,1997, the licensee forwarded to the inspector three recent QA

audits and fourteen quality control surveillance activities performed in the last

quarter of 1996 and the first quarter of 1997. These audits and surveillance

indicated that procedures are routinely reviewed as part of that activity and, where

appropriate, procedure deficiencies are identified. Also forwarded were recent

adverse condition reports identifying procedure problems.

..

.

.

26

As part of the PUP, the licensee developed the procedure basis document. The

Intent'of this document was to duplicate the procedure and to add blocks at certain

points to indicate the source or basis for key technical information such as the Final

~ Safety Analysis Report (FSAR), vendor. manual, technical specification, regulatory

commitments, etc. The Region One Procedure Working Group report considered the

basis document a strength, inspector's conversations with many plant personnel

,

indicated that they had high expectations from this document.

_

4~

' The inspector reviewed numerous basis documents during the course of this

'

inspection. While the basis document concept generally appears to be a strength,

l

the inspector noted numerous basis documents to be incomplete. The documents

l.

were essentially the procedure with one or two basis blocks added in a pro forma

j'

matter. Documents referenced in both the procedure and basis document were not

further identified in the basis document as to where the referenced material actually

i

applied. This appears to be's weakness in the practical usage of basis documents.

.

1

The licensee does not, however,' use the basis document as a procedure for plant

operations.

I

j

^The inspector did a review of some Unit 3 operations procedures with a Unit 3 staff

7e

j

engineer. Some minor technical (Hecrepancies were observed,' but were not

l

considered by the inspector to be' safety significant. In some cases, the operations

j

. engineer had difficulty in determining the source for specific information in certain

j

procedures. For instance, the calculation for one instrument setpoint was only

i-

available in the desk drawer of an l&C engineer; and the source of the setpoint was

j

not in the basis document. This was an example of one of several instances of

apparent configuration management control. However the configuration

-

i

management problem has already been identified to the licensee and is being

addressed generically by the licensee.

As noted in previous NRC inspections, design basis discrepancies have been

identified at all three Millstone units. It is possible that some procedures may not

conform to the FSAR and other procedures may conform to FSAR conditions which

are contrary to the actual design basis. As a result of an NRC 10 CFR 50.54(f)

letter, the licensee is currently performing an extensive design basis review which in

turn will be independently verified. In a letter from the licensee dated July 22,

'1996, the licensee stated, in part, the following: "...As we believe was

communicated [to the NRC in a meeting conducted on April 30,1996], there was

no commitment to complete the PUP as a conditio 1 of restarting any of the

[ Millstone] units. As part of the Operational Readiness Plan for Millstone unit No. 3,

findings resulting from the 10 CFR 50.54(f) related work will be reviewed to

determine if any procedure modifications are required prior to restarting the unit.

This will be done independent of procedure upgrades completed via the PUP..."

The licensee's corrective action plan for any required " procedure modifications" will

be assessed in the future as part of the NRC restart assessment plan,

The inspector observed that Unit 3 Operations Department had established a 67

page handbook in order to implement the DC procedures for the PUP process. This

appeared to be an uncontrolled, unapproved and unofficial procedure. The licensee

.

1

.

27

stated at the conclusion of the onsite inspection that this handbook would be made

into an officially controlled procedure. By telephone on February 26,1997, the

inspector was informed that the operations handbook had been deleted and only the

DC procedures were being used by the Operations Department as guidance for the

PUP process.

.

From January 13 through January 30,1997, the licensee's Nuclear Oversight

Group conducted an audit in the areas of document control and the maintenance of

l

quality records. This audit identified a site-wide breakdown in the control of

j

procedures in that procedures in use had not been properly updated. The

operations handbook (now deleted) was an example of an uncontrolled procedure in

use. Because of its magnitude, the adverse condition report (ACR) generated by

this audit was initially recommended for classification as a Level "A" ACR.

c.

Conclusions

l

The procedure upgrade program meets regulatory requirements and has been effective in

j

standardizing procedure formats across the site. The technical adequacy of upgraded

procedures, except for a small sample of Unit 3 operations procedures, was not a subject

for this inspection. Because of the number of individuals involved in procedure upgrade

and long period of time to upgrade the procedures (5 years), the quality of procedures

vary. A number of ACRs reviewed indicate technical problems with some procedures

already upgraded. The licensee has committed as part of their 10 CFR 50.54(f) process

and their Configuration Management Prograin to ensure procedures will meet the applicable

~

design bases. The licensee's basis documents, as implemented, did not appear to meet

the licensee's own expectations as an adequate foundation for future procedure reviewers

to have an adoquate technical basis for the information contained in each procedure.

U3 O'

Quality Assurance in Operations

07.1 General Comments (40500,92901)

The inspector reviewed station procedures, assessed planned program changes, and

discussed various quality assurance activities with representatives from the Nuclear Safety

& Oversight (NS&O) organization and Unit 3 licensing, engineering and operations

departments. The following topics were generally reviewed and evaluated during the

conduct of this inspection:

corrective action program changes (Revision 4 to the station procedure, RP-4,

addressing " Corrective Action", effective February 25,1997)

job rcAation between the Unit 3 line departments and the NS&O organization

engineering assurance recovery activities, conducted by the NS&O organization, to

include planning and preparation for an Integrated Assessment Plan imp!ementation

and review of the proposed changes to the Design Basis Document Package

upgrade program

.

