ML20136J230

Insp Rept 50-293/85-27 on 850916-20.Deviation noted:20 to 30 Ft of Redundant Drywell Detector Cable Not Installed in Conduit

ML20136J230
Person / Time
Site:	Pilgrim
Issue date:	11/07/1985
From:	Cheung L, Nimitz R, Pasciak W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20136J205	List: ML20136J201 ML20136J225 ML20136J230
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-2.F.1, TASK-3.D.3.3, TASK-TM 50-293-85-27, NUDOCS 8511250309
Download: ML20136J230 (37)

See also: IR 05000293/1985027

Text

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report No. 50-293/85-27

Docket No. 50-293

License No. OpR-35 Priority --

Category C

Licensee: Boston Edison Company M/C Nuclear

800 B6ylston Street

Boston, Massachusetts 02199

Facility Name: Pilgrim Nuclear Power Station

Inspection At: Plymouth, Massachusetts

Inspection Conducted: September 16-20, 1985  !

Inspectors: SLd d

R. L. Nimitz, Sen%r Radiation Specialist

11/ 6 ! 6 5

date

2A.0%

L. '5. Ihueng, Reactor Engin er

n//

date

t'W

A. P. Hull, Brookhaven National Laboratory

W. H. Knox, Knox Consultants, (contractor to Brookhaven

boratory)

fatin /

Approved by: . (A.o k / A__

W. J .' /lasciak, Chief, BWR Radiation

_ // (7 f[ *

/ d/ti

Prot & tion Section ( (

InspectionSummary:(

Inspection on September 16-20, 1985 (Report No. 50-293/85-27)

Areas Inspected: Special, announced safety inspection of the licensee's

implementation and status of the following task action items identified in

NUREG-0737: II.B.3, Post Accident Sampling Capability; II.F.1-1, Noble Gas

Effluent Monitors; II.F.1-2, Sampling and Analyses of Plant Effluents;

II.F.1-3, Containment High-Range Radiation Monitor; III.D.3.3, Improved

Inplant Iodine Monitoring. The inspection involved 140 hours0.00162 days <br />0.0389 hours <br />2.314815e-4 weeks <br />5.327e-5 months <br /> onsite by two

region-based inspectors and two contractors from Brookhaven National

Laboratory.

Results: No violations were identified. Several areas requiring improvements

were identified.

0511250309 851113

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DETAILS

1.0. Persons Contacted

f

! The individuals contacted during this inspection are listed in

Attachment I to this inspection report.

l 2.0 Purpose

! The purpose of this inspection was to verify and validate the adequacy of

i the licensee's implementation of the following task actions identified in

i NUREG-0737, Clarification of TMI Action Plant Requirements:

l

l Task No. Title

l II.B.3 Post-Accident Sampling Capability

l II.F.1-1 Noble Gas Effluent Monitors

j II.F.1-2 Sampling and Analysis of Plant Effluents

i

II.F.1-3 Containment High-Range Radiation Monitor

III.D.3.3 Improved Inplant Iodine Instrumentation

under Accident Conditions

' In addition, and as part of the inspection, a review was performed to

verify and validate the adequacy of the licensee's design and quality

assurance program for the design and installation of the Post-Accident

!

Sampling System (PASS). The findings in this area are presented in

Section 9 of this report.

j 3.0 TMI Action Generic Criteria and Commitments

l The licensee's implementation of the task selection actions specified in

Section 2.0 were reviewed against criteria contained in the following

!

documents:

l * NUREG-0578, THI-2 Lessons Learned Task Force Status Report and

l Short-Term Recommendations, dated July 1979,

I *

Letter from Darrell G. Eisenhut, Acting Director, Division of

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Operating Reactors, NRC,'to all Operating Power Plants, dated

October 30, 1979,

,

NUREG-0737, Clarification of TMI Action Plan Requirements, dated

l Ovember, 1980,

l

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Generic Letter 82-05, Letter from Darrell G. Eisenhut, Director,

l

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Division of Licensing (DOL), NRC, to all Licensees of Operating

Power Reactors, dated March 14, 1982,

Pilgrim Nuclear Power Station, Unit 1, Updated Final Safety Analysis

Report, dated July 11, 1982,

l

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Letter From Darrell G. Eisenhut, Director, Division of Licensing,

NRR to Regional Administrators, " Proposed Guidelines for Calibration

and Surveillance Requirements for Equipment Provided to Meet Item

II.F.1, Attachments 1, 2, and 3 NUREG-0737," dated August 16, 1982,

Order confirming Licensee Commitments on Post-TMI Related Issues,

dated March 14, 1983,

• Modification of March 14, 1983 Order, dated June 15, 1984.

Regulatory Guide 1.3 " Assumptions Used for Evaluating Radiological

Consequences of a Loss of Coolant Accident for Boiling Water

Reactors,"

• Regulatory Guide 1.97, Rev. 2, " Instrumentation of Light-Water-Cooled

Nuclear Power Plants to Assess Plant and Environs Conditions During

and Following an Accident," and

Regulatory Guide 8.8, Rev. 3, "Information Relevant to Ensuring that

Occupational Radiation Exposure at Nuclear Power Station will be As

Low As Reasonably Achievable."

i

In addition specific review criteria and/or commitments relative to each

task item numbers are included in Attachments 2-7 of this report.

4.0- Post-Accident Sampling System, Item II.B.3.

4.1~ Position

NUREG-0737, Item II.B.3, specifies that licensees shall have the capabil-

ity to promptly collect, handle, and analyze post-accident samples which

are representative of conditions existing in the reactor coolant and con-

tainment atmosphere. Specific criteria are denoted in commitments to the

NRC relative to the specifications contained in NUREG-0737.

Documents Reviewed

The implementation, adequacy and status of the licensee's post-accident

sampling, monitoring, and analysis systems were reviewed against the

,

criteria identified in Section 3.0 and in regard to licensee letters,

'

memoranda, drawings and station procedures as listed in Attachment 2 of

this Inspection Report.

The licensee's performance relative to these criteria was determined from

! interviews with the principal personnel associated with post-accident

l sampling, reviews of associated procedures and documentation, and the

l conduct of a performance test to verify hardware, procedures and personnel

capabilities.

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4.2 Findings

Within the scope of the review, the following items items were

identified:

4.2.1 System Description and Capability

The licensee.has installed a Post Accident Sampling System which is

a standard General Electric design. It has the ability to obtain

unpressurized undiluted and diluted samples of reactor coolant from

the jet pump and the RHR System. Also, samples can be obtained

from the drywell, suppression pool and reactor building atmospheres.

Redundant containment hydrogen analyzers provide a hydrogen analysis

back-up capability.

Analysis for chloride, boron, and hydrogen _are conducted in the

laboratery, using an ion chromatograph, plasma spectrometer and gas

chromatograph, respectively. Back-up analysis capability is being

negotiated with other nearby utilities. .

4.2.2 PASS Performance Testing

Grab samples of reactor coolant and of the drywell atmosphere were

collected during an operational test on September 18, 1985. During

the test, licensee personnel verified the intergrated ability to

collect and analyze samples within the time constraints of

NUREG-0737, II.B.3. (See Attachment 3 of this report.)

4.2.3.1 Reactor Coolant (Findings)

The reactor coolant sampling system is designed to obtain samples of

liquids and dissolved gasses during all modes of operation. Although

samples could be obtained from all sampling points, the following

matters requiring licensee attention were identified:

a. Demonstrate the adequacy of system purge times. Data were not

presented to demonstrate the adequacy of the purge times speci-

fied in procedures. Adequate purge time is needed to ensure

representatives sampling. (50-392/85-27-01)

b. Demonstrate the adequacy of the sample dilution method and

related equipment. Data were not available to demonstrate the

adequacy of the sample dilution method and related equipment.

Sampling dilution is a key element in the quantification of

sample results. (50-293/85-27-02) ,

c. Determine the volume of the coolant collection ball valve.

During the preoperational testing, the ball valve, wh!,ch collects

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a measured volume of coolant, was determined to be 0.14 ml,

instead of the value of 0.10 ml. The valve of 0.10 ml. was

presented in procedures. The valve, which was actually_ tested

during preops, has since been replaced, but the new valve's

volume has not been determined. (50-293/85-27-03)

d. -Repair and/or replace the flow control check valve. After the

primary system tests had been successfully conducted, the flow

control valve stuck open in a fixed position (0.6 GPM). The

design flow rate of 1 GPM could not be attained. (50-293/

85-27-04)

4.2.3.2 Containment Air (Findings)

Atmosphere samples can be obtained from the Drywell, Reactor Build-

ing and Suppression Pool.

