ML16342A113

under accident conditions;

(2) the licensee

discontinued

procedures,

training,

and calibration of the

SENTRY gas

chromatograph

for

dissolved reactor coolant

hydrogen analysis,

and failed to perform

a written

safety analysis

(per

10 CFR 50.59) of this change;

and

(3) the licensee failed

to implement

and maintain

a program (per-Technical

Specification 6.8.4) that

ensured

the capability to obtain

and analyze

samples of radioiodines

and

particulates

in plant gaseous effluents

under accident conditions.

93060i 0230 9305i 3

PDR

ADOCK 05000275

PDR

0

DETAILS

Persons

Contacted

Licensee

Baxter,

Chemistry Instructor

Boots, Director, Chemistry

Bosseloo,

Instrumentation

and Control

(I&C) Lea

Project Engineering

Group

Carlsen,

Engineer,

Regulatory

Compliance

Ehrhardt,

Engineer,

Radiation Protection

(RP)

Fong,

Engineer,

Gardner,

Senior Engineer,

Chemistry

Gray, Director,

(RPH)

Grebel,

Supervisor,

Regulatory Compliance

Helman,

ALARA Engineer

Hess, Assistant Onsite Project Engineer

Irving, General

Foreman,

Hiklush, Hanager,

Operations

Services

Holden, Director,

I&C

Houlin, Assistant to Vice President

Mosher,

Engineer,

guality Assurance

(gA)

Rising, Auditor, gA

Snyder,

Senior Chemistry Instructor

Somerville,

Senior Engineer,

Taylor, Senior Engineer

Thierry, Senior

Engineer,

Regulatory Compliance

Wessel,

Engineer,

Chemistry

p

tJ

F

d Engineer,

Onsite

tE.

S

H.,

tJ.

R

T-.

C.

R

T

tD.

+J

T

H

W.

R.

H.

tA.

tR.

tE

HRC

F.

Gee,

Resident

Inspector

H. Hiller, Senior Resident

Inspector

(t) Denotes

those individuals who participated in the teleconference

calls

on April 29 and 30,

1993.

Occu ational

Ex osure

83750

83729

(*) Denotes

those individuals who attended

the exit meeting

on April 9,

1993.

The inspector

met

and held discussions

with additional

members of

.the licensee's

staff during the onsite inspection.

At the time of the inspection,

Unit 2 was in

a refueling outage

(2R5).

The inspectors

examined this program area

by performing tours

and

independent

surveys of the facility, observing work in progress,

reviewing procedures

and records,

and interviewing cognizant personnel.

Observations

were

made regarding radiological posting

and labeli'ng,

surveys

and monitoring,

and control of posted

high radiation areas.

a.

Radiolo ical Postin

and Labelin

While conducting tours of the Auxiliary Building, Containment

Building, and various other areas within the licensee's

Protected

Area (PA), the inspectors

observed

and verified the licensee's

radiological posting

and labeling.

For those

areas

observed,

radioactive material labels,

as well as postings for radiation,

high

radiation,

and radioactive materials

areas,

were visible, accurate,

and current.

b.

Surve

s

and Monitorin

The inspectors

examined the licensee's

surveys,

survey logs,

and

instrument

issue logs.

In addition,

one inspector

accompanied

a

contract

RP technician

(RPT)

on

a routine survey of the Unit I PA,

Warehouse

A, and the

I&C instrument calibration facility.

The following licensee

procedures

were reviewed in assessing

the

licensee's

program implementation:

~

RCS-7,

"Radiation Control Standard - Surveys,"

Revision

6

~

RCP D-500, "Radiation

and Contamination

Surveys,"

Revision

9

~

RCP D-501, "Issue

and Return of Radiation Protection

Equipment," Revision I

~

RCP D-510, "Radiation

and Contamination

Survey Program,"

Revision

5

The inspectors

made specific observations

related to survey

practices,

worker usage of portal monitors

and friskers,

records of

surveys

and supervisory

review of survey information, instrument

availability,

and .adherence

to procedural

requirements.

il

(I)

Surve

Practices

Upon review of survey logs

and instrument

issue logs, the

inspectors

noted that surveys

were being performed

and

documented

thoroughly.

Routine surveys

were being performed at

the required frequency.

Observation of surveys

and discussions

with workers revealed that

RP personnel

had

a thorough

under standing of the survey program,

instrument

requirements

and limitations, techniques

for taking

and counting

smears,

and

ALARA practices.

The licensee's

routine survey practices

v.'ere

found to be .adequate

for monitoring and posting radiation

and

high radiation

areas

and were consistent with 10 CFR 20.201,

"Surveys."

