ML20134H147

Insp Rept 50-341/96-17 on 961004-1206.Violations Noted. Major Areas Inspected:Several Events Which Occurred During Refueling Outage,Plant Personnel & Evaluated Event Logs & Data

ML20134H147
Person / Time
Site:	Fermi
Issue date:	02/03/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20134H131	List: IR 05000341/1996017 ML20134H128 ML20134H147
References
50-341-96-17, NUDOCS 9702110176
Download: ML20134H147 (22)

See also: IR 05000341/1996017

Text

. .

U.S. NUCLEAR REGULATORY COMISSION

REGION 3

Docket No: 50-341

License No: NPF-43 .

Report No: 50-341/96017

!

Licensee: Detroit Edison Company (Deco)

i

Facility: Enrico Fermi, Unit 2 L

Location: 6400 N. Dixie Hwy.

Newport, MI 48166 i

Dates: October 4, 1996 through December 6, 1996

Inspectors: A. Vegel, Senior Resident Inspector

C. O'Keefe, Resident Inspector

A. Kugler, Fermi 2 Project Manager, NRR

Approved by: Michael J. Jordan, Chief, Branch 5

Division of Reactor Projects

l

9702110176 970203

PDR ADOCK 05000341

G PDR

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

. >

.

I

'

l

EXECUTIVE SUMARY

,

Enrico Fermi, Unit 2

l

NRC Inspection Report 50-341/96017

,

i

e On October 4,1996, cross-tie valves to the Residual- Heat Removal (RHR)

l reservoirs were rendered inoperable. Operators failed to recognize that

the condition exceeded a Technical Specification (TS) Limiting Condition

'

of Operation (LCO). Once the condition was recognized, and actions to

cross-tie the reservoirs were taken, operators failed to evaluate plant

indications. Operators inappropriately detemined that the reservoirs

were cross-tied when one valve had failed in the closed position and

level indications reasonably demonstrated that the reservoirs were  :

isolated. This rendered the Ultimate Heat Sink (UHS) unavailable, and i

various safety systems inoperable. The plant was in a condition

prohibited by TS for greater than 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br />. This is an apparent i

violation of several cascaded TS.  ;

e On October 13, 1996, the Onsite Review Safety Organization (ORS 0)

inappropriately approved a Technical Specification Clarification (TSC)

in an attempt to operate the plant in a condition that was prohibited by

TS rather than requesting a Notice of Enforcement Discretion (NOED) or

an amendment to the TS. This is an apparent violation of TS.

e On November 4, 1996, the plant re-entered operational Mode 5 without

performing' TS required surveillance testing of the Control Rod Block

,

Instrumentation. This is an apparent violation of TS.

These apparent violations were due to several significant root causes:

e One of the RHR reservoir cross-tie valves' (Valve F601A) disk separated

from the valve operator. A set screw on the spline was not tuck welded,

as required, to prevent the screw becoming loose and the disc from

disconnecting from the spline.

e Established periodic testing of the knx seservoir cross-tie valves would

not have detected the valve malfunction.

e Operators and work planners failed to recognize the effect de-energizing

bus 72ED had on the UHS. The planners and approving organizations of

the maintenance activity did not recognize that TS LCO had been entered.

e Operators performed an operability evolution of the UHS using non-

seismic instrumentation in lieu of valid safety-related and seismic

instrumentation that they believed was malfunctioning.

e Licensee made a TS interpretation to allow disregarding a valid TS

requirement. This was due to insufficient knowledge of the regulatory

requirements.

I

l

.

l

2

! 1

2 _ _ -_ _ - . . _ l

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

. .

\

l e Poor communications between maintenance and operations personnel were a

major contributor to missing a TS surveillance. This was compounded by

insufficient knowledge of technical specifications and inadequate

control of work activities on the refueling floor.

<

e Inadequate procedure in that all RPV bolting activities were not l

completed prior to declaring a change to Mode 4. i

l

I

1 \

i l

< l

l

!

.

,

l

,

I

3

. .

1

!

Report Details l

The inspectors reviewed several events that occurred during the refueling

outage. The inspectors independently interviewed plant personnel and

evaluated event logs and data.