The following matters requiring licensee attention were identified:

a. Demonstrate the adequacy of system purge times. Data were not

presented to demonstrate the adequacy of the purge times speci-

fied in-procedures. Adequate system purge time is needed to

ensure representative sampling. (50-293/85-27-05)

b. The capability to obtain a representative sample of containment

atmosphere should be demonstrated. (50-293/85-27-06):

1. There has been no line loss or plate-out study conducted.

2. The sa;npling assembly is not heat traced. This could lead

to excessive condensation collecting on the iodine car-

tridges. This problem would be particularly troublesome

with high humidity in containment.

c. Correct the air sampler rotometer reading for differences in air

density created by pump suction. The sampling procedure

(5.7.11) uses the rotometer reading in the calculation of the

radioiodine concentration. (50-293/85-27-07)

4.2.4 Analytical Capability (Findings)

Attachment A to BEC0's May 30, 1985 letter contains the licensee's

commitment relative to the range, sensitivity and type of analytical

capability available.

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4.2.4.'1. Chloride (Findings)

Chloride analysis of diluted coolant is conducted using an ion

chromatograph. The lower range sensitivity commitment is 0.01 ppm.

However, the water used for dilution has a higher chloride

1 concentration (approximately 0.03. ppm).

In order.to detect 0.01 ppm, analysis would have to be conducted

using an undiluted sample. The licensee's May 30, 1985 letter in-

dicated'that'an undiluted sample would be collected and retained for

up to 30 days for subsequent confirmatory analysis. However pro-

cedures did not provide for the collection and retention of an un-

diluted sample for further chloride analysis. The accuracy of +/-10%

at the 0.01 ppm level could not be achieved. (See Attachment 3 for

test results).

The following matters requiring licensee attention was identified:

• The detection limit and sensitivity of the chloride analysis

method should be clearly defined. Procedures'should be revised

to include provisions for the retention of an undiluted sample

for up. to 30 days for more detailed chloride analysis. (50-293/

85-27-08)

u4.2.4.2 Boron (Findings)

Boron analysis is conducted using two methods: Plasma Spectrometry

and Carminic Acid (Spectrometry). The analysis of spiked samples

were conducted using the primary method (Plasma Spectrometry) and

acceptable results were obtained.

Reagents needed to conduct back-up analysis using the Carminic Acid

method were not available. Hewever, they were on order.

The following matter requiring licensee attention was identified:

• Obtain and maintain supplies necessary to conduct boron

' analysis with carminic acid method. (50-293/85-27-09)

.4.2.4.3 pH Analyses (Findings)

Although the licensee has purchased the equipment for performing pH

analyses,.his procedures for this purpose were not established. The

licensee personnel indicated that this analysis.is unnecessary.

The following matter requiring licensee attention was identified:

• Clarify commitments and capabilities relative to pH analyses. I

(50-293/85-27-10)

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4.2.4.4 Gross Gamma and Isotopic Analyses (Findings)

An isotopic analysis of the PASS grab sample was compared to the

normal sink sample. The comparison was within acceptable limits.

Provisions had not been made to conduct an isotopic analysis on the

depressurized dissolved gasses in core damage procedures.

The following matter requiring licensee attention was identified:

Review and evaluate the need to perform isotopic analyses of

dissolved gasses. If a need is identified, provision should be

made to conduct isotopic analyses on dissolved gasses. (50-293/

85-27-11)

4.2.4.5 Hydrogen and Dissolved Gas (Findings)

Licensee personnel differed in their views on the type of analysis

to be conducted and how the data would be used. The core damage

assessment procedure uses the hydrogen content to assess core

damage. Procedures are available for the analysis of hydrogen,

using an ion chromatograph.

Licensee personnel indicated that the primary analysis method to

quantify hydrogen was the use of total dissolved gas. However, the

procedure for this purpose had not been completed. Additionally

licensee personnel indicated that the total gas, once determined, was

to be equated to hydrogen and used as such in core damage assessment.

Others indicated that the total gas data would be obtained but would

not be used. It was not clear that the total gas obtained was equi-

valent to total hydrogen. Equating hydrogen to total gas could lead-

to an overly conservative estimate of the hydrogen content.

Additionally, tests have not been conducted to demonstrate that the

committed range and accuracy of either dissolved ges or hydrogen

could be achieved.

The hydrogen analysis procedure required that 0.25 cc of gas be

injected into the gas chromatograph. During the test, approximately

5 cc was collected and transported to the laboratory. The bringing

of excessive amounts of gas into the laboratory could unnecessarily

increase radiation exposure.

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Although the procedure directed the operator to inject 0.25 cc into

the gas chromatograph, a method had not been developed to extract

this volume from the large syringes in which the gas is transported.

The operator indicated that an undetermined volume, up to 5 cc

would be injected directly from it. Since the calibration gas

volume is 0.25 cc, the GC will respond differently if a 5 cc volume

is used.

The following matters requiring licensee attention were identified:

(50-293/85-27-12):

a. The use of dissolved gas or hydrogen analysis data in the as-

sessment of core damage should be clearly specified. Also, it

should be demonstrated that the committed range and accuracy can

be achieved. Procedures should be established and implemented

where needed.

b. The amount of gas transported to the laboratory for analysis

should be minimized.

c. Personnel should be instructed either to follow procedures or

the procedures should be revised to reflect actual practices.

Incorrect gas volumes were injected into the gas chromatograph.

4.2.5 Additional Findings

The following additional findings were identified. These matters

require _ licensee attention:

a. Establish a routine maintenance program for the PASS. Not all

components of the PASS have been included in a regular surveill-

ance/ calibration program. Also, there does not appear to be a

formal administrative procedure for assuring that non-safety

related equipment, such as the PASS, is incorporated in the

routine maintenance (50-293/85-27-13)

b. Establish a spare parts program for the PASS.

A PASS spare parts program is under development. The licensee

is coordinating this program development with the BWR Owner's

Group. (50-293/85-27-14).

c. Establish PASS sample shipping procedures.

The procedures for the handling, loading and off-site shipment

of samples are in a draft form and arrangements are being made

through the PIMS to supply the needed shipping cask.

(50-293/85-27-15)

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d. Complete arrangement for backup off-site analyses of samples.

Arrangements for backup off-site analysis are incomplete. A

verbal agreement exists with Millstone to provide backup sup-

port. This agreement is being formalized by the licensee's

legal staff. (50-293/85-27-16)

e. Establish provisions for detection of leakage across chiller.

Provisions have not been made for the detection of leakage of

reactor coolant into the chiller cooling water. The pressure

differential between the sampling and the cooling lines could

result in the buildup of significant activity in the chiller in

the event of such a leak. (50-293/85-27-17)

f. Provide provisions to preclude PASS from exceeding its design

temperature specification.

The chiller will automatically shutdown on low or high pressure

signals. However, mechanisms have not been provided for the

interruption of the flow of the uncooled water into the PASS or

of operator actions procedures to prevent the PASS from exceed-

ing its temperature design specifications. (50-293/85-27-18)

g. Improve the safety of the access to the PASS.

Under accident conditions personnel may be required to climb a

20' ladder to the location of the PASS controls while wearing

protective clothing and SCBA gear. The ladder does not have a

cage to prevent a person from falling backward. (50-293/

85-27-19)

h. Establish Alarm Set Points for the PASS Area Radiation Monitor.

Consider using an alarm set point based on maximum dose rates

that could be encountered such that 10 CFR 50 GDC 19 dose

limits will not be exceeded. (50-293/85-27-20)

1. Evaluate the need to install charcoal filters in the exhaust

hood of the chemistry laboratory.

A charcoal filter has not been provided for the laboratory

ventilation exhaust system. During an accident radioiodine

releases may result from the operation of the gas chromatograph,

plasma spectrometer and laboratory hood. (50-293/85-27-21)

j. Calibrate the PASS radiation detector in accordance with

manufacturer's specifications.

The three radiation detectors associatsa with the PASS are not

calibrated in accordance with manufacturer's specifications and

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recommendations. For example, the manufacturer requires the

one electronic channel to be set at 2200 volts and the other

two channels at 2500 volts. Based on the calibration procedure,

all electronic channels were set at 2500 volts. Also, there was

no in-situ check of the instruments response following rein-

stallation. (50-293/85-27-22)

k. Reevaluate the adequacy of the " time and motion" study

performed for collection of PASS samples. Ensure the re-

quirements of 10 CFR 50 GDC 19 can be met.