While the inspector

was

accompanying

an

RPT on

a routine survey

of the Unit I PA,

a discrete radioactive particle ("hot

3

particle" ) was found on the asphalt

between

the north gate of

the 115'adiological

Controlled Area

(RCA) Backyard

and

Warehouse

A.

Readings with a count-rate

meter were in excess

of 50,000 corrected

counts per minute at 1/2 inch from the

particle.

The licensee

successfully

removed the particle from

the asphalt,

performed

a spectral

analysis,

and disposed of the

particle appropriately.

Additional

RPTs were sent to perform

a

more thorough survey of the area.

No other particles

were

found.

(2)

Worker Usa

e of Portal Monitors and Friskers

The inspectors

observed

worker usage of portal monitors

and

friskers at the 85'uxiliary Building and the 140'ontainment

Building access

control points.

The inspectors

noted that

workers'ontamination

monitoring practices

were adequate.

(3)

(4)

(5)

Records of Surve

s

and

Su ervisor

Review of Surve

Information

The inspectors

found the licensee's

maintenance

of survey

records to be acceptable.

Licensee

super visory review of

survey results

was performed efficiently, and survey data

and

information was disseminated

in a timely manner for use in work

planning

and radiation dose control.

Instrument Availabilit

The availability of RP instrumentation

was observed

at the

85'uxiliary

Building instrument

room.

The inspectors

noted that

few E-140 friskers

and telescoping

survey instruments

were

available for use.

The

RP Foreman

on shift indicated that the

majority of instruments

were in the Containment Building for

the outage.

The

RP Engineer responsible for coordinating

instrument calibration

and maintenance

with I&C stated that the

timeliness of instrument calibration

and maintenance still

needed

improvement,

but had improved over previous

outages.

All instruments

observed

by the inspectors

had current

calibration

and performance

test, stickers.

The inspectors

noted that Procedure

RCP D-500, "Radiation

and

Contamination Surveys," listed

an E-520

as

an available

instrument for use,

but cross-checks

against

the instrument

inventory revealed that the licensee

did not possess

any E-520

instruments.

An equivalent instrument,

an ASP-1 audible

response

with a Geiger-queller

probe,

was available in the-

licensee's

instrument inventory.

Adherence

to Procedural

Reouirements

The inspectors

reviewed survey practices for procedural

adherence.

Several

items'were

noted:

4

(a)

Review of survey logs for the period of January

1993

through Harch

1993 indicated. that routine surveys

were

performed at the frequencies

required

by Procedure

RCP

D-510, "Radiation

and Contamination

Survey Program."

(b)

Surveys

were thoroughly documented

as required

by

Procedure

RCP D-510, "Radiation

and Contamination

Survey

Program."

(c)

Other survey practices

observed

(e.g.

smear techniques,

instrument use;

preparation for surveys)

were performed

as

required

by Procedure

RCP D-500, "Radiation

and

Contamination Surveys."

Control of Posted

Hi

h Radiation Areas

Technical Specification (TS) 6.12. 1 requires that high

radiation areas

in which the intensity of accessible

whole body

radiation is greater

than

100 millirem per hour (mrem/hr) but

less

than or equal to 1000 millirem per hour {mrem/hr) at 18

inches shall

be barricaded

and conspicuously

posted

as

a high

radiation area.

Any individual or individuals permitted to

enter

such

an area

must have either:

(1)

a radiation

dose rate

survey meter,

or (2)

an alarming dosimeter

and prior knowledge

of the area

do'se rates,

or (3) accompaniment

by an individual

q'ualified in RP procedures

who carries .a survey meter,

provides

positive control over activities in the area,

and performs

'eriodic

radiation surveys.

The inspectors

reviewed four instances

in which workers

had

entered

posted

high radiation areas

(HRAs) without meeting

TS

requirements for HRA controls.

The instances,

and the

licensee's

corrective actions for each

case,

were

as follows:

{1)

On October

2,

1992, during the Unit

1

1R5 outage,

a

contract worker entered

the refueling bridge posted

HRA to

look at

a reactor vessel

in-service inspection tool.

The

worker did not have either

a survey meter,

an alarming

dose rate meter,

or coverage

by an

RPT.

In addition, the

worker had not been briefed

on radiation levels present

in

the

BRA.

The worker was observed

by

a deconner,

who

called

an

RPT,

who in turn instructed the worker to exit

the

HRA.

As corrective action for this problem,

the worker's

RCA

access

authorizagion

was suspended,

and the worker was

subsequently

terminated.

