ML20154E648

Insp Rept 50-341/87-48 on 871018-880331.Violations Noted. Major Areas Inspected:Events Surrounding Failure of Licensed Operators to Comply W/Tech Spec Action Requirements Associated W/Reactor Protection Sys Instrument Channel

ML20154E648
Person / Time
Site:	Fermi
Issue date:	05/05/1988
From:	Cooper R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20154E617	List: ML20154E613 ML20154E648
References
50-341-87-48, NUDOCS 8805200277
Download: ML20154E648 (9)

See also: IR 05000341/1987048

Text

.

.

.. .

. .

.

..

U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Report No. 50-341/87048(DRP)

Docket No. 50-341 Operating License No. NPF-43

Licensee: Detroit Edison Company

2000 Second Avenue

Detroit, MI 48226

facility Name: Fermi 2

Inspection At: Fenni Site, Newport, Michigan '

Inspection Conducted: October 18, 1987 through March 31, 1988

Inspectors: W. G. Rogers, Senior Resident Inspector

M. E. Parker, Resident Inspector

Approved By: R. Cooper, Chief k8 $1'

Reactor Projects ection 3B

S/S//J

Datt '

Inspection Summary ,

Inspection on October 18, 1987 through March 31, 1988 (Report

No. 50-341/87048(DRP))

Areas Inspected: Special unannounced inspection by a resident inspector of

the events surrounding the failure of licensed operators to comply with the

Technical Specification action requirements associated with a reactor

protection system instrument channel and of the isolation design configuration

of the primary containment radiation monitor.

Results: Three violations were identified (failure to comply with a Technical

Specification action statement, failure to provide adequate procedure content

and inadeouate primary containment isolation capability).

,

8805200277 880509

PDR ADOCK 05000341

Q DCO

- _.

- , -_ . - -

.

.

,

. .

. .

.

.

DETAILS

1. Persons Contacted

a. Detroit Edison Company

F. Abramson, Operations Engineer

• D. Gipson, Plant Manager

• P. Anthony, Compliance

1. • S. Catola, Vice President, Nuclear Engineering

1. • L. Goodman, Licensing Supervisor

J. Green, Systems Engineering

R. Laubenstein, Nuclear Assistant Shift Supervisor

J. Leman, Director, Plant Safety, Nuclear Production

• R. Lenart, General Director, Nuclear Engineering

L. Lessor, Advisor to Plant Manager

R. Lightfoot, Nuclear Shift Supervisor

1. • W. Omer, Vice President, Nuclear Operations / Plant Manager

J. Plona, Operations Support Engineer

E. Preston, Assistant Director, Plant Safety

B. Sheffel, Nuclear Production, Technical Engineering ISI

F. Svetkovich, Technical Engineer, Nuclear Production

1. • B. R. Sylvia, Group Vice President, Nuclear Operations

1. L. Wooden, Supervisor, I&C

1. L. Fron, Supervisor, Mechanical and Fluid Systems

1. P. Marquart, General Attorney

• W. Tucker, Superintendent, Operations

b. U.S. Nuclear Regulatory Commission

• M. Parker, Resident inspector

1. P. Pelke, Project inspector

1. • W. Rogers, Senior Resident Inspector

1. C. Anderson, Enforcement Specialist

1. R. Cooper, Chief, Projects Section 3B

1. H. Wong, Sr. Enforcement Specialist, OE

1. Dr. C. J. Paperiello, Deputy Regional Administrator

1. R. Knop, Chief, Projects Branch 3

1. M. Virgilio, Deputy Director, ORP

1. T. Quay, Licensing Project Manager, NRR

• Denotes those attending March 30, 1988 exit meeting.

1. Denotes those attending the April 28, 1988 enforcement conference.

2. Review of Drywell Pressure Surveillance Testing

a. Background

in Inspection Report 50-341/87007(DRP), a violation of a limiting

Condition for Operation (LCO) action statement associated with the

high pressure coolant injection / reactor core iso'ation cooling

(HPC1/RCIC) systems was identified. That violation was very

- ,

-

. . l

.

similar to the LC0 violation which occurred during this inspection

period. The previous violation also involved performance of an

I&C surveillance in which licensed operators failed to recognize

the Technical Specification implicaSons,

in response to the HPC1/RCIC violation, the licensee stated that

the violation was attributable to a lack of understanding of plant

conditions and a lack of an impact statement in the procedure.

The licensee further stated that an I&C surveillance procedures

improvement program was in place to upgrade the I&C surveillance

procedures as part of the corrective steps that would be taken to

avoid further violations. This effort included the addition of

impact statements documenting the ramifications, legal and physical,

of perforring a particular surveillance test and to specify the

plant operational conditions under which the test may be performed.