The " Time and Motion Sedy" was conducted using generic methods

before the actual procedves were developed. It is not clear

that the samples can be co'lected and analyzed within

GDC limits using the existing procedures. (50-293/85-27-23).

1. Clarify use of pressure in procedures.

Procedure 5.7.4.1.9 does not include provisions for the speci-

fication of the pressure in units of inches of Hg. Procedures

use units of psig. Subatmospheric readouts in " inches of Hg"

are the most appropriate readouts. (50-293/85-27-24)

m. Establish reliable backup power for the chiller.

l

A reliable source of backup power has not been provided for the

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chiller used for cooling the incoming reactor coolant in the

!

event of a loss of off-site power. While a backup method of {

l cooling has been devised, it has not been tested for proper

i

hose fitting. Also, the heat removal capability of this method

nas not been established. (50-293/85-27-25)

n. Identify and provide " carrying devices".

l

Procedure 5.7.3.1.2 states that a " carrying device" would be

used to transport the syringe containing radioactive gasses.

It is not clear what type of " carrying device" is to be used.  ;

, (50-293/85-27-26) '

4.2.6 Item for Improvement

The following improvement item was identified:

Consider / identify methods to ensure adherence to PASS procedures.

One operator was solely responsible for the operation of the system

and the collection of the sample. A mechanism for verifying that the

correct procedure steps had been properly followed or assuring that

the correct information had been recorded has not been provided.

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5.0 Noble Gas Effluent Monitor, Item II.F.1-1

5.1 Position

NUREG-0737, Item II.F.1-1 requires the installation of noble gas

monitors with an extended range designed to function during normal

operating and accident conditions. The criteria, including the de-

sign monitors for individual release pathways, power supply, cali-

bration and other design considerations are set forth in Table II.

F.1-1 of NUREG-0737.

Documents Reviewed

The implementation, adequacy, and status of the licensee's monitoring

systems were reviewed against the criteria identified in Section 3.0

and in regard to licensee's letters, memoranda, drawings and station

procedures as listed in Attachment 4 of this Inspection Report.

The licensee's performance relative to these criteria was determined

by interviews with the principal per sons and consultants associated

with the design, testing, installation and surveillance of the high

range gas monitoring systems, a review of the associated procedures

and documentation, an examination of personnel qualifications and

direct observation of the systems.

5.2 Findings

Within the scope of this review, the following was identified: .

5.2.1 Description and Capability

There are three atmospheric release locations at Pilgrim Nuclear

Power Station (PNPS), the free-standing main stack (to which the

effluent from the SBGTS is ducted), the reactor building exhaust

stack and the turbine building roof exhaust.

The licensee has i. stalled high re.nge ion chambers to supplement the

pre-existing normal range monitors for the main stack and for the

reactor building vent. An ion chamber has also been installed to

provide high range monitoring for the turbine building roof vent,

which is not normally monitored.

5.2.1.1 Normal Range Description

The normal range monitors for the main stack and the reactor

building vent consist of identical GE designed shielded gamma

sensitive Na! detectors which view a volume of gas in a shielded

chamber. Each has a seven decade range. From calculations supplied

by the licensee's consultant, ENTECH, the sensitivity of the main

stack monitor appears to be 4.4 x 10 -7 -4.4 uCi/cm3 and that for the

reactor building vent appears to be 6 x 10 ~0 - 0.6 uC1/cm3 .

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The turbine building is not normally monitored by installed instru-

mentation.

5.2.1.1 High Range Description

General Description

General Atomic Model RD-2A fon chambers, with a range of 10 -I -10 4

R/hr are employed for high range monitoring. For the main stack one

is installed externally to a 20" horizontal duct at its base, while

those for the reactor and turbine buildings are externally mounted

at elevated levels on the vertical ducts near the roof vents.

High Range (Main Stack) i

The licensee's submittal of February 27, 1981 indicated a range for

the stack monitor of 10 1

- 106 uti/cm 3equivalent 133

Xe concentra- I

tion. A subsequent licensee submittal of June 4,1984 indicated an

upper limit of 2 x 105 uC1/cm 3 133 Xe equivalent. The basis for these

l

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submittals could not be established at the time of the inspection.

'

From calculations supplied onsite by ENTECH, the apparent range of

l the ion chamber monitor for the high range main stack is 60 - 6 x

6 133

10 for Xe, in which case overlap with the normal range monitor

'

would not be provided. However, the attenuation of the 3/8" thick

duct well is embodied in ENTECH's calculations. When this is taken

into account, the effective range of this monitor appears to be

4

0.7 - 7 x 10 uC1/cm3 on a 133 Xe equivalent basis.

ENTECH's calculations showed that in the event of a design basis

accident, there would be an initial overlap for the immediate post-

accident mixture. However, the overlap would narrow with elapsed

time post-accident, as the average energy of the mixture decreased.

For a design basis accident, it would be very narrow before the upper

range of the low range monitor was reached some 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> later.

High Range (Reactor Butiding}

In its submittal of February 24, 1981, the licensee indicated a range

for the High Range monitor for the reactor building vent of 8 x 10 -2

3 3

to 8 x 10 uC1/cm while that of June 14, 1982 indicated an upper

4

limit of 2 x 10 uC1/cm ,3 133Xe equivalent.

From ENTECH's calculations reviewed onsite, the range of this monitor

5

appears to be 1 - 10 uC1/cm3 of 133 Xe. This monitor also showed a

range overlap for the immediate mixture which also narrowed with time

post-accident.

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High Range (Turbine Vent) ,.,

The licensee's submittal of February 17,\ 1981 indicated that the

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apparent range of the turbine vent monit'or was 1.5 x 10-2 3,5 x *

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103 uCi/cm 3

, which agreed with the upper limit indicated in the sub- - ,'

mittal of June 14, 1982. From ENTECH's calculations, its apparent >

range was approximately 3 x 10 -2 -3 x 103 uCi/cm3 of 133

Xe. The

upper limit meets the criterion of the NURCG-0737, II.F.1 Attachment -

1. Since NRR has accepted the licensee's contention that low range -

monitors are not appropriate for the turbine roof vents, as they are '

,:

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not normal release points, the question of range overlap is not ap- ,

plicable.

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5.2.1.3 General Findings (High Range Monitor)

'

The possible effect on monitor indications of the deposition of post

accident radiotodines on duct walls has not been considered in

ENTECH's calculations. The initial type calioration of the RD-2A ^

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ion chambers by the vendor (General Atomics) remains in open item

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from Inspection 50-293/83-02. However, data were supplied to dem- -

onstrate that the installed chambers are regularly calibrated

against a solid source throughout their range. Although.ENTECH's

code has not been verified with gaseous sources, they supplied a

report to demonstrate that the one used to establish the radiation

levels at the installed location of the ion chambers had been bench-

marked against an ANSI approved code.

Due to the limited range overlap between the low and high range ._

monitors for the main stack and the reactor building vent, the

ability of the former to function throughout their stipulated ranges ,

and beyond for extended period of time (up to a few days) and to re- ,

cover therefrom is an important element of the licensee's ability to  ;

follow a post-accident release. This ability has not been

documented.

Local readout and remote readout and recording in the Control Room

of the indications of the high range monitors in units of R/nr is 'b

provided. The interpretations of these indications in terms of re-

lease rates and/or concentrations can be niade by means of nomograms '

or a computer program.

Initial operator responses for unusual rele'ases are based on in- ,

dications of the low range monitors, for which appropriate alarm '

levels have been established. There are no alarms on the high range s

channels, nor are there specific provisions in emergency procedures

whereby the operator is directed to consult them. *

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The licensees indicated that an ion chamber was normally maintained

in its spare parts inventory. However, it had been utilized as a

replacement following a recent lightning induced failure at the main

stack. Another spare was on order.

5.2.2 High Range Monitors (Findings)

The installed system meets the guidance for high range noble gas

monitoring as contained in NUREG-0727, II.F.1, Attachment 1. The

following matters requiring license attention were identified:

a. Clearly specify the range of all High Range Noble Gas Monitors.