In addition, worker training on

this incident was given prior to the Unit 2 2R5 outage.

(2)

On October

26,

1992, during the Unit

1

1R5 outage,

two

contractor carpenters

entered

a posted

HRA on the

140- oot

elevation of 'the Unit

1 Containment.

A nearby

PPT

'bserved

these entries,

noted the workers'ack of

monitoring devices,

and promptly removed the workers from

the

HRA.

As correc'tive action for this problem,

both workers'CA

access

authorizations

were temporarily placed

on hold,

and

the event

was discussed

with their supervision.

Subsequent

training on the event

(and

on TS requirements

for HRA entries)

was given to the entire crew on October

27,

1992.

(3)

On Harch 15,

1993, during the Unit 2 2R5 outage,

three

contractors

were observed

by a quality control inspector

.to be working without proper radiation monitoring in the

posted

HRA adjacent to reactor

coolant

pump

(RCP) 2-2.

The workers subsequently

stated that they had 'entered

the

area

by climbing. through

a narrow gap in the floor

grating,

and

had not

known that they were entering the

HRA.

In investigating this problem,

the. licensee

noted that the

workers

had received

a pre-job briefing on radi ation

hazards

by the

RPTs at the 115-foot elevation of

Containment;

however,

the workers

had not informed the

RPTs of the full scope of the job,

and therefore

had not

been briefed

on the radiation levels present

in the

HRA

beneath

the floor grating.

As corrective action, the

workers were counselled.

(4)

On March 22,

1993, during the Unit 2

2R5 outage,

two

contractor wor'kers were found without proper monitoring

inside the pressurizer

shed

posted

HRA on the 160-foot

elevation of the Unit 2 Containment.

The workers

had

checked

in with the access'ontrol

RPT,

who had forgotten

to issue

them alarming dosimeters.

As corrective action,

both the access

control

RPT and the

workers were counselled.

During discussions

with the inspectors,

the

RP Director (RPll)

noted that the percentage

of error evidenced

by these four

instances

was relatively small in proportion to the number of

properly monitored

HRA entries during outage periods.

The

RPH

noted,

in addition, that in each

case

subsequent

surveys

and

evaluations

had -concluded that the workers

had not actually

entered

those portions of the respective

HRAs in which the dose

field exceeded

100 millirem per'hour.

The inspectors

acknowledged

these

statements,

but observed

that,

by entering the posted

HRA, the workers

had in each

case

circumvented

the final

HRA control (i.e., the posted

sign

and

barricade)

and gained

access

to the high dose fields.

The

6

IC

inspectors

also observed that the licensee's

corrective actions

thus far had not been effective in preventing recurr ence of

this problem.

The inspectors

concluded that the failure to

adhere to

HRA entry controls, in the four instances

noted

above,

constituted

a violation of TS

6'2 F 1 (50-275/93,-11-01)

In response

to the inspectors'bservations,

the

RPH stated

that, in an effort to prevent additional

improper

HRA entries,

additional controls were being evaluated.

These controls under

evaluation included:

(1) increased

emphasis

on

HRA controls

during general

employee training; (2) posting additional,

even

more conspicuous

HRA signs;

(3) using different

HRA barricades;

(4) "tightening" the

HRA boundaries (i.e., restricting the

HRA

boundaries

to the immediate

area of the high dose fields);

and

(5) inclusion of more conspicuous

markings

on radiation work

permits to remind workers of HRA monitoring requirements

prior

to beginning

a job.

With the exceptions

noted,

the licensee's

programs for controlling

occupational

exposure

appeared

effective in the areas

observed.

One

violation of NRC requirements

was identified.

3.

Post-Accident-Sam lin

84750

92701

V

This issue

was previously discussed

in

NRC Inspection

Report 50-275/93-04

and 50-323/93-04,

and was identified as Unresolved

Item 50-323/93-04-01

~

On the basis of the following discussion,

the unresolved

item is

considered

closed.

The inspectors'eview

of this area

was to determine

whether the

licensee's

program for obtaining post-accident

samples of reactor coolant

and plant gaseous

effluents

met the requirements

of TS 6.8.4.e

and the

commitments to NUREG-0737

as given in the Updated Final Safety Analysis

Report,

Chapter

9 '.2.2,

and historical

correspondence

between Pacific

Gas

and Electric Company

(PGSE)

and the

NRC.

In conducting this review,

the inspectors

reviewed applicable

licensee

procedures,

calculations,

maintenance

records,

and'elevant

historical correspondence,

and

discussed

the p'ost-accident

sampling

system

(PASS) history and

performance with various

members of the Chemistry Department

and licensee

management.