.

I

28

issuance of a " Millstone 3 Line/ Oversight Interface Agreement"

implementation of and subsequent release from a Quality and Assessment Services

+

(QAS) " Hold" of all work on safety-related eqaipment requiring the use of "non-QA"

parts

j

With the progress of the noted program revisions, recovery activities, and organizational

'

initiatives still ongoing, the impact and effectiveness of the changes have not yet provided

i

measurable results. As of the end of this inspection period, the inspector observed

increased NS&O involvement in performance monitoring, interfacing analysis, and support

of the Unit 3 rnanagement and line staffs. Such involvement has included "real time"

evaluation and feedback on routine operational activities and nonroutine' events. NRC

assessment of NS&O effectiveness (including an expectation of demonstrable results of the

,

corrective action program improvements) and specific QAS activities (e.g., Hold 97-1,

Revision 1) will continue over the course of the next severalinspection periods; covenng

the ongoing recovery, open item closure, and work associated with the startup planning for

the unit.

i

U3.Il Maintenance

l

USM1

Conduct of Maintenance

M 1.1 General Comments

I

a.

Insoection Scoce (62707/61726)

The inspector observed / reviewed all or portions of the following maintenance and

surveillance activities to verify proper calibration of test instruments, use of approved.

+

procedures, performance of work by qualified personnel, and conformance to technical

specification (TS) limiting conditions for operation.

M3-96-15203,

Battery 301 A-1 Removal

M3-96-25108,

Calibrate ITT Barton (SP 3481B01)

SP 3626.8,

" Control Building Air Conditioning Booster Pump,3SWP'P2A,

Operational Readiness Test"

The inspector found the work performed under these activities to be professional and

thorough. All activities observed were performed with the work package or surveillance

procedure present at the job site and personnel were noted to be closely following the

procedures. Review of the surveillance procedures revealed that the requirements of the

'

applicable TS were appropriately incorporated into the implementing procedure,

,

_ _

_

_

.

29

M1.2 Fix-It-Now (FIN) Conduct of Maintenance

a.

Inspection Scope (62707)

^

The licensee has implemented a number of changes to the work control process including

the development of a FIN multi-discipline work team approach to augment the way

maintenance is performed at the unit. This process is in addition to their normal work

control process. The inspector reviewed the FIN maintenance procedure and monitored

work that was performed under the FIN process to assess its implementation.

b.

Observations and Findinas

<

As part of the Unit 3 recovery process, the licensee developed a FIN work process to

reduce the time it takes to respond to operations department needs, and to perform work

more efficiently. In addition, a minor maintenance work process was developed to allow

,

qualified personnel to accomplish tasks in a more efficient manner. It was envisioned that

these work processes would contribute to reducing the corrective maintenance backlog.

These work processes were initiated in November 1996.-

Maintenance procedure MP 3705B, "Fix It Now Conduct of Maintenance," states that the

FIN work process shall not be implemented for work requiring: major plant modifications,

lengthy tag clearances, special radiological work permits that cannot be dispositioned by

the health physics (HP) team member, or repairs involving welding on specified plant

equipment. The minor maintenance procedure U3 WC 1.1, " Minor Maintenance Process

"

Controls," allows work to be performed without the generation of a work order if the work

i

is performed on non-QA equipment and the work doesn't impact plant operations or require

special work control needs.

,

i

The FIN team is comprised of mechanics, electricians, maintenance planners, and

'

instrument and control (l&C) technicians. In addition, there is a representative from the HP

and operations department, and an assigned first line supervisor (FLS). The inspector

reviewed the training records for each FIN team member and verified that they were trained

,

on the new procedure. In addition, team members appeared to be well qualified. At least

i

one member of the team was qualified for each of the matrices job tasks.

Each morning the FIN team members review all the trouble reports (TRs) that were

generated the previous day and determine which items can be performed by the team and

which need to go through the normal maintenance process. At the 6:45 a.m. morning

meeting, the FIN team FLS notifies the operations shift manager and the work planning

organization of the selected work items to ensure that these departments are cognizant of

all FIN planned maintenance activities.

The inspector attended the FIN and the 6:45 a.m. morning meeting. In addition to the off-

going and on-coming shift managers, individuals at the 6:45 a.m. morning meeting

included: the Unit Director, the operations, maintenance, l&C, and engineering department

managers, and representatives from work planning, chemistry, and HP. Each TR is

discussed then assigned to either the FIN or work planning department. Any TR that

potentially affects equipment operability is identified and an adverse condition report

.

l

.

30

generated. The inspector noted that all work on the protected train was assigned to the

work planning department.

There have been approximately one hundred TRs generated each week during the month of

January 1997. Of these, the FIN team completed approximately fifty-five percent. A

review of the TR backlog, those over one week old, revealed that the number has been

declining. The inspector monitored selected activities and reviewed those work items that

were performed by the FIN team for the month of January. A review of the work activities

revealed that work orders had been generated for all safety-related work activities in

accordance with procedure U3 WC 1.1. As a result of the high number of TRs being

generated and the minor maintenance work activities performed, the FIN team has been

unable to work off any corrective maintenance backlog items.

c.