The reasons for the inconsistencies between the range of capa-

bility of the high-range monitors as can be derived from the

reports by ENTECH and from information supplied by the licensee

to NRC should be investigated. Also range overlap should be

clearly specified. (50-293/85-27-27)

b. The ability of the low range monitors to function for sustained

periods of time at concentrations close to and beyond their

upper range for and to recover therefrom during a post-accident

sequence should be established. If this cannot be satisfac-

torily acccomplished, provisicns for turning off the power to

thc,and/or bypassing them during these periods of time should

be c.onsidered. (50-293/85-27-28)

c. The possible effect of radioiodines deposited within ducts on

the response of the high range monitors should be considered.

If it is appreciable, considerations should be given to relo-

cating the detectors to a shield cave within which they view a

suitable volume of an off-line aliquot of the stack / vent flow.

(50-293/85-27-29).

6.0 Sampling and Analyses of Plant Effluents, Item II.F.1-2

6.1 Position ,

NUREG-0737, Item II.F.2, requires the provision of a capability for the

collection, transport, and measurement of representative samples of radio-

active iodines and particulates that may accompany gaseous effluents

following an accident. It must be performable without exceeding

specified dose limits to the individuals involved. The criteria

including the design basis shielding envelope, sampling media, sampling

consideration, and analysis considerations are set forth in Table

II.F.1-2.

.

.

15

,

I

'

6.2.1' Description and Capability

The licensee has elected to utilize the pre-existing normal sampling

arrangements for compliance with NUREG-0737, II.F.1. Attachment 2.

As such, they are integral wfth the low-range gas monitors which are

routinely employed for the sampling of iodines and particulates in

the effluent from the main stack and the reactor building vent.

.The probe for the main stack is installed at an elevation of about

25', at which point it samples the combined flow from the SBGTS duct

of 4,000 cfm and of stack dilution air at a rate of 16,000 cfm. The

sampling line is heat traced and has a relatively-short run to the

sampling station, which is located at the base of the main stack.

The probe for the reactor building vent is installed at the 163'

level near the top of this 9' x 9' exhaust duct. The licensee's-

calculations indicate that this probe is sized to be isokinetic for a

stack flow rate of 65,000 cfm. No correction features were available

for other flow rates, such as the much lower ones which would occur

during accident conditions. The sample line has over a 100' vertical

run, which is followed by a relatively short horizontal run to the

sampling and monitoring station, which is located on the turbine deck

adjacent to the stack. This sample line is not heat traced.

For routine sampling, a particulate filter and charcoal base iodine

collection canister are installed in'a filter holder which is mounted

with quick disconnect fittings on the sample station rack ahead of

'

the low range gas monitor. This filter holder is unshielded. The

emergency procedure calls for.the replacement of the in place can-

ister with one containing a silver zeolite collection medium at the

onset of accident-conditions.

The licensee's procedures envision the use of a nomogram to estimate-

the 131 I activity released to the environment, based on gamma dose

rates of the cartridge. However, this is not specifically called for

in the sampling procedures for the main stack and the reactor build-

ing vent. Also the nomograms which are contained in the procedures

permit only an estimation of the activity on the filter. Further-

more, as written, these procedures call for the setting aside of

sample holders with contact dose-rates that are above the limit for

" hand-touching". The sample holders are placed in a shielded cave

adjacent to the sampling station, prior to any other measurement of

them.

These procedures also call for a purge step prior to the removal of

the filter holder from the sample rack. During the inspection, it

was found that the purge enters the sample line subsequent to the

holder. As a result, the purge only purges the gas monitor chambers.

The sample racks also contain a number of unlabeled valves.

.

.

16

The procedure for sampling the effluent from the turbine building for.

iodines and particulates under the accident conditions calls for the

use of a portable air sample pump and filter holder to collect a 10

minute sample at a pre-established location on the turbine deck lev-

el. It also calls for the measurement of the gamma dose rate of the

filter with a survey instrument and the estimation of the activity on

the filter, using the same nomogram that is contained in the stack

and reactor building vent emergency sampling procedures.

-The licensee was unable to furnish any calculations that at the

shielding design basis, plant personnel could remove samples, replace

the sampling media and transport samples to the on-site analysis

facility with radiation exposures that would not exceed the GDC-19

criteria.

A review of the licensee's manual dose assessment procedures indi-

cated that they provided only for the use of assumed iodine to noble

gas ratios and did not provide for the use of the measured radio-

fudine activity on effluent sampling media for the estimation of

field iodine dose rates.

6.2.2 Sampling Plant Effluents (Findings)

The following matters requiring licensee attention were identified:

! a. Establish provisions in procedures for the sampling of the main

stack and the reactor building vent (similar to those now con-

tained in these for sampling in the turbine building) for the

measurement of samples of radiation levels up to and including

design basis, so as to accomplish continuous sampling throughout

a post-accident sequence. (50-293/85-27-30)

b. Perform a time and motion study to ascertain that the system

design will make it possible to remove and transport design

basis samples within GDC-19 criteria. All appropriate source

terms should be used for this study. (50-293/85-27-31)

c. Provide the necessary nomograms and calculator / computer pro-

cedures, whereby measurement of the collected radioiodine

activities can readily be translated into release concentra-

tions and rates. (50-293/85-27-32)

d. Develop correction factors for the non-isokinetic sampling rates

for the range of stack flow rates of the unit vent especially

for those anticipated under accident conditions. (50-293/

85-27-33)

. - - - - ~ . , .

.

~

-

17

. e. -Evaluate the capabilities of the sampling system to collect

'

representative samples under accident condition. (50-293/

85-27-34)

.

6.2.3 - Additional Items for Review / Resolution

In order to reduce the anticipated dose rate of the collected sample,

and/or to facilitate its analysis, the licensee should review and/or

resolve (as appropriate) the following matters: (50-293/85-27-35)

f a. Provide shielding for the effluent sample holders.

' b. The provision of' labels and a flow diagram for the valves and

indicators on the sample racks.

c. The heat tracing of the sampling line for reactor building vent,

.

'in view of the possibility that it may contain steam leakage or ,

- moisture therefrom under accident conditions.

'

d. The provision of featJres which will- enable the purging of in-

place sample canister and nearby sample' lines with an clean air

supply, prior.to their removal for transport and analysis,

e

e. The addition to manual dose assessment-procedures of nomograms

which would make it possible to estimate field iodine dose rates

on-the basis of measured radiciodine activity (or release rates

derived therefrom).

7.0 Containment High-Range Radiation Monitor, Item II.F.1-3

7.1 Position .

NUREG-0737, Item II,F.1-3, requires the installation of two in-con-

tainment radiation monitors with a maximum range of~1 rad /hr to 10'

,- rad /hr (beta and gamma) or alternatively l' R/hr to 107 R/hr(gamma

,

only). The monitors shall be physically separated to view a large

portion of~ containment and developed and qualified to function in an

accident environment. :The monitors are also required to-have an

energy response as specified in NUREG-0737,~ Table II.F.1-3.

~

-

Review Criteria

.

L .The implementation adequacy, and status of the installed in-contain-

ment high range' monitors were reviewed against the criteria set

.forth in Section 3.0 of this report and in regard to interviews with

cognizant licensee personnel, licensee letters, station procedures,

as-built prints and drawings as listed in Attachment 5 to this

, Inspection Report.

.

4

,w- e r e r~~--.e ,<-m , , - - , ,- ~-wr .,e, ,r~r- -

e ,, --ww, , . - - -w,n-- wr

.

.

18

The licensee's performance relative to these criteria was determined

by:

• Interviews with cognizant personnel;

Review of applicable operational and emergency plan procedures;

Review of applicable lesson plans and training records;

• Review of calibration data; and

• Direct observation of installed equipment.

7.2 Findings

Within the scope of this review, the following was identified:

Installation / Placement

The licensee has installed four- (4) gamma sensitive ion chambers for

monitoring of gamma radiation emanating from primary containment.

Two of the detectors monitor the torus at the North and West areas

of the Torus respectively (-17' elevation). The remaining two de-

tectors are mounted in close proximity to each other (about 2 feet

apart) at approximately the north mid plane area of the Drywell.

These latter two detectors project into primary containment through

two caped penetrations. The location and number of the detectors was

verified. The number and location of these detectors was found

acceptable to the NRC.

The detectors read out at the Post-Accident Monitoring (PAM) Panel,

located in.the Main Control Room. The read outs were operating

properly.