Discrepancies

were noted regarding

two aspects

of the licensee's

program:

(1) sampling

and analysis of dissolved

hydrogen in reactor

coolant;

and

(2) sampling

and analysis of radioiodines

and particulates

in plant gaseous

effluents.

Samolin

and Anal sis of Dissolved

H droaen

in Reactor Coolant

{1)

Requirements.

Criteria

and

Commitments

TS 6.8.4 requires

the licensee,

in part, to establish,

implement.

c.nd

ma ntain

a program l".hich will ensul e t,1e

capability to obtain

and analyze reactor coolant under accident

conditions.

The program must include (1) training of

personnel,

(2) procedures

for sampling

and analysis,

and (3)

provisions for maintenance

of sampling

and analysis. equipment.

r

UFSAR Section 9.3.2.2,

"Post-LOCA Sampling System," further

describes

the licensee's

methods of implementing this program,

including the system capability for quantifying dissolved

hydrogen in reactor

coolant.

The

UFSAR commitment states

that

all sampling

and analyses,

"as required

by NUREG-0737

(November

1980),"

can

be done within a 3-hour period.

NUREG-0737

(November 1980), "Clarification of THI Action Plan

Requirements,"

Section II.B.3, provides criteria for

postaccident

sampling capability.

These criteria include:

(a)

The capability of providing, within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> of deciding to

take

a sample,

quantification of either hydrogen

gases

or

total dissolved

gases

in the reactor coolant;

(b)

The capability of performing backup grab samples for any

parameter

using inline monitoring;

and

(c)

Use of the dose limits of General

Design Criteria

(GDC) .19

of 10 CFR. 50, Appendix A (i.e.,

5 rem whole body, .75.rem

extremities)

as the design basis for any individual

performing postaccident

sampling

and analysis.

PG&E Letter DCL-85-047, dated

February

1,

1985,

committed to

having the Diablo Canyon Unit 2 SENTRY

PASS operable prior to

criticality.

Enclosure

1 to this letter stated that the

SENTRY

PASS was "the permanent

system for providing post accident

sampling capability,"

and reiterated

PG&E's commitment to

meeting the criteria of NUREG-0737,Section II.B.3.

Finally,

10 CFR 50.59 requires,

in part, that the licensee

'hall

maintain records of changes

to the facility or procedures

described

in the

UFSAR, including

a written safety evaluation

that provides the basis for determining that the change

does

not involve an unreviewed safety question.

'I

Back round

The

SENTRY

PASS method of quantifying dissolved

hydrogen in

reactor coolant consisted of a shielded off-line system,

A

small, pressur'ized

sample of reactor coolant liquid was

,isolated within the system

and evacuated'o

a gas

bomb.

The

offgas

was routed through

a shielded

gas

chromatograph

(GC) to

obtain

a hydrogen concentration

readout.

After several

years of using the

SENTRY

PASS,

the licensee

developed

a concern

regarding moisture carryover from the

8

liquid sample through the gas

bomb and into the

SENTRY GC.

With routine use of the system (for training and operational

checks),

the carried-over moisture periodically wetted

down the

GC columns,

rendering the

GC inoperable until the columns could

be replaced.

Since the

SENTRY

GC was also

used for analyzing

dissolved

oxygen in the reactor coolant

and hydrogen in the

containment

atmosphere,

this matter

was of additional

concern.

~h

As an improvement to the system,

the licensee

decided to add

in-line reactor coolant hydrogen monitors.

These monitors,

called exosensors,

were designed with a gas-permeable

membrane.

Hydrogen passing

through'the

membrane

reacted with a sulfuric

acid solution,

and this reaction

was measured

to quantify the

hydrogen concentration.

The licensee's

evaluation of adding the exosensors,

performed

in accordance

with 10 CFR 50.59,

was completed in December

1987 'he evaluation stated that the

SENTRY

GC would continue

to be maintained

as

an alternate

method of analysis.

The

evaluation

concluded that the

VFSAR analysis

was unaffected

by

this change (i.e.,

by adding the exosensors).

The Unit 2 exosensor

was installed

and plant-accepted

on

December

8,

1989,

and the Unit

1 exosensor

was installed

and

plant-accepted

on January

15,

1990.

After more than

a year of

observing

exosensor

performance

in both units, operation of the

exosensors

was

added to the

PASS emergency

procedures

on March

27,

1991.

Due to continued

problems with moisture carryover in the

SENTRY

GC, the licensee

discontinued

use of the

GC for dissolved

reactor coolant hydrogen analysis (for both Units

1

and 2).