Conclusions

The licensee is implementing the FIN work process in a conservative manner. Any work on

the protected train equipment is not being assigned to the FIN team. All monitored work

activities performed by the FIN team was performed in accordance with the unit and

station procedures. FIN team members appeared to be well qualified. No safety concerns

,

were identified from the specific activities observed.

U3 M8

Miscellaneous Maintenance issues

M8.1 Plant Insoection-Tours (62707. 92902)

The inspectors conducted inspection-tours of several areas of the plant during this

inspection period, observing work in progress and raising some questions regarding

completed field installations. As appropriate, discussions were held with workers, field

supervisors, and support personnel (e.g., health physics technicians). While most field

observations and questions were resolved prior to completion of the inspection-tours, the

following issues required followup, as documented below:

snubber removal on a residual heat removal (RHS) line in the engineered safety

features building; authorized by plant design change record (PDCR) MP3-90-003.

The inspector reviewed the applicable PDCR and design change notice DM3-P-154-

90, verifying proper re-analysis of the RHS piping system and control of the snubber

elimination list. Since design criteria discussed in ASME Code Case N-411 were

used in the pipe stress re-verification, the inspector reviewed the related discussion

of seismic design response spectra, provided on NRC Regulatory Guides 1.60 and

1.61, in the final safety analysis report (FSAR); and confirmed NRC approval for the

use of ASME Code Case N-411 at Unit 3.

white " frothing" of oil observed in the site glass for the speed increaser on the "B"

charging (CHS) pump, located in the auxiliary building.

The inspector discussed this observation with the responsible system engineer, who

confirmed that the subject " frothing" was likely due to the turbulence caused by the

.

.

31

meshing of gears in the CHS pump speed increaser. The inspector reviewed the

results of the most recent chemical analysis performed on this pump and identified

no adverse conditions or additional concerns.

treatment of reactor plant sampling (SSR) tubing runs and flexible hose connections

inside the containment building as ASME class 2 components, as discussed in the

Unit 3 FSAR.

.

The inspector reviewed the fabrication installation control drawings for the SSR

piping and common header connections from the containment penetrations to the

i

steam generator blowdown lir.es (i.e, a review of approximately 60 Stone &

Webster Engineering Corporation isometric drawings); and confirmed proper ASME

j

Code classification of the subject sample lines.

1

an unrestrained trolley assembly, located on a structural beam at the lower elevation

j

(-24'6") of the containment building, in proximity to some safety-related trisodium

1

phosphate baskets.

Inspector followup of the status of this assembly revealed that the trolley had been

installed as a temporary component during the first refueling outage in 1987, but

never removed. The licensee removed the unauthorized trolley and issued an

adverse condition report (ACR) M3-97-0563, documenting the concern that current

procedural controls for " incomplete work" were not being followed. Subsequently

(note: after the conclusion of this inspection report period), the licensee issued

another Condition Report (CR) M3-97-0850, documenting inadequate corrective

action implementation relative to a licensee event report, LER 3-96-003, involving

unauthorized temporary I-beams over safety-related equipment.

CR M3-97-0850 also documented current licensee findings of heavy, unrestrained

tools and chain falls located in proximity to safety-related equipment. Based upon

the discovery of the unrestrained trolley assembly, as well as the more recent

licensee-identified issues of CR M3-97-0850, the inspector determined that

additional licensee management attention to such " seismic II/l" concerns would be

prudent. The Significant items List (SIL) enclosed with the NRC Restart Assessment

Plan for Millstone Unit 3 documents an item for " Resident Emphasis: Seismic II/l".

This issue, with emphasis upon the new problems documented above, will be

tracked as an inspector followup item (IFl 423/97-01-07) to evaluate both the

timeliness and adequacy of further corrective measures in this area.

Overall, the plant inspection-tours revealed improvements in Unit 3 areas of housekeeping,

material conditions, and work controls. With the exception of the problem with

unrestrained equipment, noted as an IFl above, the licensee provided adequate response to

the inspector questions and field observations and demonstrated continued progress in the

physical enhancements to the plant field conditions.

-

.

'

32

U3.lli Enaineerina

U3 E8

Miscellaneous Engineering issues

E8.1

IOnen) Unresolved item 50-423/96-01-08: Slave Relav/Overlan Test Deficiencies

a.

Inspection Scope (92903)

In 1993 the licensee identified slave relay and other testing deficiencies. As a result of

those findings, the unit director established an overlap testing task force to review the

adequacy of overlap testing for the reactor trip and engineered safeguard systems circuitry.

These reviews were completed in 1993 and the licensee later credited these reviews with

accomplishing the reviews requested in NRC Generic Letter 96-01, " Testing Of Safety-

Related Logic Circuits."

l

in 1996, the licensee performed a review of safety and non-safety related functions,' as '

described in the Final Safety Analysis Report (FSAR) and/or the Safety Evaluation Report

(SER) to determine if the functions were properly tested. The scope of that review

included systems that were considered to be accident mitigating or risk significant as

'

defined in the Maintenance Rule (10 CFR 50.65).

The inspector reviewed selected test procedures, elementary electrical and logic drawings,

Open item Reports (OIRs), ACRs and other documents associated with the testing review

efforts to assess the effectiveness of the reviews.

b.