Procedures

The licensee has established procedures which included monitor

response curves. An estimate of core damage may be made by:

1. obtaining a reading from a detector;

2. determining time after shutdown; and

3. usirig an appropriate curve to estimate core damage. Different

curves are used for the Torus and Drywell detectors.

,

e

.

..

19

.The monitor response curves incorporate corrections for inherent

shielding of the steel penetration liner in which the detectors are

inserted.

Procedures are in place for calibration and periodic verification of

operability of the detectors.

Environmental / Seismic Qualification

See section 9 of this report.

Training / Qualification of Personnel

The licensee has provided training and qualifications of both oper-

ations and radiological controls personnel in use/ interpretation of

detector readings. Appropriate personnel are provided periodic re-

fresher training in use/ interpretation of the read outs.

Calibration-Range

The detectors were found to be calibrated in accordance with

NUREG-0737. The range of the detectors is consistent with

NUREG-0737 Table II.F.1-3.

7.3 Acceptability

The installed system can be considered to meet the guidance specified

in NUREG-0737, Attachment II.F.1-3. However, the following matters

requiring licensee attention were identified:

a. Fully establish and implement the operator Training Modules for

operation /use of the PAM Panel (Module CT-SH/IG-S-120). l

(50-293/85-27-36).

b. Train and qualify appropriate personnel on Procedure 5.7.5,

" Estimating Core Damage". Personnel have not been trained in

the new procedure. (50-293/85-27-37)

c. Include North / East Torus Monitor Response Curve in Procedure

5.7.5. The procedure contains only the Drywell monitor response

curve. However, the Drywell response curve is being incorrectly

used with the Torus monitors (50-293/85-27-38)

d. Review dose / damage response curves for the Dryweli Monitors.

At less than 100% core damage, no other radiation ';curce (e.g.

primary lines in area of detectors) are used as contributors to

the detector readings. A review should be performed to ensure

that sources in the area of the detectors (other than the gas-

eous activity in primary containment) do not adversely effect

the core damage estimates. (50-293/85-27-39)

i

.

.

20

8.0 Improved In-Plant Iodine Instrumentation Under Accident Conditions,

Item III.D.3.3

8.1 Position

NUREG-0737, Item III.D.3.3, requires that each licensee shall

-provide equipment and associated training and procedures for

accurately determining the airborne iodine concentration in areas

within the facility where plant personnel may be present during an

accident.

Review Criteria

The implementation, adequacy and status of the licensee's in plant

iodine monitoring under accident conditions were reviewed against

the criteria in Section 3.0 of this report and in regard to the

documents identified in Attachment 6 to this Inspection Report. The

licensee's performance relative to these criteria was determined by:

• Interviews with cognizant licensee personnel;

• Review of applicable operational and emergency plan procedures;

• Review of applicable lesson plans and training records;

• Direct observation of performance during a walkthrough; and

• Verification of equipment availability and storage.

8.2 Findings

Within the scope of this review, the following was identified:

Equipment

The licensee was found to have appropriate equipment for sampling

and quantifying airborne iodine in areas within the facility where

plant personnel may be present during an accident.

Procedures

The licensee has established procedures for calibration / operation of

the equipment used to determine concentration of airborne iodine.

Training

The licensee provides initial training and refresher training to

personnel responsible for calibration / operation of airborne iodine

collection and analysis equipment.

.

.

21

8.3 Acceptability .

The licensee's program to sample and analyze iodine against a back-

ground of noble gases can be considered to meet the guidance spec-

ified in NUREG-0737, section III.D.3.3. However, the following

matters requiring licensee attention were identified:

a. The licensee should obtain additional SAM-2s. Currently all

SAM-2s (4) have been assigned to specific locations. The

licensee does not have any spare units. Spare units should be

obtained in the event an assigned SAM-2 becomes defective or

needs calibration. (50-293/85-27-40)

b. Certain aspects of the procedures for calibration and use of

the SAM-2 are written in such a manner that literal reading of

the procedures would cause errors in quantification of airborne

iodine: (50-293/85-27-41)

.

--

the method for determination and use of iodine measurement

efficiency is not clear / consistent between SAM-2 calibra-

tion /use procedures,

--

' CPM' is not defined (i.e. net or gross)

c. Some of the procedures for use of the SAM-2 do not include the

correction factors for determination of total iodine dose. The

procedures only include correction factors for dose due to

I-131. (50-293/85-27-42)

d. No verification of acceptability of air flow calibration

devices used to calibrate iodine air samplers has been

performed (50-293/85-27-43)

e. Procedures for use of the SAM-2s for analysis of iodine

activity collected on sample media allow the analyses to be

made in up to a radiation field of 4 mR/hr. The licensee

should perform and document an evaluation that demonstrates that a

SAM-2 can detect 1x10 7uci/ml in a 4 mR/hr field. (50-293/

85-27-44)

9.0 Quality Assurance (QA) and Design Review

9.1 General Design and Installation Records

The following design records were reviewed:

The design and installation records reviewed in this area are

presented in Attachment 7 (Section A) to this report.

___

.

.

22

The inspector found that there were sufficient sample points to

obtain samples after the accident and that the installation

documents were readily retrievable. No unacceptable conditions

were identified.

9.2 Alternate Power Supplies to' PASS

The Pilgrim SER on PASS, Section 2.0, criterion 1 indicated that

during loss of off-site power, alternate power sources were avail-

able for both gas and liquid sampling systems that could be

energized in sufficient time to meet the three hour sampling and

analysis time limit.

The inspector reviewed pertinent documents concerning the power

supplies to the PASS and its associated electrical equipment, (e.g.,

sample cooling water pumps, valves that require operation to draw

samples) to ascertain whether all power supplies can be switched to

the diesel generator after a loss of off-site power.

Items examine included:

BECO Drawing S-E-155 " Station Electrical Single Line Composite

Diagram" Rev. 2 dated November 18, 1984.

Bechtel Drawing No. E8 " Single Line Meter & Relay Diagram 4160V

Breaker A409, 480V load Center 88 480 Volt". Revision E5.

BECO Drawing M239 " Post Accident Sampling & H and 0 Analyzer

2 2

System" Revision El dated September 13, 1985.

The inspector verified that the power supplies to the PASS could be

switched to the diesel generator after a loss of off-site power. A

self contained cooling unit was used to cool the sample. The unit

would be inoperable after loss of off-site power. Should this

happen after an accident, water from Fire Main System can be

supplied to the sample cooler by manually open two hand valves (one

for supply, one for discharge to sump). The water from Fire Main

System does have sufficient head to perform this function.

No unacceptable conditions were identified.

9.3 Environmental Qualification of PASS Valves

The Pilgrim SER on PASS, section 2.0, Criterion 3 indicated that the

PASS valves which were not accessible after an accident were environ-

mentally qualified for the conditions in which they need to operate.

_

.

23

The PASS solenoid valves required to be operable for post accident

sample collection are SV-5065-63 through SV-5065-86. (24 valves

total)

The inspector randomly selected the EQ files of 4 valves (SV-5065-63

and 64, SV-5065-69 and 70) for review. The documents reviewed in-

cluded the Equipment Qualification Evaluation Sheet (EQES) for each

of the 4 valves, all dated June 11, 1985, and Wyle's Qualificattun

Report 47066-50V-8, Revision B dated June 21, 1985.

Within the scope of this review, the following matters requiring

licensee attention were identified:

a. The model of the installed solenoid valves differed from the test

model. The licensee tried to qualify the installed model by

similarity. However, evaluation of similarities and differences

between the installed model and the test model (e.g. organic

materials used and their thermal aging effect, coil temperature

rating, physical size and construction etc.) were not performed.

(50-293/85-27-45)

b. The valve manufacturer recommended the 0-rings be replaced once

every five years. This was not addressed in the EQ file, Jus-

tification should be provided if this recommendation is not to

be implemented. (50-293/85-27-46)

c. Information Notice 84-68 identified field cable degradation

when connected to high power solenoid valves. This cable degra-

dation was caused by substantial temperature increase in the

solenoid housing. The effect of this should be addressed in

the EQ file. (50-293/85-27-47)

9.4 Physical Observation of System Installation

The inspector physically observed the installed PASS and discussed

its operation with the licensee's I&C personnel. The licensee

stated that the instruments on the PASS panel were calibrated during

the preoperational test early this year. Although there were no

evidences that the calibrations of the PASS instruments were

overdue, there were no calibration schedules set up for these in-

struments.