I. Doerations

l

01 Conduct of Operations

01.1 General Comments (71707)  ;

i

On three occasions during the refueling outage, between October 4 and  :

November 4,1996, technical specification requirements were not met. Two

of the three events are discussed below in the OPERATIONS area while the j

third is discussed in the MAINTENANCE area. , ;

!

01.2 Residual Heat Removal (RHR) Reservoir Cross-Tie Lines Were Not Opened

Der Technical Soecification (TS) Reauirements

a. Insoection Scone (93702)

The inspectors independently reviewed the various documentation

associated with the October 4,1996 loss of cross-tie capability to the  !

RHR reservoirs. The inspectors also interviewed the appropriate i

operations personnel and management.  !

b. Observations and Findinas

On October 4, Operations deenergized bus 72ED in preparation for

maintenance. This action removed power to motor operated valves (MOVs)

Ell 50-F602A and F6028. These MOVs are in one of two cross-tie lines for

the Residual Heat Removal (RHR) reservoirs. Technical Specification (TS)

3.7.1.5, Action C, requires that with one cross-tie line for the

Ultimate Heat Sink (UHS) RHR reservoir inoperable, the valves in the

other cross-tie line shall be opened and deenergized within eight hours.

About Eight hours and nine minutes after one cross-tie line was rendered

inoperable, licensed operators realized that this action statement had

not been completed, so the UHS was declared inoperable. The operating

division of shutdown cooling was declared inoperable as a result, which

was reported to the NRC Operations Center per 50.72(b)(2)(iii)(b). The

cross-tie valves in the other division (Ell 50-F601A and F6018) were

promptly opened and deenergized as required. This event was documented

in DER 96-1288.

Shortly after upening valves F601A and F601B to comply with Action C,

control room operators identified that level indications between the two

reservoirs did not agree, as should be expected with open cross-tie

lines. A non-licensed operator was dispatched to compare local (non-

seismic and non-safety related) indications. Each reservoir had two

local level indicators, and the operator determined that three of the

four agreed, with one of the detectors on the Division I reservoir

reading higher. The operation shift was satisfied that the pools were

successfully cross-tied. Operations considered the UHS to be operable.

4

_ . _ _ _ _ _ _ _ _ __. . _ _ _ _ _ _ _ _ _ _ __

7

. .

'

,

'

About 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> after the original de-energizing of bus 72ED, water was '

3

'

added to the Division I reservoir with no noticeable change in the

Division 2 reservoir water level. The UHS level divergence was

} investigated, and operators determined that the pools were not cross-

tied. Divers were sent into the UHS and determined that the F601A valve

was actually shut while it indicated open. The motor operated actuator

did not cause valve movement. Operators then opened the other cross-tie

valves to comply with TS 3.7.1.5, Action C. Subsequent investigation by ,

the licensee identified a loose set screw in the bull gear on the F601A

actuator. This event was documented in DER 96-1289.

!

c. Conclusions i

i

The inspectors determined that on October 4,1996, cross-tie valves to i

the Residual Heat Removal (RHR) reservoirs were rendered inoperable.

Operators failed to recognize that the condition exceeded a Technical j

Specification (TS) Limiting Condition of Operation (LCO). Once the  !

condition was recognized, and actions to cross-tie the reservoirs were

taken, operators failed to evaluate plant indications. Operators

inappropriately determined that the reservoirs were cross-tied when one

valve had failed in the closed position and level indications reasonably

demonstrated that the reservoirs were isolated. This rendered the

Ultimate Heat Sink (UHS) and various safety systems inoperable.

The inspectors concluded that the plant was in a condition prohibited by

TS for greater than 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br />. This is an apparent violation of several

cascaded TS.

TS 3.7.1.5 requires the Ultimate Heat Sink, comprised of two one-half

capacity residual heat removal (RHR) reservoirs with the capability of

being cross-connected, shall be OPERABLE with...(g) two reservoir cross-

connect lines, each with two OPERABLE motor operated cross-connect

valves.

e Action (c) of TS 3.7.1.5 requires with one or more reservoir

cross-connect valves inoperable, within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> open and de-

energize both valves in at least one cross-connect line and verify

that these valves remain open and de-energized at least once per 7

days. Otherwise, declare both reservoirs inoperable and take the

ACTION of e. below.

e Action (e.2) of TS 3.7.1.5 requires that in OPERATIONAL CONDITIONS

4 or 5, declare RHRSW system, the EESW system and the diesel

9enerator cooling water systems inoperable and take ACTION

required by Specifications 3.7.1.1, 3.7.1.3 and 3.7.1.4.