Observations and Findinas

The licensee task force reviews performed in 1993 identified procedural deficiencies,

circuitry which required design changes, incorrect drawings and FSAR errors. Three LERs

were issued as a result of technical specification violations that occurred due to testing

deficiencies (LERs 3-93-005,-010,-017). In each case the affected circuit performed

satisfactorily when tested. The followup reviews of the FSAR and SER performed in 1996

identified a number of questions regarding testing adequacy and the licensee was

continuing to disposition the associated OIRs. The OIRs that had been dispositioned to

date had not identified logic testing deficiencies of the types discussed in GL 96-01.

The inspector performed an independent review of testing that was performed on reactor

protection and engineered safety features logic circuits to assess the adequacy of the test

procedures. These reviews included the steam generator low level reactor trip and

emergency feedwater pump start testing and portions of the emergency diesel generator

start, load shed and load sequencing testing. The inspector also reviewed several OIRs to

assess the significance of the issues and the adequacy of the licensee resolutions.

The inspector's review of the steam generator low level channel testing included the

following procedures:

SP 3444A01 (Rev. 04) - Steam Generator Water Level Channel Calibration

SP 3443A21 (Rev.10) - Protection Set Cabinet i Operational Test

_ _ . _ _ . _ _ _ _ _ _ . _ _ _ _ . _ _ . _ . _ _ _ _

. . _ _ .

. _ _ _ . . .

i

e

.

4

l

33

.

I

SP 3446B11_(Rev. 09) - Train A Solid State Protection System Operational Test

'

~ The review included testing of the circuitry from the steam generator level transmitters to

i

the output devices, the reactor trip breakers and the auxiliary feedwater pump controls.

The test procedures were thorough and no problems were identified, in addition to the

testing of the automatic reactor trip, the inspector also reviewed the testing of the manual

i

reactor trip push buttons that is performed in'accordance with test SP 3446F331, "SSPS

}

Refueling Tests". The inspector found_that the manual push buttons were properly tested

_

and ensured they would each independently open the reactor trip breakers on either the

shunt trip or undervoltage trip mechanism.

The inspector also reviewed test procedures and selected drawings associated.with the

i

emergency diesel generator starting, load shedding and load sequencing functions. The -

j

following procedures were included in this review:

- SP 3646.A.1 (Rev.12) - Emergency. Diesel Generator A Operability Test- -

')

)'

SP 3646A.5. (Rev. 05) - Offsite Power Transfer Operability Test

-*

i

SP 3646A.8 (Rev.14).- Slave Relay Testing' Train A

j

[

SP 3646A.12 (Rev. 07) - Emergency Diesel Generator A Lockout Test -

-

SP 3646A.15 (Rev.11) - Train A Loss of Power Test

I

1

'

~

SP 3646A'.17 (Rev. 09)- Train A ESF With LOP Test

• .

SP 3646A.19 (Rev. 03) - SIS Transfer of DG From Test to Standby

• -

SP 3646A.21 (Rev. 05) - DG Auto Start on ESF Signal

j

SP 3448E51 (Rev. 01) - Diesel Sequencer Train' A Actuation Timer Test

j

SP 31447MA (Rev. 01)- MP3 Bus 34C Loss of Power Channel Calibration

'

Although this review did not include 100% of the circuitry, the inspector identified the

d

following testing issues:

During the loss of power testing, the EDG receives a start signal from relay 27Y2 in

the bus undervoltage logic circuit. A set of contacts from a different relay in the

]

undervoltage logic feeds a loss of power signal to the emergency generator loading

.

sequencer (EGLS) resulting in an additional EDG start signal. The existing test

]

procedures were not adequate to properly verify the operation of the parallel

signals.

,

The bus undervoltage relays provide trip signals to the emergency bus tie breaker

j

and the feeder breaker from the reserve station' service transformer (RSST). The tie

<

breaker trip circuit contains _three parallel trip circuits and the testing does not verify

j

each path. The RSST breaker trip was not included in any of the surveillance test

j

procedures.

1

1

The EDG starting control has two circuits that are designed such that a start signal

to either circuit will result in the opening of both air start solenoid valves even if one

or more circuit component failures may have occurred. The current testing'does not

independently test the redundancy designed into the circuits.

J

. - _ _ _ . _ . . _ _ . . .

_

. -

_ . _ _,

_ . _ . _ . _ . _ - . _

_ _ _ - . . _ _ _ . . _.

. .

.

34

j

Procedure SP 3646A17 contains notes that state:

-

'

"If possible, the SW pump not tested in lead during SP 3646A.15 should be lead."

t

!

"The CCP pump not tested in SP 3646A.15 should be tested here: the other must.

4

!

be in " PULL-TO-LOCK.""

- "The CHS pump not tested in SP 3646A.15 should be tested here; the other pump

'

must be in " PULL-TO-LOCK.""

,

These notes are intended to ensure that all of the equipment is tested either in SP

3646A15 or SP 3646A17.' These notes are worded as to provide

recommendations, rather than requirements, and as such may not ensure that the

1

standby and swing pumps get tested as required by technical specifications.