Within the scope of this review, the following matter requiring

licensee attention was identified:

Review and evaluate the need to incorporate the periodic cal-

ibration of PASS instrumentation into the routine maintenance /

calibration program. Critical instruments should be included

in such a program. (50-293/85-27-48)

_

.

.

24

9.5 Containment'High Range Radiation Monitor (CHRRM)

9.5.1 EQ of CdRRM Detectors

.The inspector reviewed the EQ files of the CHRRM detectors to

ascertain whether the files contained sufficient evidence that

these detectors were qualified for the environmental conditions

in which they need to operate after an accident.

There were four CHRRM detectors in Pilgrim, two for Drywell

(RE1001-606A&B), and two for Suppression Pool (RE1001-607A&B).

These detectors were manufactured by General Electric Company.

RE1001-606A&B were mounted in capped pipe sections extending

into the drywell atmosphere from drywell outside wall. RE1001-

607A&B were located in the secondary containment adjacent to

the suppression pool.

The following documents were reviewed:

• BECO Equipment Qualification Evaluation Sheets for RE1001-

606A&B, RE1001-607A&B.

• Advanced Systems Engineering Memo. Qualification of Gamma

Sensitive Ionization Chambers for Cams Post-Accident En-

vironmental Conditions, Report No. 943-81-003, April 24,

1981.

• Environmental Qualification Test Report of Raychem WCSF-N

' Nuclear In-Line Cable Splice Assemblies for Raychem Cor-

poration Menlo Park, California. Report No. 58442-1,

May 15, 1980.

The inspector physically observed the installation conditions

of these four CHRRM detectors.

The following matters requiring licensee attention were

identified:

a. The radiation exposure qualification data was not

contained in the EQ files. This data should be made

available for NRC review (50-293/85-27-49).

b. The coaxial cables for the radiation detectors at the

Drywell were found laying on the floor and were subject to

being stepped on. The licensee's May 18, 1982 letter in-

dicated the cables were-in conduit. This appears to be a

,'

Deviation from the information provided to the NRC.

(50-293/85-27-50)

. _. . . . _

_ __ _

. _ .__ _ ._ _.

. _ _ _ _ _ _ _ - -

.

.

25

9.5.2 Power Supply to CHRRM Detectors

The inspector reviewed pertinent documents to ascertain that

safety related redundant emergency power supplies were used in

these four detectors.

The documents reviewed are identified in Attachment 7

(Section B) of this report.

The inspector verified that these four radiation detectors were

powered by safety related redundant power supplies channels A &

B.

1

Within scope of this review no unacceptable conditions were f

identified. j

)

9.6 Environmental Qualification of Containment Hydrogen Monitors

(Analyzers)

The inspector reviewed the EQ files of Containment Hydrogen Monitors

(C172 and C173) to ascertain whether the file contained sufficient

evidences that these hydrogen analyzers were qualified for the en-

vironmental conditions in which they were required to operate after

an accident.

The following documents were reviewed:

BECO Equipment Qualification Evaluation sheets for Hydrogen

Analyzer C172 & C173, both dated June 28, 1985.

Test Report 1035-1 " Prototype Qualification Test for Hydrogen

Analyzer System K-III & K-IV" Revision 1 dated September 1981

(COMSIP Inc.).

Each of the containment hydrogen monitors was a rather complex

system. It consisted of numerous solenoid valves and automatic

sequencing circuits, sample pump, pressure regulator, flow meters,

etc. and a control cabinet.

Within the scope of this review, the following matter requiring

licensee's attention was identified:

Page 18 of the Test Report 1035-1 described the yearly, 5 year

and 10 year maintenance requirements for the H 0 analyzer.

22

The yearly maintenance requirement states " carefully inspect for

degradation, replace as necessary". This description appears to

provide less than acceptable guidance relative to performing an

inspection of this safety related system. The qualification l

l

_ _ _ _ _ . _ _ _ _ _ . _

.

.

.

26

maintenance for this-system was not available for review. The

licensee. should clearly identify the inspection acceptance

criteria. (50-293/85-27-51)

10. Exit Interview

The Post-Accident Sampling and Analysis Team met with licensee rep-

resentatives at the conclusion of the inspection on September 20,

1985. The Team Leader summarized the purpose, scope, and findings

of the inspection.

At no time during the inspection was written material provided to

the licensee.

_

--

.

.

ATTACHMENT 1

TO INSPECTION REPORT

50-293/85-27

PERSONS CONTACTED

A. Licensee Personnel

• C. J. Mathes, Station Manager

• K. P. Roberts, Outage Coordinator

• J. F. Crowder, Senior Compliance Engineer

• A. Shatas, Acting Chief Chemical Engineer

• B. Eldridge, Assistant Chief Radiological Engineer, Operations-

• E. T. Graham, Compliance Management, Group Leader

J. Smallwood, Senior Chemical Engineer

• L. Dooley, Training Supervisor, Technical

• W. Hoey, Senior ALARA Engineer

• T. L. Sowdon, Radiological Section Head

• E. Rochelle, Energy Support Services

• T. Kelley, Bartlett Nuclear

• N. Eisenmann, CYGNA

• R. Andrew, I&C Engineering

• R. Velez, Project Manager, EQ

• E. .

De Lemos, Project Manager, TMI Modifications

• D. Sanford, Manager, Training

• P. J. Moraites

R. Fairbank, Deputy Manager of Engineering,-Nuclear

J. Pawlak, Power System Group Leader

L. Perfetti, Power System Engineer

R. Sherry, Assistant System Chief Maintenance Engineer

J Burbank, I&C Technician

.

M. Akhtar, Senior Modification Engineer

B. NRC

• M. McBride, Senior Resident Inspector

0ther members of the licensee's staff were contacted and/or participated

in an exercise of post-accident and effluent monitoring systems during

the inspection.

• Denotes attendance at the exit interview on September 20, 1985.

- _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _

.

.

ATTACHMENT 2

TO INSPECTION REPORT

50-293/85-27

Documentation for NUREG-0737, II.B.3

Pilgrim Nuclear Power Station Emergency Procedures .

-

2.2.113 "H2/02 Analyser and C-19 Systems," dated August 21, 1985.

-

5.7.4.1.1 " PASS Small Volume Liquid Sample From Jet Pump Flow

Sensing Line," dated September 4, 1985.

-

5.7.4.1.2 " PASS Undfluted Liquid Sample and Dissolved Gas Grab

Sample From Jet Pump Flow Sensing Lines," dated

September 4, 1985.

-

5.7.4.1.3 " PASS Small Volume Liquid Sample From Residual Heat

Removal System," dated September 8, 1985.

-

5.7.4.1.4 " PASS Undiluted Liquid Sample and Dissolved Gas Grab

Sample From Residual Heat Removal System," dated

September 4, 1985.

-

5.7.4.1.7 " PASS Dissolved Gas Measurement From Jet Pump Flow Sensing

Lines," dated September 4, 1985.

-

5.7.4.1.8 " PASS Dissolved Gas Measurement From RHR," dated September 4,

1985.

-

5.7.4.1.9 " PASS Iodine / Particulate Sample From Drywell," dated

September 4, 1985.

-

5.7.4.1.10 " PASS Iodine / Particulate Sample From Torus," dated

l September 4, 1985.

! -

5.7.4.1.11 " PASS Iodine / Particulate Sample From Reactor Building,"

dated September 4, 1985.

1

.

'

-

5.7.4.1.12 " PASS 14 ml Gas Sample From Drywell," dated September 4

1985.

-

5.7.4.1.13 " PASS 14 ml Gas Sample From Torus," dated September 4,

l 1985.

-

5.7.5 " Estimating Core Damage," dated August 21, 1985.

,

I _ _ _ - _ _ _ -

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

.

.

2

-

6.5-305 " PASS Source Calibration of PASS Radiation Monitoring

Instruments," dated March 27, 1985.

-

7.11.1 " Analysis of Liquid Samples for Baron (By Spectrophoto-

metry) Under Accident Conditions," dated June 12, 1985.

-

7.11.2 " Analysis of Liquid Samples for Boror (By Plasma Spectro-

metry) Under Accident Conditions," dated June 12, 1985.

-

7.11.3 " Analysis of Liquid Samples for Chloride Under Accident

Conditions," dated July 17, 1985.

l

!

-

7.11.4 Radioisotopic Analysis of Liquid Samples Under Accident

Conditions," dated June 12, 1985.

