Text

Insp Repts 50-317/85-16 & 50-318/85-14 on 850624-28. Deficiency Noted:Failure to Comply W/Limiting Condition of Operation for post-accident Sampling Sys,Per Tech Spec 3.7.13 & TMI Items II.B.3,II.F.1 & III.D.3.3
	ML20134A753
Person / Time
Site:	Calvert Cliffs
Issue date:	08/06/1985
From:	Cheung L, Clemons P, Knox W, Lessard E, Shanbaky M, Weadock A, Jason White; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML20134A727	List: IR 05000317/1985072; ML20134A726; ML20134A753;
References
	RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-2.F.1, TASK-3.D.3.3, TASK-TM 50-317-85-16, 50-318-85-14, NUDOCS 8508150409
	Download: ML20134A753 (35)
	v • d • e

o-

,

ENCLOSURE 1

U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report Nos. -50-317/85-16

50-318/85-14

Docket Nos. 50-317

50-318

License'Nos. DPR-53 Category C

DPR-69

Licensee: Baltimore Gas and Electric Company.

P. O. Box 1475

Baltimore, Maryland 21203

Facility Name: Calvert Cliff Nuclear Power Plant, Units 1 and 2

Inspection At: Lusby, Maryland

Inspection Conducted- June 24 p 28, 1985

Inspectors: .

C

J./. Whl ,' S o' Ra at n Specialist date

V. I S V

iat pecialist ' da'te

A. A.fe

M~k, or - eer

rndatd

(

L. S. JeheuI' M, R

ati /

.

',

'cialist

rMer

date'

h P. E. C ym Y ~, WsWVf

', 60kh ven National efate '

[W. Knox [o6tra atory

.

~

E. Less gd, Contractor,'Brookhaven National date

Laboratory

j

Approved by: MD/a., / / 9ff

M. M. Shanbaky( Chief, fP Radiation 'date~

Safety Tection

8508150409 850008

PDR ADOCK 05000317

G PDR

.

__

_ _

J .-

,

Inspection Summary:

Inspection on June 24 - 28, 1985 (Report Nos. 50-317/85-16; 50-318/85-14)

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

mentation and status of the following task actions identified in NUREG-0737:

'

II.B.3, Post-accident sampling of reactor coolant and containment atmosphere;

II.F.1-1, Noble gas effluent monitors; II.F.1-2, Post-accident effluent moni-

toring; II.F.1-3, Containment radiation monitoring; and, III.D.3.3, In plant

radiofodine measurements. The inspection involved 190 hours0.0022 days 0.0528 hours 3.141534e-4 weeks 7.2295e-5 months by four region-

based inspectors and two contractors from Brookhaven National Laboratory.

Results: Several deficiencies were identified. The following deficiencies

appear to represent violations of NRC requirements: Failure to comply with the

Limiting Condition of Operation specified by Technical Specification 3.7.13,

" Post Accident Sampling"; Failure to assure environmental qualification of the

Containment High Radiation Monitors pursuant to NUREG-0737 as confirmed by the

NRC Confirmatory Order dated March 14, 1983; Failure to perform surveillances

of gasecus effluent monitoring instrumentation used for post-accident monitor-

ing pursuant to Technical Specification 4.3.3.8, " Radiation Gaseous Effluent

,

Monitoring Instrumentation"; and, failure to implement a personnel training

l program pursuant to the requirements of Technical Specification 6.15, " Iodine

l

Monitoring." Additionally, licensee had not implemented and maintained the

, post-accident sampling system with respect to submittals that were confirmed by

l an NRC Confirmatory Order, dated March 14, 1983.

!

,

l

L

'

o

i

DETAILS

1.0 Persons Contacted

J. A. Tiernan Manager, Nuclear Power, Baltimore Gas & Electric

(BG&E)

P. G. Rizzo Supervisor, Technical Training, Calvert Cliffs Nuclear

Power Plant (CCNPP)

R. E. Denton General Supervisor, Training, CCNPP/BG&E

R. L. Wenderlich Supervisor, Operations Quality Assurance Auditing

L. E. Salyards Senior Engineer - Licensing, BG&E

M. J. Miernicki Principal Engineer - Licensing, BG&E

G. F. Wall Engineering Analyst. CCNPP/BG&E

C. L. Rayburn Emergency Planning Analyst, CCNPP/BG&E

R. B. Sydnon Supervisor - Electrical and Control, CCNPP/BG&E

N. L. Millis General Supervisor - Radiation Safety, CCNPP/BG&E

P. T. Crinigan General Supervisor - Chemistry, CCNPP/BG&E '

B. N. Proctor Technical Support Engineer, CCNPP/BG&E

G. C. Wolf Technical Support Engineer, CCNPP/BG&E

A. Marion Senior Engineer - Electrical Engineering, BG&E

Denotes attendance at the exit interview conducted on June 28, 1985.