' ' "

~ Based on these findings th'e' inspector questioned the adequacy of the previous overlap ' -

+

-testing. reviews that were credited with ensuring that test procedures are adequate to

ensure the technical specification test requirements are met. As a result, the licensee

A:

developed an action plan to address this concern. . The' planned actions included: . _ ..e

A review for overlap test issues of several circuits by comparing electrical

- *

schematic drawings and logic drawings against plant surveillance test pro:edures to

'

ensure that all portions of the logic circuitry, including the parallel logic, interlocks,

bypasses and inhibit circuits, are adequately covered in the surveillance procedures

"

to fulfill the TS requirements. These reviews were to include the loss of power

schemes (undervoltage and degraded voltage), one ESF actuation system, and one

4

reactor trip instrumentation functional unit.

The revision of the surveillance testing for the loss of power initiation logic to

'

adequately verify parallel logic.

The performance of a root cause analysis to determine the cause of missed contacts

to determine if the failure was a generic issue that applied to the overall effort of

- the overlap task force. Based on the findings of the root cause analysis any

additional corrective actions would be determined.

- Subsequent licensee reviews identified additional testing deficiencies. The licensee

performed a self-assessment of the Unit 3 response to Generic Letter 96-01 and concluded

that the response did not meet the requirements of the letter. This was based on the

above specific findings where actuation' contacts were not tested, a lack of auditable

documentation and a difference in philosophy between the 1993 overlap task force and the

requirements of the Generic Letter. This issue was documented in ACR M3-97-0529 and

additional corrective actions were being developed by the licensee at the end of this

inspection period.

,

k-,-

- - + ,

c-

,s.w-

-

,-,e

- .- ,

, - ~

-

y-

c

y

- . _ . . . _ _ . - _

_ _ _ . _ _

,

!

..

35

c.

Conclusion

,

":

- ' ~ The inspector found that the licensee had expended significant resources in the past to~

i

review the logic testing and had identified and corrected numerous deficiencies. The test'

procedures reviewed were generally thorough. However, as discussed above, some test

-

deficiencies continued to exist. The failure to ensure all contacts are operable could result

,

in significant undetected problems. For example, if the emergency bus tie and feeder

1

breakers failed to trip on an undervoltage signal, the EDG output circuit breaker would be

prevented from closing to' energize the bus to power necessary safety equipment. The

inspector also noted that a more thorough self-assessment would have been appropriate

,

prior to the licensee submission of a response to GL 96-01.

]

.

.

This item remains open pending NRC review of additional licensee corrective actions, and

)

the assessment'of the significance of any additional findings and the results of additional

tes:ing that is performed.

i

.

E8.2 (Closed - Part of SIL ltem 67) ACR M3-96-0621: Potential For Overloadina Station

Blackout (SBO) Diesel

a.

Insoection Scope (92903)

This ACR identified a concern that the station blackout (SBO) diesel generator could be

overloaded if a safety injection or containment depressurization signal occurred while the

-

SBO diesel was supplying power to an emergency bus.

-

~

b.

Observations and Findinas

-

. In the event of a loss of all ac power,-the SBO dieselis manually started in accordance .a

Y

with procedure ECA O.0, " Loss of All'AC Power." Prior to energizing the bus from the

SBO diesel, the procedure directs the operators to place the control switches for large

)

loads in the pull-to-lock position. This action blocks the automatic start of the loads.

Following the energization of the bus, the operator then manually starts loads needed to '-

cope with the station blackout condition,

i

c.

Conclusions

- The inspector reviewed the associated ACR, procedure and elementary electrical drawings

)

and concluded that the licensee had appropriately reviewed and dispositioned this ACR.

This item is closed, (representing partial closure of SIL ltem 67).

o

.

l

'

36

IV Plant Support

Millstone Units 1,2, and 3

R1

Radiological Protection and Chemistry Controls

a.

Inspection Scope (83750)

i

The inspector reviewed the licensee's radiation protection programs established at each

unit and for the site. A review of specific work performed, the programs for maintaining

occupational radiation exposures as low as is reasonably achievable, and tours of the

various radiologically controlled areas (RCAs) were conducted by the inspector.

b.

Observations and Findinas

Unit 1

i

i

The inspector toured portions of the reactor and liquid radwaste buildings as part of the

j

inspection at Unit 1. The inspector noted a significant decrease in the number and size of

posted contaminated areas in the unit, which was described to the inspector as part of.the

unit's clean-up policy. The inspector noted that while some work was being performed in

the reactor building at the time of this inspection, significant radiological work, especially in

the drywell and on the refueling floor had yet to commence. The inspector also toured the

Xenon / krypton building, which houses some of the off-gas treatment systems and delay

tanks. This structure and its component systems had undergone a significant

refurbishment during 1996. Two areas within the structure were appropriately posted and

controlled as high radiation areas. The upper level of the structure housed two glycol

chiller systems, one of which was still under refurbishment.

For 1997, the licensee had established a goal of not more than 399 person-rem. This goal

'

is based on completing significant work and having the unit ready for restart of operations

during 1997. As described in a previous NRC Inspection Report (50-245/96-09), the unit

has significantly increased the number of personnel assigned to perform work planning and

ALARA functions. Seven persons within the Radiation Protection Department are now

assigned to ALARA, and two technicians are on loan as work week managers. Each of the

ALARA personnel have been assigned specific work packages and/or work areas for

planning purposes, and are responsibla for coordination with engineering and the work

groups to ensure proper ALARA controls are incorporated into the work packages.

Additionally, an ALARA Committee has been established, which includes all department

managers.