Cascaded TS 3.7.1.1, ACTION (c) requires that in OPERATIONAL CONDITION 5

with the RHRSW subsystem (s), which is associated with an RHR loop

required by Specification 3.9.11.1 inoperable, declare the associated

RHR system inoperable and take ACTION required by Specification

3.9.11.1.

e TS 3.9.11.1, ACTION requires with no RHR shutdown cooling mode ,

loop OPERABLE, within I hour and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> j

l

5

_

. .

j

l

!

!

'

thereafter, verify the OPERABILITY of at least one alternate l

method capable of decay heat removal. Otherwise, suspend all l

operations involving an increase in the reactor decay heat load l

and establish SECONDARY CONTAllWENT INTEGRITY within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.  ;

i Cascaded TS 3.7.1.3, ACTION requires that with one EESW system -

l subsystem inoperable, declare the associated EECW system subsystem

inoperable and take the ACTION required by Specification 3.7.1.2.  ;

e TS 3.7.I.2, ACTION (b) requires in OPERATIONAL CONDITION 4 or 5, ,

determine the OPERABILITY of the safety-related equipment l

associated with an inoperable EECW system subsyst.en and take the i

ACTIONS required by the applicable Specifications.

Cascaded TS 3.7.1.4, ACTION requires with one or more diesel generator ,

cooling water subsystems inoperable, declare the associated diesel  ;

generator inoperable and take the ACTION required by Specification l

3.8.1.2.

e TS 3.8.1.2, ACTION (b) requires that with less than the above

required A.C. electrical power sources (One onsite A.C. electrical

power sourco, Division I or Division II, consisting of two

emergency diesel generators] OPERABLE, suspend CORE ALTERATIONS, '

handling of irradiated fuel in the secondary containment,

operations with a potential for draining the reactor vessel and ,

crane operations over the spent fuel storage pool when fuel  :

assemblies are stored within. l

!

01.3 Failure to Meet TS Reauirements for Control Rods  ;

a. Insnection Scone (93702)

The inspectors reviewed documentation associated with a October 13,

1996, On-site Review Safety Organization (ORS 0) approved Technical

Specification Clarification (TSC)96-003. The inspectors also

interviewed various operations personnel and management.

b. Observations and Findinas

During the refueling outage, with several control rods withdrawn in

defueled cells to permit reactor vessel inspections by camera, a problem

was encountered with the refueling bridge. The withdrawn control rods

had blade guides removed to permit room for inspection cameras, and thus

could not be reinserted for lack of support. Reinsta11ation of blade

guides would have required the use of the refueling bridge. However,

the refueling bridge power supply cable shorted and was repaired during

the camera inspections. When the problem with the bridge was repaired,

the refueling' bridge interlock surveillance was required to be performed

before the bridge could be declared operable. This required briefly

placing the mode switch in Startup to verify interlocks functioned.

However, footnotes in Technical Specification Table 1.2 and Technical

Specification Surveillance Requirement 4.9.1.1 to Technical

i Specification 3.9.1, required that all control rods be fully inserted

{

!

! 6

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

-

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

_.

. .

,

.I

b

i

i

prior to placing the mode switch in a position other than Shutdown or l

Refuel for surveillance performance.  ;

i ,

,

The inspectors determined that the footnotes were slightly different. i

l The footnote to Table 1.2 stated "the reactor mode switch may be placed i

in Run, startup/ Hot Standby, or Refuel position to test the switch i

interlock functions and related instrumentation provided that the  ;

i control rods are verified to REMAIN FULLY INSERTED [ capitals added for '

i

emphasis] by a second licensed operator or other technically qualified  !