The I&C surveillance improvement program was to be completed by

January 31, 1988.

To nrevent another violation before the completion of the I&C

surveillance improvement procram, the licensee began generating

interim impact statements. The interim impact statements were

being generated as a surveillance became due but were not formally

incorporated into the procedure or approved by the Onsite Review

Connittee (0SRO). The interim impact statement would then be

formalized into that particular surveillance procedure when the

procedure was revised under the total 1&C surveillance procedure

improvement effort.

b. 1.imiting Condition for Operation

On October 24, 1987, the Nuclear Assistant Shift Supervisor (NASS)

signed on Plant Operations Fanual (P0M) 44.020.015, Revision 3,

"NSSS - Drywell Pressure, Division I, Channel A Response Time fest;

C71-N650A and C71-N050A," to allow the instrument and control (I&C)

repairman to perform a response time test of transmitter C71-N050A

and master analog trip unit C71-N650A for the nuclear steam supply

shutoff system (NSSSS) drywell pressure input. Transmitter

C71-N050A is one of two Division I NSSSS high drynll pressure

channels for the reactor protection system instrunantation and the

isolation actuation instrumentation required by Technical

Specification (T.S.) 3.3.1 and 3.3.2, respectively.

At 12:25 p.m. EDT, on October 24, 1987, the I&C repairman defeated

the high drywell pressure instrumentation for channel A, as part of

the response time test. This action rendered the channel inoperable

and the licensee entered into the Technical Specification two-hour

action statement. T.S. Table 3.3.1-1 and Table 3.3.2-1, Table

Notation (a) allows a channel to be placed in an inoperable status

for up to two hours for required surveillance testing without

placing the trip system in the tripped condition provided at least

one operable channel in the same trip system is monitoring that

parameter.

2

_

-

.

-

.

4

At 2:25 p.m. EDT, the two hour Technical Specification allowance

for having the drywell pressure channel inoperable was exceeded

and went unnoticed by all personnel involved with the surveillance.

It was not until 3:15 p.m. EDT, when the I&C repairman reported

to the operating shift that they were having trouble with the

surveillance, that personnel recognized a possible time constraint

problem with the Technical Specifications. The drywell pressure ,

transmitter was subsequently returned to operable status at

3:37 p.m. EDT. This resulted in the high drywell pressure channel

being inoperable from 12:25 p.m. to 3:37 p.m., a total of three

hours and twelve minutes.

This is considered an apparent violation of Technical Specification 3.3.1 and 3.3.2 for failing to place the drywell pressure Channel A

in the tripped condition or to return the inoperable channel to

operable status within the two hour period (50-341/87048-01(DRP)).

c. Inspector Followup

The inspector reviewed the procedures and discussed the event with

the licensee personnel involved. From these reviews the inspector

ascertained that:

(1) An interim impact statement had been generated for this

surveillance procedure. However, in preparing the interim

impact statement for incorporation into the surveillance test

package, the Shif t Operations Advisor (50A) made an error in

that he incorrectly assumed that 'echnical Specifications

allowed the trip channel to be out of service for surveillance

testing for three hours when Technical Specifications only

allows the channel to be out of service for testing for two

hours. This error was further compounded when the operations

engineer concurred with this impact statement.

(2) The incorrect infomation in the non-0SR0 approved interim

impact statement was in contradiction to OSR0 approved

Procedure P0M 44.020.015, Step 4.13.

(3) Even the incorrect time restraints in the interim impact

statement were not adhered to since the chonnel was not placed

in the tripped condition after the three hour time limit had

expired.

(4) No formal means existed for tracking short-tem LCOs. Procedure

POM 21.000.18, "Out-of-Specification Log," does not re% ire

short-term LCOs to be placed in the out-of-specifica* ion log.

(5) I&C personnel were aware that response time testing has

typically taken one shift to complete and they were aware that

this particular response time test would take longer than the

two-hour LC0 limit, but this knowledge was not transmitted to

the operating authority.

3

. _ .

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

-

.

. . . .

(6) During the surveillance, the licensee only performed Sections 6.1,

6.2 and 6.3 of POM 44.020.015. This consisted of performing the

response time test for transmitter C71-N050A and master Analog

Trip Unit C71-N650A. This would not have resulted in any

isolations or actuations. Had this been fully understood prior

to performing the surveillance, drywell pressure Channel A could

have been placed in the tripped condition with no adverse

consequences and full compliance with Technical Specifications

would have been achieved,

d. Impact Statement Review

On February 17, 1988, the inspector reviewed POM 44.020.015 to

ensure that the interim impact statement discussed previously had

been corrected to adequately reflect Technical Specification

requirements and had been formally incorporated into the procedure

thereby receiving OSR0 approval. During this review it was observed

that the impact statement had been incorporated into the procedure

and approved by OSRO. However, during incorporation of the impact

statement into the procedure the licensee failed to incorporate the

correct impact statement. Specifically, the impact statement still

allowed the channel to be inoperable for up to three hours without

placing the channel in the tripped condition instead of the two hours

recuired by Technical Specifications 3.3.1 and 3.3.2.