On January 15,1997, the licensec identified, through its Adverse Condition Reporting

(ACR) process (ACR # M1-97-0094) that fan HVE-14, which exhausts portions of the

Radwaste Storage Building, was potentially an unmonitored release pathway, as the fan

was not connected to the main plant stack, and no radiological effluent monitoring

equipment was located with this fan. The inspector discussed this ACR with a licensee

representative during this inspection, and reviewed a reportability evaluation performed bt

the site Engineering Department which analyzed the significance of this ACR. The

inspector determined that the reportability determination was invalid in that the evaluation

~

1

4

37

erroneously addressed a building that was different from the Radwaste Storage Building.

Subsequent to this finding, the licensee wrote another ACR (ACR # M1-97-0282) to

document this error, and subsequently determined that a notification to the NRC was

required, which was made on February 6,1997. Failure to monitor effluents released to

the environment from the Radwaste Storage Building to demonstrate compliance with

applicable regulatory limits, including 10 CFR 20.1301, is a violation of 10 CFR 20.1302.

(VIO 245/97-01-08)

Unit 2

The inspector toured various portions of the Unit 2 RCA, including the Auxiliary and

Containment Buildings, as part of this inspection. In general, all areas were determined to

be in compliance with NRC requirements for radiological postings and control of radioactive

material. On February 5,1997, the inspector observed the removal and subsequent

transfer of two highly radioactive pieces of debris previously found in the reactor vessel. A

metal nut and a tie wrap, each reading in excess of 10 Roentgens per hour on contact,

were removed from a storage bucket that was being kept in the refueling cavity,

transferred to a lead pig, and moved from the Containment Building, through the Auxiliary

Building and outside the unit to a designated storage area. The inspector observed the pre-

job briefing, which included a discussion of engineering controls for the minimization of

personnel exposure, and the conduct of all work until the shield pig with the objects was

removed from the unit. This activity also involved significant coordination between the

unit operations department, the health physics department, the site security organization

"

and the self-directed work group. Due to the careful planning process used, total exposure

for this work was less than 25 millirem.

For 1997, the unit established a goal of not more than 182 person-rem. Since the last

specialist inspection (50-336/96-09), the unit had flooded up the refueling cavity and

successfully off-loaded the reactor fuel to the spent fuel pool. Significant strides in

improving the unit ALARA program have also been made. ALARA coordinators have been

identified in each of the major work departments within the unit, and the Unit ALARA

coordinator was in the process of developing a training program for them. A unit ALARA

program procedure was also drafted, however, it was not issued at the time of this

inspection. Discussions with the ALARA coordinator, Health Physics Manager and Unit

Directr,r indicated the intent to establish an ALARA Committee, to include the major

depa:tment heads and their ALARA coordinators.

The inspector interviewed the health physics manager, and reviewed the documentation

associated with three ACRs (M2-97-0086, M2-97-0091 and M2-97-0142) written to

identify improper entries to the RCA which occurred during a ten day period in January

1997. In two of the instances, workers entered the RCA without having signed in on a

RWP, and without having on an electronic docimeter, as directed by the unit radiation

protection staff. In the third instance, a fire watch entered the unit with an electronic

dosimeter that had not been properly turned on. When discovered through self-checking,

the fire watch remained in the RCA with the non-functioning electronic dosimeter unti!

completion of the fire watch round. Procedure RPM 5.22 requires radiation workers to

comply with written instructions, including RWPs, from the radiation protection staff.

Although the safety significance of each of these events is individ ally low, as each worker

'

.

1

i

e

i

38

was wearing a thermoluminescent dosimeter (TLD) which is utilized to determine dose of

record, the number of events in such a short duration are of concern. Additionally, the fire

watch event may highlight a problem with the training given and the perception of the

workers performing this task relative to other plant requirements, such as radiological

j

safety. Both of these issues were discussed by the inspector with the unit Health Physics

Manager and Unit Director, and with the station Vice President - Work Services. Short-

)

term corrective actions taken by the licensee included posting a health physics technician

at the main RCA entrance to ensure that personnel entering the RCA were wearing a

functional electronic dosimeter. Long-term corrective actions were not identified, however,

at the time of this inspection. Failure to adhere to the licensee's radiation protection

program, specifically procedure RPM 5.22,is a violation of 10 CFR 20.1101. (VIO

336/97-01-09)

Unit 3

The inspector toured various portions of the Unit 3 RCA, including the containment and

auxiliary buildings. Since the last specialist inspection (50-423/96-07) a significant

reduction in the number of leaking valves was observed, based on the reduced number of

catch containments observed. The unit continues te have very low dose rates in most

i

,,

areas, and significant portions of the Containment rematri accessible as clean areas.

The inspector reviewed the licensee's ALARA program, including the 1997 ALARA goal of

not more than 170 person-rem. The unit focus in ALARA has been to improve the work

order process, to include having RWP and ALARA control information included in the work

-

orders, in addition, the licensee has begun to include detailed maps and pictures of areas

and systems to be worked in the work order package. This is the result of a significant

campaign completed in 1996 by the unit radiation protection staff to photograph over

30,000 components in the RCA.