' member of the unit technical staff." The footnote to the technical ,

specification shrveillance requirement stated "the reactor mode switch  ;

may be placed in the Run or Startup/ Hot Standby position to test 1

interlock functions provided that ALL [ capitals added to emphasize the )

i

difference in the footnotes) control rods are verified to REMAIN FULLY l

j INSERTED [ capitals added for em>hasis) by a second licensed operator or l

other technically qualified mem>er of the unit technical staff." The '

,

clarification was written to interpret that "all control rods" of

i Specification 3.9.1 and that the term control rods applies only to " core

.

cells containing fuel and does not include rods withdraw or removed in

J

accordance with 3.9.10.2." This clarification is in agreement with

Improved - Standard Technical Specifications; however, improved

l specifications are not approved for Fermi.

J

i On October 13, ORSO approved Technical Specification Clarification (TSC)

!96-003. Based on the interpretation contained in TSC 96-003, the

, refueling bridge interlock surveillance was performed on October 13,

i resulting in the Mode Switch being unlocked and placed in Run and

Startup with some control rods withdrawn. Fermi did not request an

amendment to their existing technical specifications or a waiver of the

, current requirements.

1

i TS 3.9.10.2 requires that any number of control rods and/or control rod

i drive mechanisms may be removed from the core and/or reactor pressure

vessel provided that at least the following requirements are satisfied

i

until all control rods and control rod drive mechanisms are reinstalled

, and all control rods are inserted in the core....(a.) the reactor mode

, switch is OPERABLE and locked in the Shutdown position or in the Refuel

position per Specification 3.9.1, except that the Refuel position "one-

rod-out" interlock may be bypassed, as required, for those control rods

and/or control rod drive mechanisms to be removed, after the fuel

assemblies have been removed as specified in TS 3.9.10.2 (b through e).

l c. [gnelusions

! The inspectors determined that operators had entered Technical

l Specification 3.9.10.2, to allow withdrawing the control rods for the

inspections. This TS required that the mode switch remain locked in

i Refuel or Shutdown until all control rods were fully inserted. This was

in conflicted with the licensee's use of TSC 96-003. At the Residents'

request, NRR Technical Specification Branch reviewed this icsue, and

determined that Fermi should have complied with TS 3.9.10.2. The

j appropriate action should have been to request a N0ED or amend their

4 cnrrent technical specifications. This is an apparent violation of TS

l 3.9.10.2.

.

7

-

- . .

.- - -

1

I

02.0 System Description

02.1 Description of the RHR Reservoir

The Fermi Ultimate Heat Sink (VHS) is comprised of two 50 percent

Residual Heat Removal (RHR) reservoirs, which can be interconnected

through either of two cross-tie lines, each containing two ball valves. l

The sink is divided to minimize the impact of a below grade breach of

the reservoir but can be cross-connected to ensure the accident analysis

volume of UHS water is available. The technical specification does not

prohibit operation of the reservoirs either cross-connected or not

cross-connected. Each line has a normally open valve with the other

valve in the line normally closed.

03.0 Seouence of Events

The following sequence of events were determined by the inspectors from

reviews of various parameter chart recording and process computer alarm

recordings.

03.1 RHR Reservoir

11:11 am October 4 Bus 72ED was de-energize rendering RHR reservoir

cross-tie valves Ell 50-F602A and F602B

inoperable. (This prevented cross-connection

through the affected line) (The other cross- ,

connect line has valve F601A closed and F601B l

open)

Operating crew does not recognize that they were

in an 8-hour LCO per T.S. 3.7.1.5, Action C.

7:20 pm October 4 Operating crew recognize that they were in T.S.

LCO. They declare the operating division of

shutdown cooling inoperable. Valve F601A was

directed to be opened (F6018 was already open).

Actions for TS 3.7.1.1, 3.7.1.3, and 3.7.1.4

were also entered. Operations verified that no

core alterations or activities with the

potential to drain the vessel were in progress

or scheduled. TS 3.9.11.1 was the most

limiting. Also, the following systems were

affected; secondary containment, star.dby gas

treatmer.t, control center HVAC, D.C. power

sources, A.C. power sources, and various

electrical power components and systems.

7:41 pm October 4 Valve F601A indicates open in the control room

and operators believe that requirements of TS 3.7.1.5 and the associated cascading TS action

requirements were met. The LC0 Actions were

exited.

8

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

. .