Other members of the licensee's staff were also contacted and/or partici-

pated in exercises of the post-accident sampling and the effluent moni-

toring systems during the inspection.

2.0 Purpose

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

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

NUREG-0737, Clarification of TMI Action Plan Requirements:

Task No. Title

II.B.3. Post Accident Sampling Capability

II.F.1-1 Noble Gas Effluent Monitors

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

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

III.D.3.3 Improved Inplant Iodine Instrumentation

under Accident Conditions

3.0 Executive Summary

The following summary is a overview of the most significant findings of

this inspection.

3.1 Post Accident Sampling Capability, Item II.B.3

This review indicated that while the licensee considered the system

to be installed June 1,1983, pursuant to commitments contained in

i

,

O

,

3

1

the NRC " Order Confirming Licensee Commitments on Post-TMI Related

Issues," dated March 14, 1983, the item was not implemented and main-

tained in accordance with those commitments, in that the system was

never demonstrated nor could it function as described in submittals

to the NRC.

At the time of this inspection, in-line analytic components (i.e,

Boron Analyzer and pH Analyzer) were inoperable; and other equipment

(i.e., Radioisotopic Analyzer and Hydrogen /0xygen Analyzer) still

remained to be demonstrated as able to function as specified in sub-

mittals to the NRC. Additionally, it was determined that certain

valves necessary to establish sample flow through the system would

not operate during a system demonstration; and the system's ability

to provide diluted grab samples for backup analysis was not reliable

since a known dilution factor could not be verified.

Upon issuance of a specific Technical Specification referencing post-

accident sampling on February 22, 1985 (i.e., Section 3/4.7.13) the

licensee did document the systems continued inoperability, indicating

that the preplanned alternate method of processing samples was in

effect.

This inspection determined that the licensee was utilizing a sampling

technique involving the station's routine sample sink and post-acci-

dent sampling apparatus originally developed to meet interim require-

ments of NUREG-0578 to meet the backup sampling capability require-

ments of NUREG-0737. This sampling technique had not been submitted

to or evaluated by the Commission as to its adequacy in fulfilling

the requirements of NUREG-0737, Item II.B.3. Review of this sampling

method revealed that it was not preplanned, in that procedures were

not commensurate with actual system configuration, nor were personnel

trained in the method; and it was not a practicable sampling scheme

for post-accident conditions in that it likely involved incurring

personnel exposure in excess of GDC-19 criteria and could not be

demonstrated as a workable solution to post-accident sampling.

3.2 Noble Gas Effluent Monitors, Item II.F.1-1

The licensee is using two systems, the Wide Range Gas Monitoring

(WRGM) system and the Main Steam Effluent Radiation Monitor (MSERM)

system to meet the noble gas effluent pathway monitoring requirements

of NUREG-0737 for noble gas monitoring; however, the following con-

cerns were identified during this inspection:

The following concerns were identified relative to the WRGM system:

A spare parts inventory for timely system repair is not main-

tained by the licensee.

Problems with vent stack flow instrumentation require the use of

a default flow rate value.

l

'

D

.

4

The majority of required surveillance and maintenance procedures

for this system have not been developed. Consequently, formal-

ized training in these procedures has not been given.

The MSERM system will be used to monitor noble gas releases to the

atmosphere through the main steam line pathway. This system has not

been declared operational by the licensee. Equipment has been instal- -

led but final calibrations have not been completed. Licensee commit-

ments require operability for the Unit 1 system by the end of the

current outage and by December 31, 1985, for the Unit 2 system.

The following concerns were identified relative to the MSERM system:

Formal procedures and training controlling the operation and

upkeep of this system have not been developed.

Calibration data showing monitor response to noble gas activity

(in pCi/cc) was not available.