Further review identified that the licensee had taken the same

action with the associated drywell pressure response time

procedures:

POM 44.020.016, Revision 20, Drywell Pressure, Division II, Channel B,

Response Time Test

POM 44.020.017, Revision 20, Drywell Pressure, Division I, Channel C,

Response Time Test

P0M 44.020.018, Revision 20, Drywell Pressure, Division II, Channel D,

Response Time Test

This is considered an apparent violation (50-341/87048-02(DRP)) of

Technical Specification 6.8.1.d for failing to properly implement

the requirements of T.S. 3.3.1 and 3.3.2 into surveillance

procedures,

e. Sunmary

Although this LC0 violation was caused by inadequate communications

between I&C and Operations, this appears to be indicative of a

breakdown in the overall understanding and appreciation of Technical

Specifications. These sare elements were apparent in the

50-341/87002 violation and as such the corrective actions taken in

response to that violation were inadequate to preclude the current

violation. Also, this event pointed out the weaknesses in the

licensee's system for not tracking short-term Limiting Conditions

for Operation (LCOs).

.

4

-

.

-

. . .

.

3. Primary Containment Radiation Monitoring System (PCRM) Design Configuration

a. Background

The primary containment radiation monitoring (PCRM) system is

configured in a parallel arrangement with the drywell hydrogen /

oxygen sample panel. Both systems no mally operate continuously

during reactor operation and sample the drywell atmosphere from

five zones through primary containment penetrations X-48a through

X-48e. Each of these five penetrations has an air-operated remote

manual isolation valve (T50-F401A, F402A, F403A, F404A, and F405A)

and an associated local manual valve (T50-F033A, F034A, F035A,

F036A, and F037A).

The original isolation design for the PCRM system and the drywell

hydrogen / oxygen sampling system was an acceptable alternative to

GDC 56 described in the FSAR Section 6.2.4. Containment isolation

requirements were achieved using a single isolation valve (T50-F401A

through F405A) and this was based on a closed system outside the

containment. The basis of a single remote manuhl isolation valve

is described in the UFSAR Table 6.2.2 and Section 6.2.4.2.2.3.2.

This design assumed that the PCRM system would operate following a

loss-of-coolant accident (LOCA) and the PCRM system would be in

compliance with the closed system requirements.

In January 1984, the licensee determined that the PCRM system did

not comply with the specific closed system requirements. Specifi-

cally, the system was not qualified for containment design pressure

and problems were noted with the sei."nic and material certifications

provided by the vendor. The licensee subsequently determined that

the PCRM was a non-essential system following a LOCA ard should be

automatically isolated upon receipt of a LOCA signal. As such, in

early 1984, the two automatic isolation valves (T5P-Fl50 and F451)

and the two manual isolation valves (T50-F063 and ' N4) were added

to isolate the PCRM on a LOCA signal (high drywell pressure).

The automatic isolation valves were added to provide the isolation

of the nob r.:n-essential PCRM system and was intended to return the

system to that of a closed system configuration. The licensee

believed this configuration was an acceptable alternative to GDC 56.

This configuration resulted in providing two barriers in the event

of a LOCA and failure of the PCRM boundary, the first barrier being

the newly installed automatic isolation valves (T50-F450 and F451)

and the second barrier being the remote manual isolation valves

(T50-F401A through F405A).

b. Licensee Event Report (LER)

On October 17, 1987, during implementation of Engineering Design

Package (EDP) 1786 on the primary containment radiation monitoring

(PCRM) system, the primary containment isolation was questioned as

to the adequacy of containment integrity. The isolation boundary

utilized was two solenoid-operated valves (T50-F450 and F451) which

5

.

.

. . . .

.

'

are not primary containment isolation valves. The containment

isolation valves for this penetration (X48a throu

manual isolation valves (T50-F401A through F405A)gh and receive X48e)

no are remote

automatic isolation signals. The primary containment isolation

valves were open dur'ng implementation of this EDP. It was further

detennined that valves T50-F450 and F451 were not properly qualified

to satisfy containment integrity requirements.

On October 17, 1987, at 8:50 p.m. EDT, the licensee determined this

was a potential loss of the primary containment integrity. The

licensee took immediate action to isolate the primary containment

boundary. The Division 1 PCMS was subsequently shut down and

isolated. The isolation consisted of closing primary containment

isolation valves T50-F401A through F405A for the Division 1 PCMS.