.

h

Shortly after opening ,

valve F601A: Operators notice difference in RHR reservoir  !

level on control room safety related

indications. A operator was dispatched to

investigate. Local, non-seismic level

indications have 3 of 4 in agreement with each

other. The fourth indicator was out of

calibration since 1993 and could not be ,

calibrated during several attempts since 1993. j

.

Operations crew determine that the cross-tie

l

line was open and that the UHS was operable.  ;

11:14 pm October 4 With six minutes romaining of a required four-

hour notification, NRC was notified via ENS

(Event # 31100) of inoperable shutdown cooling.

This notification was subsequently retracted on

October 5 because the licensee deter. wined that

in addition to the loss of a cress-connect line,

a division of electrical power would also be

needed. Therefore, this was beyond the "alone"

stipulation of the 10 CFR 50.72 criteria. (This

was considered to be valid if valve F601A was

OPEN).

l

1:49 pm October 5 Water was added to the Division 1 RHR reservoir,

operators noticed that the level in Division 2

did not change. (Observation of control room

indications). A diver was requested to inspect

the cross-tie line valves.

3:49 pm October 5 Valve F602B (one of the two originally affected

when bus 72ED was de-energized) was manually

opened and valve F601A was closed for inspection

of the reservoir by the diver.

With completion of this action, unbeknownst to  !

the operators, the UHS was returned to operable I

status. l

'

5:00 pm October 5 Operators observed that the indications for the

(about) two reservoirs were equalizing. Division I

reservoir was increasing and the Division II was

,

decreasing to an equalization level. Operations

!

determined that the cross-connection through

l valves F601A/B (established at 7:41 pm on

l October 4) was not open. It was determined that

j TS 3.7.1.5 and cascading TS 3.9.11.1 was not met

(since TS 3.7.1.1, 3.7.1. 3, and 3.7.1.4 were

,

exited on October 4, they were also not met and

'

'

not recognized by the licensee). The plant was

determined to be in a condition prohibited by

TS.

9

- . _

- - - . _-.

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

. .

I

. 1:59 pm October 6 NRC was updated via ENS. The original 11:14 pm

! on October 4 notification was updated and in  ;

effect nullified the retraction. Update does not i

clearly state that T.S. 3.7.1.5 was not met for

the entire period.

LATER Diver determines that valve F601A did not open ]

when operated from the control room (valve '

,

position lights indicated open).

October 21 The failure of valve F601A was determined to be

.

a loose set screw on the valve operator spline

bushing.

03.2 Failure tt, Meet TS Reauirements for Control Rods

Sept 27 Plant was shutdown for fifth refueling outage

October ?! First fuel shuffle completed. At stopping

4

point, a number of peripheral cells were

defueled, the control rods withdrawn, and the

blade guides removed. In-vessel camera

inspections were begun.

9:50 pm October 11 Refueling bridge blew a main line fuse.

Investigation shows the collector brush assembly

for the power cable takeup reel shorted. Enter

LCO 96-0572

October 13 OSRO approves Technical Specification

Clarification 96-003 to permit retesting

refueling bridge.

1:30 pm October 13 Surveillance 24.623, " Reactor Manual

Control / Reactor Mode Switch / Refueling Platform -

Refueling Interlocks," performed. Mode Switch

in Startup/ Hot Standby for about 47 minutes, in

Run for about 7 minutes. Returned to Refuel and

locked upon completion.

6:30 pm October 13 Exit LCO 96-0572. Refueling bridge declared

operable.

4.0 Root Cause and Ma.ior Contributors to the Events ,

I

Based on interviews of appropriate personnel, the inspectors determined l

the following root causes and contributors existed during and prior to

the events.

,

10

i

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

. . J

04.1 RHR Reservoir

The following factors contributed to the event.

e One of the RHR reservoir cross-tie valves (Valve F601A) disk

separated from the valve operator. A set screw on the spline was

not tacked to prevent loosening and becoming disconnected.

'

e Operators and work planners failed to recognize the effect of de- ,

4

energizing bus 72ED had on the UHS.

1

• Operators did not recognize that the plant was in a condition

i requiring action to meet an LCO.

e Operators performed an operability assessment of the UHS using

, non-seismic instrumentation in lieu of valid safety related and

seismic instrumentation that they believed was malfunctioning. l

j 04.2 Failure to Meet TS Reauirements for Control Rods

i

~

e Licensee made a TS interpretation to allow disregarding a valid TS

, requirement.

d

e Insufficient knowledge of regulatory requirements.