On

August 18,

1992, this method of analysis

was removed

from the

PASS emergency

procedures,

and training in this method

was

discontinued for PASS operators.

Comparisons of SENTRY

GC

reactor coolant hydrogen to normal reactor coolant hydrogen

samples,

formerly used to determine calibration of the

SENTRY

GC method,

were

no longer performed after this date.

NRC Evaluation

As previously noted in

NRC Inspection

Report 50-275/93-04

and

50-323/93-04,

the Unit 2 exosensor

performed

much more poorly

than the Unit

1 exosensor.

From August

1,

1992, to the time of

the onsite inspection,

the licensee

had considered

the Unit 2

exosensor

to be inoperable

during the following periods:

0

Peri od

August 24,

1992 - October 15,. 1992

November 6,

1992,

December

10,

1992

December 31,

1992 - February

12,

1993

February

25,

1993

April 9,

1993

T~otal

Da

s

53

35

44

During the

same period,

the Unit

1 exosensor

was only

inoperable for brief periods of routine maintenance.

The inspector

noted the following deficiencies related to the

licensee's

PASS capabilities for reactor coolant dissolved

hydrogen:

(a)

By invalidating the procedures,

training,

and calibration

for the

SENTRY

GC method of analysis,

this method

no

longer met the program requirements

of TS 6.8.4.

Despite

this fact, the

SENTRY

GC method

was still listed

as

an

alternate

method of analysis

under Equipment Control

Guideline

(ECG) 11.1,

"Post Accident Sampling System."

(b)

Other alternate

methods

given in

ECG ll.1 included normal

laboratory analysis of liquid grab samples

obtained

from

the

PASS sample lines or normal reactor coolant

sample

lines.

Using these

methods,

however,

the licensee

was

unable to perform sampling

and analysis within the design

basis

dose criteria of GDC 19 under accident conditions

(as outlined by NUREG-0737 criteria).

(c)

As

a result, for both Units

1

and 2, inline monitoring of

reactor coolant dissolved

was not supplemented

by

backup grab sampling capability,

as outlined by NUREG-0737

criteria.

(d)

During the periods of exosensor

inoperability in Unit 2,

no method of sampling

and analyzing reactor coolant

dissolved

under accident conditions

met the

criteria of NUREG-0737.

(e)

The licensee's

decision to discontinue

the procedures,

training,

and calibration of the

SENTRY

GC for reactor

coolant dissolved

hydrogen analysis

was not within the

scope of the

10 CFR 50.59 evaluation

performed in December

1987.

The licensee

had not performed

any additional

evaluation of this change'to

the

PASS capability.

(5)

Licensee

Evaluation

and Corrective Action

The inspectors

noted that enclosures

to the licensee's

Februart',

1985,

commitment letter had listed liquid grab samples

(drawn from either the

PASS lines or normal reactor coolant

sample lines)

as alternate

methods of meeting

NUREG-0737

10

P

criteria.

In discussions

with the inspectors,

the licensee

stated that

PGFE

had never performed evaluations

to demonstrate

that these alternate

methods

could. meet the design basis

dose

criteria of GDC-19.

The licensee

had incorrectly assumed

that

normal laboratory analysis of liquid grab samples

were always

an available alternative to the exosensor

or SENTRY GC.

During an April 14,

1993,

conference call, the licensee

discussed

proposed

corrective actions with members of NRC

Region

V management.

On April 15, the licensee

submitted

PGEE

Letter

DCL-93-089, which committed to the following corrective

actions:

(a)

The

SENTRY

GC method of quantifying reactor coolant

dissolved

hydrogen would be upgraded

by adding

a liquid

coalescing filter to reduce moisture carryover

and extend

thermal conductivity detector life.

This upgrade

would be

completed in Units

1

and

2 by April 21,

1993.

The letter

noted that the

SENTRY

GC method

was considered

a remote

grab sample.

{b)

The licensee

had already revised applicable

procedures

to

include the

SENTRY

GC method of quantifying reactor

coolant dissolved

hydrogen.

In addition, calibration

and

interim training for this method would be completed

by

April 21,

1993.

Training of all

PASS assigned shift

chemistry

and

RP technicians

would be completed

by Hay 15,

1993.

(c)

The Unit 2 exosensor

would be returned to operable status

prior to restart of Unit 2.

During an April 21,

1993,

conference call, the licensee

stated

that the design

change installing coalescing filters in the

SENTRY

GC lines

had

been

completed

in both units.

In addition,

the licensee

stated that interim training and calibration

had

been completed,

and that the Unit 2 exosensor

had

been repaired

and declared

operable.