Site Health Physics

The site health physics group is responsible for providing calibration and dosimetry

services, health physics engineering and health physics support to the units and to other

site-wide organizations. As part of this inspection, a review of certain activities was

i

coni:cted by the inspector.

l

The self-directed work group includes six health physics technicians whose primary focus

is to support the activities of the Waste Services Department. As previously noted in the

discussion on Unit 2, above, this group was involved in the transfer of two highly activated

pieces of material from the Unit 2 vessel to a storage location outside the unit. In addition,

the inspector also reviewed the performance of this group during a recently completed

waste transfer evolution.

As part of the Liquid Radwaste Remediation Project at Unit 1, waste concentrates spilled

inside the "A" Concentrator cubicle were removed in 1996. In addition to being

i

radioactive, this material also contained asbestos, and thus required specialized engineering

controls for handling, as required by the Occupational Safety and Health Administration

(OSHA). On January 24,1997, ten barrels of this material, each containing several

7

V

e

39

plastic-wrapped bags of the waste, were transferred to a processing liner, which ultimately

was to be solidified and buried as low-level radioactive waste. Due to OSHA requirements,

this transfer was conducted inside a tent-like structure erected around the liner by a team

'

of five specially trained contractors. Because of the OSHA requirements, the health

physics technicians from the self-directed work group could not enter the tent once work

began. Based upon interviews conducted by the inspector with all members of the

j

contractor work team, the contractor industrial hygienist, and the self directed work group

technicians, and a review of documentation associated with this activity (RWP, ALARA

review, pre-job briefing package, post-job review) the inspector determined that the work

i

was appropriately controlled in accordance with NRC regulations. While all five contractors

were contaminated on their person and/or clothing upon completion of the work, this was

'

not the result of a breakdown of radiological controls. None of the contaminations resulted

,

in a significant radiological exposure.

'

c.

Conclusions

Unit 1

Noticeable reductions in the amount of contaminated spaces within the unit were

observed. ALARA planning and staffing of the ALARA group have significantly improved,

however, the effectiveness of this cannot be determined until more radiologically

significant work commences. One violation of NRC requirement involving an unmonitored

release path was identified.

Unit 2

Good work planning and control was observed for the transfer of highly irradiated materials

j

from the reactor vessel. Significant changes in the planning and control of radiological ,

work is under development. A violation of NRC requirements involving poor radiological

worker practices was identified. While short-term corrective actions were implemented by

the licensee at the time of this inspection, long-term actions had not yet been identified.

Unit 3

Contamination control improvements, especially the reduction in the number and need for

catch containments, was observed during tours of the RCA. Incorporation of RWP and

ALARA information into the work orders was an improvement, although the effectiveness

of this will have to be evaluated once significant radiological work resumes.

Site Health Physics

Appropriate support to Unit 2 was observed during the transfer of highly irradiated

material. Appropriate work controls were implemented during the transfer of asbestos

contaminated concentrates wastes.

$

40

R8

Miscellaneous Radiological Protection and Chemistry Issues

A recent discovery of a licensee operating their facility in a manner contrary to the Updated

Final Safety Analysis Report (UFSAR) description highlighted the need for a special focused

review that compares plant practices, procedures and/or parameters to the UFSAR

descriptions.

While performing the inspections discussed in this report, the inspector reviewed the

applicable portions of the UFSAR that related to the areas inspected. The inspector

verified that the UFSAR wording was consistent with the observed plant practices,

procedures and/or parameters, except in the area of the management organization and

responsibilities for radiation protection. Section 12.5 of the Unit 1 UFSAR and Section

11.2.3 of the Unit 2 UFSAR make reference to Section 12.5.1 of the Unit 3 UFSAR for a

full description of the health physics organization and reporting functions. This description

no longer is accurate due to the restructuring and unitization of the Radiation Protection

Program. The Work Services organization recognized the need to update the Unit 3 UFSAR

to reflect the management changes and identified it to the Site Licensing Director by

memorandum, dated November 29,1996.

V. Manaaement Meetinas

X1

Exit Meeting Summary

The inspectors presented the inspection results to members of licensee management at the

conclusion of the inspectian. The licensee acknowledged the findings presented.

X1.2 Final Safety Analysis Report Review

A recent discovery of a licensee operating their facility in a manner contrary to the final

safety analysis report (FSAR) description highlighted the need for additional verification

that licensees were complying with FSAR commitments. All reactor inspections will

provide additional attention to FSAR commitments and their incorporation into plant

practices, procedures and parameters.

While performing the inspections which are discussed in this report the inspectors

reviewed the applicable portions of the FSAR that related to the areas inspected.

inconsistencies were noted between the wording of the FSAR and the plant practices,

procedures and/or parameters observed by the inspectors, as documented in Sections

U3.E8.1 and R8.

<

, _ _-__

_

__

. . -

_.

.-.~. _ - --

._

..

_ _ _ .

- - _ _ . _

b

f

41

INSPECTION PROCEDURES USED

~

IP 37550:

Engineering

IP 37551:

Onsite Engineering

,

1

IP 40500:

Licensee Self-Assessments Related to Safety issues inspections

]

IP 61726:

Surveillance Observations

i

l

IP 62707:

Maintenance Observations

1'

l

IP 71707:

Plant Operations

IP 71750:

Plant Support Activities

j

IP 83750:

Occupational Radiation Exposure

'

IP 92700:

Onsite follow-up of Written reports of Nonroutine Events at Power Reactor .,

Facilities

,

i

IP 92901:

Follow-up Operations

IP 92902:

Follow-up Maintenance

IP 92903:

Follow-up Engineering

.