05.0 Safety Sianificance

i 05.1 RHR Reservoir

The consequence of this event was minimal because of the conditions of

l the plant during the event. The plant was in the seventh day of an

refueling outage with little decay heat, no activities in progress that

j could result in draining the vessel, no demand for emergency diesels,

and little heat load on the emergency cooling systems. However, the

safety significance of this event was moderate to high due to the number i

e

and specific systems effected. l

e  !

05.2 Failure to Meet TS Reauirements for Control Rods

I NRR Technical Specification Branch determined that the safety l

4

significance of this event was low because the cells with withdrawn

control rods were defueled. This event would not have violated improved

technical specification if improved technical specifications were

applicable to Fermi. However, this event signifies a significant

weakness in using technical clarifications to resolve conflict between

technical specifications without either amending or requesting waiver of

the requirements with a N0ED.

l

11

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

.

. - . . . ~ - - . .

. .

06.0 Corrective Actions

The following corrective actions were either caserved by the inspectors

or verified through documentation reviews. j

!

06.1 RHR Reservoir

The licensee implemented some short term corrective actions. The

'

l

l affected valve (F601A) was repaired. All four cross-tie valves' spline

.

bushing set screws were recessed and lock-tighten.

l

06.2 [ailure to Meet TS Reauirements for Control Rods

The licensee withdrew the technical specification clarification (TSC

96003) on December 20, 1996. The licensee reviewed other current TSCs

>

for similar problems and found none. Currently, the licensee has not

issued a LER or DER documenting this issue.

II. Maintenance i

M1 Conduct of Maintenance

M1.1 Mode Chanae Resultina in Missed TS Surveillance ,

a. Inspection Scone (93702)

The inspectors reviewed various logs and documents associated with the

November 4,1996, event when the plant re-entered Operational Mode 5,

from Mode 4 without performing TS required surveillance. The inspectors

also interviewed both maintenance and operations personnel. The

inspectors also interviewed the appropriate maintenance supervisor.

b. Observations and Findinas

On November 4,1996, the plant re-entered Operational Mode 5, from Mode

4, when a reactor vessel head flange bolt was inadvertently detensioned.

Upon identification that not all reactor vessel head flange bolts were

tensioned, the licensee recognized that they were in Operational Mode 5

and reviewed surveillance requirements. Based on this review, the

i licensee determined that Technical Specification (TS) 4.0.4 requirements

were not met, in that not all surveillances were completed prior to

entry into Operation Condition 5. In this case, the surveillance

requirements for TS 3.3.6, " Control Rod Block Instrumentation," for

Operational Condition 5 were not completed. Because this surveillance

had expired before re-entry into Mode 5, there were less than the

!

minimum required operable channels of intermediate range monitors per

trip function.

l

!

!

l

f

12

- __ - . _ _ . ._

_

. .

.

c. Conclusions

The inspectors determined that on November 5, 1996, the plant re-entered

operational Mode 5 without performing TS required surveillance testing

of the Control Rod Block Instrumentation.

Technical Specification 4.0.4 requires, in part, that entry into an

Operations Condition shall not be made unless the surveillance

requirements associated with the Limiting Condition for Operations have

been performed. On November 5, 1996, entry was made into Operation 1

Condition 5, without the surveillance requirements for Technical '

Specification 3.3.6, " Control Rod Block Instrumentation," being

performed. This is an apparent violation of TS 4.0.4.  !

M3.0 Seouence of Events i

M3.1 Mode Chanae Resultina in Missed TS Surveillance

Initial Conditions: Operational Mode 5

6:02 pm November 4 Head tensioning operations initiated. All 68

head studs installed and hand tightened.

7:27 pm November 4 First pass tensioning (5400 psig) complete.

9:04 pm November 4 Second rass tensioning (7200 psig) complete.

Operations was informed of completion of second

pass. Mode change from Operational Mode 5 to 4

was made. Surveillance for Technical

Specification 3.3.6, " Control Rod Block

Instrumentation," would have been due soon if

the plant remained in Mode 5. With the plant in

Mode 4, the surveillance was no longer required.