Finally, the licensee

stated that

calibration of the exosensor

would be verified against

normal

laboratory analysis of reactor coolant

upon restart of Unit 2.

{6)

Conclusion

The inspectors

determined that the licensee

had failed to

implement

and maintain

a program that ensured.

the capabilities

for sampling

and analysis

required

by TS 6.8.4,

'in the

following aspects:

(a)

At the time of discontinuing the procedures,

training,

"no

calibration for the

SENTRY

GC method of quantifying

reactor coolant dissolved

hydrogen,

the licensee

had

removed grab samplino capability

as

a bacl:up to inline

11

monitoring.

{b)

During the periods of Unit 2 .exosensor

inoperability,

no

alternate

method

was available capable of quantifying

reactor

coolant dissolved

under the specified

.

accident conditions without exceeding

GDC-19 dose

criteria.

The inspectors

concluded that these failures constituted

an

apparent violation of TS 6.8.4 {50-323/93-11-02).

. The.

inspectors

concluded,

further, that the failure to perform

a

written safety evaluation of the effects of invalidating the

SENTRY

GC method of quantifying reactor coolant dissolved

hydrogen constituted

an apparent violation of 10 CFR 50.59 (50-

523/93-11-03).

In an effort to establish

the overall safety

consequence

of

these deficiencies,

the inspectors, noted that normal laboratory

analysis of reactor coolant dissolved

hydrogen while meeting

GOC-19 dose criteria was possible

under

some accident

conditions,

and that the conditions outlined in NUREG-0737

represented

worst-case

accident conditions.

The inspectors

also noted,

however, that the licensee's

PASS equipment control

'uidelines

and surveillance test procedures

did not establish

the severity of the accident for which these

"alternate

methods" could be used.

In addition, the inspectors

reviewed portions of the licensee's

emergency

plan

and discussed

the use of the reactor coolant

dissolved

hydrogen analysis with applicable

members of the

licensee's

emergency

organization.

The inspectors

noted that

this analysis

was not used

as

a primary means of assessing

core

damage

under accident conditions.

Rather, it was considered

supplemental

information used

in verification and long-term

analysis of core conditions.

b.

Sam lin

and

Anal sis of Radioiodines

a'nd Particulates

in Plant

Gaseous

Effluents

Re uirements

Criteria

and Commitments

TS 6.8.4 requires

the licensee,

in part, to establish,

implement,

and maintain

a program which will ensure

the

capability to obtain

and analyze

samples of radioiodines

and

particulates

in plant gaseous

effluents under accident

conditions.

The program must include (1) training of

personnel,

(2) procedures

for sampling

and analysis,

and

(3)

provisions for maintenance

of sampling

and analysis

equipment.

UFSAR Section

11.4 further describes

the licensee's

methods cf

implementing this program,

including

a basic description of i'...,

use

and functions of the midrange plant vent iodine monitor

12

{RE-32) and high-range plant vent iodine sampler

(RX-40).

In a letter to=the

NRC dated April 15,

1982,

PG&E committed to

meeting the criteria of NUREG-0737,Section II.F. 1, Attachment

2 for the gaseous

iodine and particulate

sampling

system in

Diablo Canyon Unit 1.

In a letter to the

NRC dated July 21,

1982,

PG&E reiterated this commitment,

and stated that RX-40

had

been

made operational

as of July 8,

1982.

NUREG-0737

(November 1980), "Clarification of THI Action Plan

Requirements,"

Section II.F. 1, Attachment 2, provides criteria

for the licensee's

post-accident capability of quantifying

radioiodines

and particulates

in plant effluents.

These

criteria include:

(a)

Concentration of 100 uCi/cc of gaseous

radioiodines

and

particulates,

deposited

on the sampling media;

(b)

30 minutes of sampling time;

(c)

An. average

gamma energy of 0.5 NeV;

and

{d)

Design of the sampling

system

such that plant personnel

could remove

samples,

replace

sampling media,

and

transport the

samples

to the onsite analysis facility with

radiation exposures

not in excess

of the

GDC-19 dose

criteria.

In addition,Section II.B.2 of NUREG-0737 provides the source

term to be used in reviewing plant shielding design

and in

determining the accessibility of vital. areas

during post-

accident operations.

(2)

~Back round

As originally installed,

the licensee

had several

monitors

and/or samplers

capable of quantifying plant vent radioiodines

and particulates

under various conditions:

(a)

RE-24, the normal

range monitor,

was located

on the

115'levation

of the Fuel Handling Building in the Plant Vent

Room.

RE-24 provided

a local readout

and alarm function,

as well

as

a remote

alarm in the Control

Room.