-

7.11.5 " Radioisotopic Analysis of Gas Samples Under Accident

Conditions," dated June 17, 1985.

-

7.11.6 " Radioisotopic Analysis of Iodine Cartridges Under Accident

Conditions," dated June 21, 1985.

l

-

7.11.7 " Radioisotopic Analysis of Particulate FtIters Under

Accident Conditions," dated June 12, 1985.

-

7.11.8 " Analysis of Dissolved Gas Sample (By Gas Chromatograph)

Under Accident Conditions," dated September 11, 1985. '

-

TP 84-226 " Pre-operational Test for the Post Accident Sampling

System," dated October 26, 1984.

Other Pilgrim Nuclear Power Station Procedur_es

-

" Interim Safety Evaluation by the Office of Nuclear Reactor Regulation

Relative to Technical Specifications Requested by Generic Letter 83-36,"

dated July 5, 1985.

I -

" Safety Evaluation by The Office of Nuclear Reactor Regulation Post-

Accident Sampling System (NUREG-0737, Item II.B.3) " dated July 1,1985. ,.

I

l

l

l

.

._

, .

.

ATTACNeeEMT 3

TO 198SPECTIOtt REPORT

50-293/85-27

COMPARISOM OF AMALYTICAL RESULTS

A. Chemical Analysis

Bo ron

The test data were:

Analysis Licensee NUREG-0737 Actual Commitment

Standard Results Coenitment Pecuireme m %Erro r Meet

100 pps 90 ppa +/- 10% +/-50 ppa -10% Yes

500 ppm 460 ppe +/- 1G% +/-5% - 3% Yes

1000 ppa 1000 ppa +/- 10% +/-5% 0% Yes

Chloride

1 ppa 1.13 ppe +/- 10% 10% +13% No

2 ppa 1.99 ppa +/- 10% 10% - 1% Yes

5 pps 4.97 ppa +/- 10% 10% -

1% Yes

10 paa 9.34 ppa +/- 10% 10% - 7% Yes

Q. Isotopic Anasysis

8toraa I Licensee NUREC-0737 Comeitaent

Isotope Sink PASS Cormitment Requirements %Erro r Meet

5-131 1.29D-4 1.53E-4 -50/+200-% -50/+200% +19% Yes

8-132 4.62E-3 4.75E-3 -50/+200% -50/+200% + 3% Yes

8-133 1.67E-3 2.30E-3 -50/+200% -50/+200% +30% Yes

I-134 1.39E-2 1.65E-2 -50/+200% -50/+200% +19% Yes

8-135 5.01E-3 4.67E-3 -50/+200% -50/+200% - 7% Yes

- _- _ , .. - . .

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

.

.

ATTACHMENT 4

TO INSPECTION REPORT

50-293/85-27

Documentation for NUREG-0737. II.F.1-1.2

Pilgrim Nuclear Power Station Emergency Procedures

-

2.2.55 " Reactor Building Exhaust Radiation Monitoring System,"

Rev. 6, dated May 1, 1985.

-

5.7.2.18 "Offsite Dose Projections and Protective Action Guides for

the General Public," Rev. 4, dated June 8, 1983.

-

5.7.2.22 "Use of the Emergency Dose Assessment System (EDAS), Rev.

dated

-

5.7.2.23 "Use of Off-site Dose Rate Nomograms", Rev. 3, dated

June 8, 1983.

-

5.7.2.25 "Use of HP-85A Off-site Dose Calculators", Rev. 3, dated

August 7, 1985.

Pilgrim Nuclear Power Station Drawings

-

P. + I. D. M-282, " Plant Ventilation Diagram," Rev. E-4, dated

March 25, 1985.

Pilgrim Nuclear Power Station Construction Package

-

PDCR 79-62, " Noble Gas Effluent Monitoring System".

Licensee Correspondence

-

W. H. Deacon, Actg. Mgr., Nuc. Opns BECO to D. G. Eisenhut, Dir. DOL,

dated January 25, 1982.

-

W. H. Deacon, Actg. Mgr., Nr Opns BECO to D. G. Eisenhut, Dir. 00L, dated

April 16, 1982.

-

D. B. Vassallo, Chief, ORD #2 to A. V. Morisi, Mgr. Nuc1 Opns BECO, dated

May 10, 1982.

-

A. V. Mortsi, Mgr. Nuc. Opns BECO to 0. B. Vassallo, Chief, ORB #3,' 00L,

dated May 18, 1982.

, . _-

.

.

2

4

'.

-

A. V. Morisi, Mgr. Nuc. Opns, to D. G. Eisenhut, Dir. DOL, dated June 9,

1982.

-

D. B. Vassalo, Chief, ORB #2 to A. V. Morisi, Mgr. Nuc. Opns BECO Boston,

dated August 9, 1982.

'

-

A. V. Morist, Mgr. Nucl. Opns BECO, to D. B. Vassallo, Chief, ORB #2 00L,

dated September 14, 1982.

-

W. D. Harrington, Sr. VP, Nuclear BEC0, to D. B. Vassallo, Chief ORB #2

00L, dated August 9, 1984.

l

-

D. B. Vassallo, Chief, Operating ORB #2 DOL to W. D. Harrington, Sr. VP

Nuclear, BECO, dated December 17, 1984.

-

W. D. Harrington, Sr. VP Nuclear, BECO to D. B. Vassallo, Chief, ORB #2 '

DOL, dated February 27, 1985.

-

W. D. Harrington, Sr. VP Nuclear, BECO to D. B. Vassallo, Chief, ORB #2

DOL, dated May 18, 1985

,

= -

D.-B. Vassallo, Chief, ORB #2 00L to W. D. Harrington, Sr. VP Nuclear,

BECO, dated July 5, 1985.

'

-

W. D. Harrington, Sr. VP Nuclear, BECO to D. B. Vassallo, Chief ORB #2 o

. DOL, dated May 30, 1985. ,

NRC Memoranda

,

-

W. V. Johnston Asst. Dir. Materials, CEB to G. C. Lainas, Asst. Dir. for

Oper. Reactors, DOL, dated December 14, 1984,

i

) Pilgrim Nuclear Power Station Drawings

J

-

. Drawing No. M239, Sh. 1-3, " Post Accident Sampling and H2 and 02 Analyzer

System," Rev. E3, dated September 13, 1985.

,

Pilgrim Nuclear Power Station References

11.t.1-1 and II.F.1-2

Procedures

-

5.7.3.3 Sampling, Transport and Analysis of Effluent Iodines and

Particulates from the Main Stack Under Emergency Condi-

, tions, PIL 52-F1, dated December 28, 1982.

.,

- _ . . _ _ . _ _ . . . _ . _ _ . . . _ . . - _ , - - _ . - _ _ _ . . _ _ . _ . _ _ . _ _ _ _ _ . _ _ . . _ . _ _ _ _ _ _ _ . _ _ _ _ _

___

.

1

o

3

-

5.7.3.4 " Sampling, Transport and Analysis of Effluent Iodines

and Particulates from the Reactor Building Vent Under

Emergency Conditions", Rev. 3, PIL 52-01, dated

December 28, 1982.

-

~5.7.3.5 " Sampling, Transport and Analysis of Effluent Iodines and

Particulates.from the Turbine Building Under Emergency

. Conditions", PIL 52-115, Rev. 2, dated December 28, 1982.

-

7.4.2.9 " Source Calibration of General Monitor High Range Noble

Gas Monitors", Rev. 1, dated November 24, 1984.

Pilgrim Nuclear Power Station Memoranda

-

R. A. Smith BECO to A.R. Trudeau, BECO, "Representativenats of

Samples from the Main Stack and the Reactor Building.Ver.t", PNPS

i File HP-83-151, dated March 31, 1981.

!

-

R. A. Smith BECO to A. R. Trudeau, BECO, "NRC Inspection 83-12/

Follow-up Item 83-02-02, Representativeness of Samples from the Main

Stack and the Reactor Building Vents" PNPS File HP-83-283, dated

June 30, 1983.

ENTECH Engineering Co. Reports. '

i- -

P10368 "A nomogram for Correlating the Effluent Sample-Line Filter

Radioactivity to the Amount of I-131 Released to the

Atmosphere", dated May 24, 1981.

-

P100-R2 " Didos-III, A Three-Dimensional Point - Kernel Shielding Code

for Cylindrical Sources", dated December 1980.