9:56 pm November 4 Adjustment pass IAW Procedure 35.710.08 l

initiated.

1:50 am November 5 The Adjustment Pass for final set of four studs

completed.

Later, Maintenance personnel find that stud nut

1. 27 was inadvertently loosen enough to move by

hand.

Stud #27 was re-tensioned to 7200 psig.

2:15 am November 5 The Nuclear Shift Supervisor (NSS) was notified

that maintenance personnel found stud # 27

loose.

2:35 am November 5 Refuel Coordinator went to control to fully

brief NSS on situation. NSS recognized that

when stud #27 was inadvertently loosen that the

plant re-entered Mode 5, and TS 4.0.4

13

.

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

!

i

i

! requirements not met because of an expired

.

technical specification required surveillance.

4

l All. other studs were subsequently checked, no

3

other problems identifled.

M4.0 Root Cause and Ma.ior Contributors

j M4.1 Mode Chanae Resultina in Missed TS Surveillance

e Poor communications between maintenance personnel and Refuel Floor

Coordinator with operations.

4

e Insufficient knowledge of technical specifications.

e Inadequate control of work activities on refueling floor.

i e Inadequate procedure in that all RPV bolting activities were not

completed prior to declaring a change to Mode 4.

j M5.0 Safety Sianificance

! M5.1 Mode Chance Resultina in Missed TS Surveillance

i

j The safety consequence: and significance of this event was minimal.

However, the significance of the root cause, failure to recognize the

impact of plant conditions on technical specification requirements was

! high and of importance.

M6.0 Corrective Actions

M6.1 Mode Chance Resultina in Missed TS Surveillance

The licensee will revise the administrative procedure MOP 13, " Refueling

Operations," to define some actions for changing from Mode 5 to 4.

Procedure 35.710.008, " Reactor Vessel Head Detensioning and Tensioning,"

will also be changed to provide thumbrules for adjustments to stud

tension. A caution or note will also be provided that will require

stopping and getting the refueling floor coordinator verification if

more than a turn of adjustment is required. Checks of stud elongation

data will be made between the reactor cavity and.the official record l

before adjustments will be made to ensure the correct adjustments were

'

made. Finally, the mode change will be made after all trin passes were ,

completed and stud elongation is within tolerances for all studs. l

However, these changes were not developed before the end of the l

inspection and were not planned to be completed until the end of May,

1997.

!

i

14

_

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

. .

V. Manacement Neetines

j X1 Exit Meeting Summary l

The inspectors presented the inspection results to members of licensee

management at the conclusion of the inspection on December 17, 1996. The

licensee acknowledged the findings presented.

l The inspectors asked the licensee whether any materials examined during the

'

inspection should be considered proprietary. No proprietary information was

identified.

,

! .

i

!

"

t

!

i

i

!

,

t

i

!

i

i

<

15

. .

PARTIAL LIST OF PERSONS CONTACTED

Licensee

R. Delong, Superintendent, System Engineering

T. Dong, NSSS, Technical Engineering

P. Fessler, Plant Manager, Operations i

J. Plona, Technical Director  :

P. Smith, Director, Nuclear Licensing '

W. O'Connor, Manager, Nuclear Assessment

N. Peterson, Supervisor Compliance

A. Antrassian, Licensing Engineer

J. Moyers, Director Nuclear Quality Assurance

R. Newkirk, Supervisor, Licensing i

R. Eberhardt, Director, Nuclear Training

LIST OF ACRONYMS USED

CCHVAC Control Center Heating Ventilation Air Conditioning

CFR Code of Federal Regulations

DECO Detrnit Edison Company

DER Deviation Event Report ,

EECW Emergency Equipment Cooling Water i

HVAC Heating Ventilation and Air Conditioning

LER Licensee Event Report

M0V Motor Operated Valves

NRC Nuclear Regulatory Commission l

NSS Nuclear Shift Supervisor '

OSR0 Onsite Review Organization

RHR Residual Heat Removal

RHRSW Residual Heat Removal Service Water

SOE Sequence of Events

S0P System Operating Procedure

TS Technical Specification

TSC Technical Specification Clarification

UHS Ultimate Heat Sink

16