In

addition,

RE-24 was equipped with a particulate filter "nd

silver zeolite iodine cartridge that could

be

removed

and

analyzed

as

a grab sample.

The approximated

useful

range

of RE-24 was

from

1 E-7 microcuries

per cubic centimeter

(uCi/cc) to

1

E-4 uCi/cc.

(b)

RE-32, the midrange monitor,

was located adjacent to PE-2-'.

on the 115'levation of the Fuel Handling Building.

PE-

32 provided

remote monitorino, strip chart

'. ecorder,

-nc.'

13

alarm functions in the Control

Room,

and could be remotely

operated

and purged.

In addition,

RE-32 was equipped with

a particulate filter and silver zeolite iodine cartridge

that could

be removed

and analyzed

as

a grab sample.

The

approximated

useful

range of RE-32 was from 1.3 E-7 uCi/cc

to 3 E-3 uCi/cc.

(c)

RX-40, the high-range

sampler,

was located outside at the

85'levation against

the north wall of the Fuel Handling

Building.

RX-40 was provided with sampling functions only

(no monitoring or alarm functions),

and

was operated

remotely from the Control

Room.

A particulate filter and

silver zeolite iodine cartridge

was

housed

in a shielded

lead pig assembly,

which could be extracted

into

a larger,

lead-shielded

cart for transport to the Technical

Support

'enter

for analysis

as

a grab sample

under accident

conditions'.

The approximated

useful

range of RX-40 was

from

1 E-9 uCi/cc to

1

E+2 uCi/cc.

(3)

S stem Modifications

The licensee

was in the pro'cess of installing upgrades

in both

units for various process

and effluent radiation monitors.

As

part of this upgrade

program in Unit 1,

RE-24 had

been

removed,

and

a portable

RADECO pump assembly

had

been installed in the

former RE-24 sampler location.

This assembly

was provided with

a particulate filter and silver zeolite iodine cartridge for

periodic grab sampling.

In review of Unit

1 Control

Room logs,

the inspectors

noted

that the licensee

had taken both

RE-32

and

RX-40 out of service

for a 12-day period from February

26

March 9,

1993.

As part

of continued monitor upgrade

work, the monitors

had

been in

various conditions of inoperability during this time (e.g.,

clearances

hung with supply breakers

open,

leads lifted, heat

trace inoperable).

(4)

NRC Evaluation

'he

inspectors

noted that log entries

on February

26,

27,

and

28 gave

a list of monitors out of service

and then stated,

"Ho

alternate

sampling available for PASS" [referring to the plant

vent radioiodine/particulate

monitoring capability only].

Although subsequent

log entries

recorded

the

same monitors

as

being out of service

from February

26 - March 9,

no entries

regarding

PASS sampling capability were

made after the

February'8

entry.

p

The inspectors

asked

the chemistry engineer

responsible for

PASS what alternate

means

had

been provided for post-accident

sampling of plant vent radioiodines

and particulates

during th=

period in ouestion.

The chemistry engineer

stated that,

base

on discussions

between operations

and chemistry personnel,

the

licensee

had decided that the portable

RADECO pump temporarily

installed at the

RE-24 sampling lines provided

an acceptable

means of sampling under post-accident

conditions.

Based

on

that decision,

no log entry had been

made regarding

inoperability after February

28.

The inspectors

noted that licensee

Surveillance Test Procedure

{STP) G-14, "Operability Determination of Post Accident

Sampling Program," listed

RE-24

as

an alternate

post-accident

sampling point for plant vent radioiodines'nd particulates.

RE-24 was also listed

as

an alternate

sampling point in

Equipment Control Guideline

(ECG) 11. 1,

"Post Accident Sampling

System."

The inspectors

noted,

however,

the following

'concerns:

{a)

The grab sample cartridge at the

RE-24 location

had not

been provided with any shielding.

In addition, the

RADECO

pump temporarily attached

to RE-24 sampling lines required

local operation.

Under post-accident

conditions, this

would require prolonged

exposure

in a location

approximately

2 3 feet. from the plant vent.

(b)

Accessing

the

RE-24 sampling location

on the 115'uel

'andling Building would require passing

the filter banks

on the 115'allway.

Under post-accident

conditions,

these filter banks would become highly radioactive,

creating

a considerable

radiological

source

term in the

hallway.

(c)

Time and motion studies

had been

performed to ensure

the

capability to obtain

and analyze

a sample

from RX-40 under

accident conditions;

however,.no

such studies

had

been

performed for sampling

from RE-24.