- P103EC1 " Pilgrim Station Unit 1 -Main Stack and Reactor building Vent

Monitors: Correlation Between Monitor Readings and Off-site

Whole Body Gamma Doses", Volumes I-III, dated June 1980.

-

P103-EC3 " Pilgrim Station Unit 1, Emergency Plan Engineering Calcula-

tions to Correlate Radiation Monitor Readings with Plant Re-

leases and Off-site Dose Rates", Vol. I-III, dated June 1980.

l

-

P100-R3 "A Nomogram for the Interpretation of I-131 Field Sample

Measurements Without the Need of Numberical Calculations", dated

l January 1981.

!

Licensee r +

- ondence

l

l

-

W. ' :iarritt, Mgr. Nuc. Engr. , BECO to D. G. Eisenhut, Dir. DOL, dated

February 27, 1981.

l

i

l

l

l

f

r

.

.

.

4

-

A. V. Morist, Mgr. Nuc. Opns. BECO to T. A. Ippolito, Chief, OR8 #2, DOL,

dated October 16, 1981.

-

T. A. Ippolito, Chief ORB #2, 00L to A. V. Morisi, Mgr. Nuc. Opns. BECO,

dated December 8, 1981.

-

D. B. Vassallo, Chief ORB #2, DOL to A. V. Morisi, Mgr. Nuc. Opns. BECO,

dated March 1, 1982.

-

W. H. Deacon, Accts. Manager, Opns, BECO to D. G. Eisenhut, Dir. 00L,

dated April 16,-1982.

-

' D. B. Vassallo, Chief, ORB #2, 00L to A. V. Morisi, Mgr. Nuc. Opns.,

BEC0, dated May 10, 1982.

A. V. Morisi, Mgr. Nuc. Opns, BECO to D. B. Vassallo, Cheif ORB #2 00L,

dated June 4, 1982.

_ A.1V. Morisi, Mgr. Nuc. Opns, BECO to D. G. Eisenhut, Dir. 00L, dated

-June 9, 1982.

-

D. B. Vassallo, Chief, ORB #2 00L, to A. V. Morisi, Mgr. Nuc. Opns. BECO,

dated August 25, 1982.

P. H. Leach, Proj. Mgr., ORB #2, to W. D. Harrington, Sr. VP Nuclear

BECO, dated November 7, 1984.

-W. D. Harrington, Sr. VP Nuclear BECO to D. B._Vassallo, Chief ORB #2

DOL, dated May 28, 1985.

NRC Memoranda

-

W. E. Kregar, Asst. Dir. Rad. Prot. DSI, to G. C. Lainas, Asst. Dir.

Safety Assessment D0L, dated October 23, 1981.

-

R. W. Houston, Asst. Dir. Rad. Prot. DSI to T. Novak, Asst. Dir. OR. DOL,

dated January 29, 1982.

-

R. W. Houston, Asst. Dir. Rad. Prot. DSI to T. Novak, Asst. Dir. OR. 00L,

dated April 26, 1982.

-

T. Eipen, R-1, Inspection Report 50-293/82-13 dated May 15, 1982.

.

-

M. Shambecky R-1, Inspection Report 50-293/83-02.

-

G. C. Lainas, Asst. Dir. OR, 00L to D. M. Crutchfield, Asst. Dir. Safety

Assessment, DOL, dated October 17, 1984.

.

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l

i

,

,

ATTACHMENT 5

i

TO INSPECTION REPORT 50-293/85-27 l

Documentation for NUREG-0737 ,

ITEM, II.F.1.-3, High Range '

Containment Monitor

G. E.- Product System and Sensors Engineering MEMO No. 127-82-004, dated

i December 20, 1982.

,

'-

Letter, W. H. Deacon (BECO) to D.13. Eisenhut (NRC), dated March 25, 1982 (

(BECO ltr. 82-91).

.

I Letter, H. R. Balfour (BECO) to D. B. Vassallo (NRC), dated May 18, 1982

(BECO Ltr.82-145). t

Letter, D. B. Vassallo (NRC) to A. V. Morisi (BECO), dated May 13, 1983. [

Letter, W. D.- Harrington (BECO) to D. 8. Vassallo (NRC), dated June 1,1983. 7

NRC Inspection Report 50-293/83-02, dated February 23, 1983.

!

Procedure No. 5.7.5, Revision 0, estimating Core Damage, dated March 6, 1985. ,

!

Procedure No. 5.7.2.18, Revision 4, Offsite Dose Projections and Protection

Action Guides for the General Public, dated June 8, 1983.

! Procedure No. 6.5.-296, Revision 1, Source Calibration of the Containment High

Radiation Monitoring System, dated February 10, 1984.

s Varicus Emergency Plan Qualification Cards.

!

• *

1 daedJu$y$b,i98. *

r

l Procedure No. 2.3.2.4, Revision 5, Panel 904 Left Control Room, dated i

December 14, 1984.

1

t

'

Procedure No. 5.7.2.22, Revision 3, Use of the Emergency Dose Assessment

System (EDAS),datedAugust7,1985.  ;

Procedure No._2.2.124 Revision 4, Containment High Rad Monitor System, dated '

May 22, 1985,

i

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D'

o

ATTACHMENT 6

TO INSPECTION REPORT

50-293/85-27

Documentation for NUREG-0737

Item III.D.3.3. Inplant Iodine

Letter, D. 8. Vassallo (NRC) to A. V. Morist (BECO), dated April 8,1982.

Letter, A. V. Morisi (BECO) to D. B. Vassallo (NRC), dated May 5,1982 (BECO)

Ltr82-118).

Procedure No. 5.7.2.19, Revision 0, In-Plant I-131 Air Sampling and Analysis,

dated April 1, 1981.

Procedure No. SI-RP.5700, Revision 0, Calibration of Eberline RAS-1

Regulation Air Samples, dated August 22, 1985.

Procedure No. SI-RP.4701, Revision 0, Operation of the RADS Co Model H809V

Air Sampler, dated August 22, 1985.

Procedure No. PNPS SI-RP 4700, Revision 0, Operation of the RAS-1 Am Sampler,

dated August 22, 1985.

Procedure No. SI-RP.5701, Revision 0, Calibration of Radus Model H809V Air

Sampler, dated August 22, 1985.

Procedure No. 6.5-287, Revision 9 Calibration of Eberline SAM-2, dated

July 12, 1985.

.

_ _ _ _ _ _ _ _ _ _

.

.

ATTACHMENT 7

PASS QA AND DESIGN / INSTALLATION RECORDS

A. General Design and Installation

Boston Edison P&ID No M239 " Post Accident Sampling and H &0

2 2

Analyzer System". Sheet 1 Revision E4 dated June 1985, Sheet 2 Re-

vision E3 dated September 13, 1985, Sheet 3 Revision E4 dated

September 13, 1985.

GE Document No. C5474-SP-1 "BWR Generic PASS Design Requirements"

Revision 1 dated November 24, 1980.

• GE Document No. C5475-SP-5 "LOCA Sampler Installation" Revision 2

dated March 3, 1981.

• Boston Edison Field Revision Notice FRN 80-31-120 dated April 21,

1983 " PASS Modification Sample Supply and Return Piping Installa-

tions & Tie-in".

• Boston Edison FRN 80-31 Attachment A "Special Procedures for Welding

to Valcor Engineering Model V5265-5295 Solenoid Valves - to avoid

overheating of valve, rubber seat welding."

, Boston Edison FRN 80-31-142 " Work Installation for the Removal and

Replacement of Valve Nos. SV-5065-72, SV-5077 and 5078, Liquid

Sample Return Line Containment Isolation Valves". 4 pages

B. Power Supplies to High Range Containment Monitors

• BECO Schematic Diagram E550 Sheet 1 " Containment High Radiation PAM

System - Channel A" Revision El dated May 7, 1985.

• BECO Schematic Diagram E550 Sheet 2 " Containment High Radiation PAM

System - Channel B" Revision El dated May 7, 1985.

• BECO Drawing SE 155 " Station Electrical Single Line Composite

Diagram, 4.16KV & 480V AC System" Sheet 1, Revision E10 dat.d Feb-

ruary 18, 1985 and Sheet 2 Revision Ell dated December 25, 1984.

• BECO Wiring Diagram M227A3 " Post Accident Monitoring C170" Sheet 1,

Revision E2.

• Beco Wiring Diagram M227A5 " Post Accident Monitoring C171" Sheet 1,

Revision E2.