This resulted in the plant being in a seven days and 30 day Limiting

Condition for Operation as a result of having the primary containment

H2/02 monitoring system and radiation monitoring system, respectively,

out of service. After investigation into the isolation, the licensee

determined the inadeouate isolation was a reportable event and on

October 17,1987, at 9:30 p.m. EDT the licensee made the applicable

notifications per 10 CFR 50.72 for a primary containment integrity

violation.

Local leak rate testing (LLRT) on T50-F450 and T50-F451 was

satisfactorily completed on October 18, 1987, and the applicable

out of service log was subsequently cleared for these valves. This

allowed the licensee to isolate the PCMS radiation monitor utilizing

T50-F450/F451 and open the designated containment isolation valves

T50-F401A through F405A. This action allowed the H2/02 monitor to

be placed in service and took the plant out of the seven day LC0.

However, a 30 day LC0 was still in place for not having the PCMS

radiation monitor in service.

Discussions with the resident inspector concerning the configuration

resulted in the inspector questioning the current design configuration

of the PCRM system to meet 10 CFR 50, Appendix A, General Design

Criterion (GDC) 56 requirements.

c. Exemption Request

On October 27, 1987, in DECO Letter No. NRC-87-0211, the licensee

requested a temporary exemption from the requirements of 10 CFR 50,

Appendix A, Criterion 56 (GDC 56), Primary Containment Isolation.

This request was a result of review by the NRC in detennining that

the current design configuration, for the Division I primary

containment monitoring system, did not meet the requirements of

GDC 56. This exemption request along with other correspondence

identified the licensee's proposed course of action to return the

. primary containment radiation monitoring system to service utilizing

the current isolation design.

On November 13, 1987, the NRC granted to the licensee an exemption

to General Design Criterion 56 of Appendix A to 10 CFR Part 50.

!

6

L _ _ _ - _ _ _ _ _ _ _ _ _

h

.

.- . .

This exemption permitted postponement of full compliance with GDC 56

for the primary conteinment radiation monitoring isolation until

startup following the planned local leak rate testing in March

1988. To support operation pending incorporation of modifications

of the PCRM isolation, while the exemption is in effect, the

licensee comitted to upgrade the effectiveness of the isolation

scheme as described in NRC letter, dated November 13, 1987. This

action included treating valves T50-F450, F451, F040, F046, F063

and F064 as primary containment isolation valves, in a manner

consistent with any other valve listed in Technical Specifications.

The licensee also comitted to revise the Emergency Operating

Procedures and enhance operator training as an interim compensatory

measure.

On January 29,(License Amendment) change which results from1988, the lice

Specification

modifications to bring the PCRM isolation design up to the standards

set forth in GDC 56. On March 29, 1988, NRR issued Amendment 17 to

the operating license in response to the January 29, 1988 letter,

d. Inspector Followup

In reviewing the PCRM design, the licensee was unable to find any

correspondence accepting this configuration as an acceptable  !

alternative to GDC 56. Review of the UFSAR identified that it had

not been updated to reflect the current design configuration and

categorized penetrations X48a through X48e as Engineered Safety

Feature (ESF) or ESF-related system penetrations and that these

penetrations are attached to a closed system. Technical

Specifications (TS) had also not reflected this modification to

include the additional automatic isolation valves (T50-F450 and

F451). In addition, the licensee's program in 1984 to install

these valves failed to properly maintain these valves in accordance

with the applicable requirements of GDC 56,10 CFR 50 Appendix J

and other testing requirements (functional testing, logic testing,

positive indicator checks, LLRT testing, etc.).

The PCRM containment isolation system design as described in UFSAR

Section 6.2.4 does not reflect the as-built system as required by

10 CFR 50.34(b). The deviation between the UFSAR and the as-built

system was not evaluated in accordance with 10 CFR 50.59. This is

an apparent violation (50-341/87048-03(DRP)).

e. Enforcement Conference

An Enforcement Conference was held in the Region 111 Office on

April 28, 1988 to discuss the PCRM containment isolation system

design. No new information was provided.

4. Exit Interview (30703)

The inspectors net with licensee representatives (denoted in Paragraph 1)

on March 30, 1988, and informally throughout the inspection period and

7

- .. _ _

__ ._

. . - _

.

...

.

. . .

'

i

'

summarized the scope and findings of the inspection activities. The

inspectors also discussed the likely informational content of the

inspection report with regard to documents or processes reviewed by the

inspectors during the inspection. The licensee did not identify any

such documents / processes as proprietary. The licensee acknowledged the

findings of the inspection.

!

l

,

>

l 8

l

- - _ -_ - -. - __ L