(d)

Despite listing RE-24

as

an acceptable

alternate for RX-40

in the above

procedures,

the licensee

had never determined

that the

RE-24 design

allowed collecting

a sample

under

accident conditions within the

GDC-19 dose criteria, nor

was

such

a determination

made during the February

26 -,

Harch

9 period to justify dependence

on RE-24.

At the exit interview on April 9,

1993,

the inspectors

asked

the licensee

to perform

a dose calculation to determine whether

samples

could have justifiably been obtained

under specified

accident conditions using the temporary

RADECO pump at the

RE-

24 location.

(5)

Licensee

Evaluation

On April 13,

1993,

the licensee

provided the inspectors

~:ith

czlcuTation of source

term and resultant

dose entitled

l5

"Extemporaneous

Use of RE-24's 'Old'ample Line for Post

Accident Sampling in Connection with CAP E-2

5.

EP RB-12 in the

Event

RX-40 Is Not Operable."

This study accounted for

exposure

from (I) the sampling cartridge,

(2) the filter bank

hallway,

(3) outside

area

exposure,

and (4) plant vent exposure

while obtaining the sample.

The study concluded that

an

individual obtaining

and analyzing

such

a sample would receive

approximately 0.48

rem whole body dose

and 4.8 rem extremity

dose,

without the use of shielded

sampling or transport

casks.

Based

on the

above study, -the licensee

concluded that the use

of this proposed

sampling location would have provided

an

acceptable

means of obtaining

and analyzing radioiodines

and

particulates

in plant gaseous

effluents under accident

conditions.

(6)

Further

NRC Evaluation

and Conclusion

The inspectors

reviewed the licensee's

study,

and noted the

following concerns:

(a)

The licensee's

study had not -used the applicable

source

term provided in NUREG-0737,Section II.F.I, Attachment

2

(as committed in the April 15,

1982, letter), for

evaluating the dose from the, radioiodine cartridge

and

particulate filter.

Using the applicable

source

term from NUREG-0737,

the

inspectors

recalculated

the whole body dose from handling

the sampling cartridge for 90 seconds

to be approximately

25 rem.

(b)

The licensee's

study had not used

the analysis

methods of

NUREG-0737,Section II.B.2 for evaluating accessibility of

the Plant Vent

Room on the 115'levation of the Fuel

Handling Building during post-accident

operations.

(c)

Emergency

Procedure

(EP)

RB-12,

"Hid and High Range Plant

Vent Radiation Honitors,." Appendix I, Section 4.b,

stated

the following:

Should radioactivity levels of plant vent

effluents rise to the operating level of the

RE-

29 monitor, it would be virtually impossible to

detect radioiodines

in the presence

of the noble

gases.

Because of the high radiation levels

from the plant vent itself, personal

exposures

from entering the monitor area to obtain grab

samples

would be prohibitive.

For this reason,

a 'separate

iodine grab sampler

(RX-40) has,been

installed.

16

(d)

The inspectors

concluded that the licensee's

analysis

was

inadequate,

and that the licensee's

conclusion regarding

the adequacy of this sampling location

was in error.

In

subsequent

discussions

held on April 29,

1993,

the

licensee

acknowledged

the inspectors'bservations.

The

licensee

stated that

an additional possible

sampling

location would have

been from the

new RE-24

and

RE-24R

monitors,

which had not yet been declared fully

operational

at the time in question.

The inspectors

concluded that,

from February

26 to Harch 9,

1993, the licensee

had failed to implement

and maintain

a

program which ensured

the capability to obtain

and analyze

samples of radioiodines

and particulates

in the plant gaseous.

effluents under accident conditions, constituting

an apparent

violation of TS 6.8.4

(50-275/93-11-04).

Regarding the overall safety consequence

of these deficiencies,

the inspectors

noted that the Plant Vent

Room on the

115'levation

of the Fuel Handling Building would be accessible

while meeting

GDC-19 dose criteria under

some accident

conditions,

and that the conditions provided in NUREG-0737

represented

worst-case

accident conditions.

The inspectors

also noted,

however, that the licensee's

PASS equipment control

guidelines and'urveillance test procedures

did not establish

the severity of the accident for which this "alternate"

sampling location could be used.

In addition, the inspectors

noted that'Emergency

Procedure

(EP)

RB-9, "Calculation of Release

Rate,"

proposed

using the RX-40

sampling data

as

one input to

EP RB-11,

"Emergency Off-Site

Dose Calculations."

The inspectors

noted, further, that these

calculations of release

rate

and resultant off-site dose could

be used in establishing

protective action recommendations

under

certain accident conditions.

c.

Additional Observations

Related to