.

,-

..

..

~. -

-.

--

.

,.

-

v

i

i

1

42

ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

URI 50-245/97 01-01

U 1.02.1

Spent Fuel Pool Cleanliness

I

URI 50-245/97-01-02

U 1.03.1

Operations Procedure Adequacy

1

URI 50 245,336,423/

U 1.05.1

Inaccurate Personal Qualification Statements

97-01-03

URI 50-245/97-01-04

U 1.E1.1

Resolution of A-46 Program Outliers

URI 50-245/97-01-05 -

U1.E2.2

Low Flow Operation of Containment isolation Check

i

Valve 1-CU-29

l

NCV

U2.08.2

Failure to Enter TS Action for inoperable Nis

IFl 50-423/97-01-06

U3.01.1

Interpretation of TS Language and LCO Actions

IFl 50-423/97-01-07

U3.M 8.1

Seismic 11/1 Concerns

VIO 50-245/97-01-08

R1

Failure to Monitor Gaseous Effluents from the

i

Radwaste Storage Building

VIO 50-336/97-01-09

R1

Entering RCA w/o Electronic Dosimeter or Signing

RWP

Closed

LER 50-245/96-03

U 1.E1.1

LER 50-336/96-15

U2.02.2

VIO 50-336/94-17-10

U2.08.1

Operation Outside Systtim Design Parameters

'

URI 50-336/96-04-09

U 2.M8.1

Troubleshooting Controls

URI 50-336/96-06-06

U2.M8.2 High Pressure Safety injection Check Valve Backflow

Testing

j

URI 50 336/95-07-06

U2.E8.1,

Condensate Storage Tank Siphon Break

-

-

URI 50-336/95-11-03

U2.E8.2

10 CFR 21 Reportability Review

Discussed

VIO 50-245/95-42-01

U 1.08.1

Failure to Prevent Work Which had the Potential for

Draining the Reactor Vessel

URI 50-336/96-01-04

U2.02.2

Loss of DC Bus Event

URI 50-423/96-01-08

U3.E8.1

Slave Relay / Overlap Test Deficiencies

Sianificant items List

i

Unit 3 SIL #67

Partial Closure

,

,

.- . - . - .

,

'

i

43

LIST OF ACRONYMS USED

ACR(s)

adverse condition report (s)

AISC

American Institute of Steel Construction

ALARA

as low as reasonably achievable

ANSI /ANS

American National Standards Institute /American Nuclear

ASME

American Society of Mechanical Engineers

CCP

reactor plant component cooling

CES/NTE

component engineering services / nondestructive test engineering

CFR

Code of Federal Regulations

CHS

charging system

CR(s)

condition report (s)

DCN

design change notice

DG

diesel generator

EDG

emergency diesel generator

EGLS

emergency generator loading sequence

ESAS

engineered safeguards actuation system

i

ESF

engineered safety feature

FIN

Fix-It-Now

FLS

first line supervisor

FME

foreign material exclusion

GIP (s)

generic implementation procedure (s)

GL

Generic Letter

gpm

gallons per minute

HELB

high energy line break

HPSI

high pressure safety injection

ICAVP

Independent Corrective Action Verification Program

IFl

inspector follow item

IR(s)

inspection Reports (s)

LCO

limiting condition for operation

LER(s)

licensee event report (s)

M&TE

material & test equipment

NDT

nil ductility transition

NRC

Nuclear Regulatory Commission

NRR

Nuchar Reactor Regulation

NSIC

Nut.,ar Safety Information Center

NS&O

nuclear safety and oversight

NUREG

Nuclear Regulation

.OCA

Office of. Congressional Affairs

OIR(s)

open item report (s)

OJT

on the job training

. OSHA

Occupational Safety & Health Administration

PAO

Public Affairs Office

PDCR

plant design change record

'

PDR

Public Document Room

PiR(s)

plant information report (s)

PUP

procedure upgrade program

QA

quality assurance

.

._

_

_

..

. . .

~ . .

-

. - _ . .

..

.

. . - .

_.

d

. o

e

'

44

,

i

QAS

- Quality and Assessment Services

OPTR

quadrant power ti!! ratio

RBCCW

reactor building closed cooling water

RCA

radiologically controlled area

RFO

refueling outage

RHS

residual heat removal

RI

Region I

i

^

RPS

reaction protection system

RSST

reserve station service transformer

RWCU

reactor water cleanup

RWP(s)

radiation work permit (s)

SBO

station blackout

SER(s)

safety evaluation report (s)

'

SPO

Special Projects Office

,

SQUG

seismic qualification utility group

"-

SSER

supplementel safety evaluation report

SSR

reactor plant samples

i

TBSCCW

turbine building secondary closed cooling water

'

TLD(s)

thermoluminescent dosimeter (s)

i

TR(s)

trouble report (s)

TS(s)

technical specification (s)

UFSAR

updated final safety analysis report

USl

unresolved safety issue

VIO

violation

,

l

WC

work control

I

i

4

r

{

.j .

1

I

e

d

l

3

5

a

m

1

-

.~

. _ . .

, . , . . - -

I