Information was not available demonstrating that the attenuation

of low range gammas by main steam piping had been considered in

determining detector response.

3.3 Sampling and Analysis of Plant Effluents, Item II.F.1-2

The licensee is utilizing the grab sampling capability of the Wide

Range Gas Monitor (WRGM) system to meet the radiciodine and particu-

late effluent sampling requirements of NUREG-0737. The WRGM

system allows diversion of the main vent stack sample stream through

shielded, quick disconnect particulate and iodine filters. These

filters can then be manually transported to the filter analysis point.

Based on a review of system capabilities the WRGM system was deter-

mined to be unacceptable in meeting the sampling requirements of

NUREG-0737, Item II.F.1-2. Of primary concern is the failure of the

licensee to demonstrate the system is providing representative iodine

and particulate sampling; and failure to perform operability surveil-

lance of the system in accordance with technical specifications.

Other concerns identified during this inspection include:

Failure to address adequacy of the installed sample line heat

tracing to provide adequate heating under all ambient tempera-

ture conditions.

Failure to provide adequate procedures and personnel training

for filter removal, handling, and subsequent analysis.

i

l

- , _-

9

4

5

Failure to perform an adequate time and motion study for filter

retrieval and analysis which takes into account all radiation

sources.

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

The licensee's implementation of this NUREG-0737 requirement generally

appeared to be in accord with the specifications. Since this system

is monitoring inside of containment and is expected to function in

accident conditions, such as LOCA, the installation was specified to

be environmentally qualified by NUREG-0737.

The licensee indicated in submittals to the NRC that the system was

installed pursuant to NUREG-0737 requirements. An NRC Confirmatory

Order documented this commitment. However, direct observation of the

installation in Unit I revealed that certain protective sleeving

necessary to assure environmental qualification of the monitors'

electrical connectors at the internal containment penetrations were

not installed due to a maintenance oversight. Such lack of protec-

tive sleeving would compromise the system's operation in accident

environments.

3.5 Inplant Radioiodine Monitoring, Item III.D.3.3

The licensee's implementation of this NUREG-0737 requirement gener-

ally appeared to be in accord with the specifications. The

licensee's Technical Specification 6.15 requires the implementation

of a personnel training program for monitoring radioiodine. Such a

program is defined in the licensee's procedures which specifies a

yearly requirement for training.

However, at the time of this inspection it was found that no personnel

have been trained in this area since February 1984, indicating that

the licensee has failed to implement the program as specified by

procedures.

4.0 TMI Action Plan Generic Criteria and Commitments

The licensee's implementation of the task actions specified in Section

2.0 were reviewed against criteria and commitments contained in the

following documents:

NUREG-0578, TMI-2 Lessons Learned Task Force Status Report and Short-

Term Recommendations, dated July 1979.

Letter from D. G. Eisenhut, Acting Director, Division of Operating

Reactors, to all Operating Power Plants, dated October 30, 1979.

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

November, 1980.

9

.

6

Generic Letter 82-05, letter from D. G. Eisenhut, Director, Division

of Licensing, to All Licensees of Operating Power Reactors, dated

March 14, 1982.

NRC " Order Confirming Licensee Commitments on Post-TMI Related Issues",

dated March 14, 1983.

Regulatory Guide 1.4, " Assumptions Used for Evaluating Radiological

Consequences of a loss of Coolant Accident for Pressurized Water

Reactors".

Regulatory Guide 1.97, Rev. 2, " Instrumentation for Light-Water-

Cooled Nuclear Power Plants to Assess Plant and Environs Conditions

During and Following an Accident".

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".

5.0 Post Accident Sampling System, Item II.B.3.

5.1 Position

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

capability to promptly collect, handle and analyze post-accident

samples which are representative of conditions existing in the

reactor coolant and containment 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-acci-

dent sampling and monitoring systems were reviewed against the cri-

teria identified in Section 4.0 of this report and in regard to

licensee letters, memoranda, drawings and station procedures as

listed in Attachment I.A.

5.2 System Description

The Calvert Cliffs Post Accident Sampling System (CE-PASS) was

designed and built by Combustion Engineering. The system is designed

to permit in-line analysis of the chemical and isotopic content of

reactor coolant. The system also has provisions for the collection

of diluted reactor coolant samples for laboratory analysis.

The CE-PASS is common to both units. It consists of a control panel,

a sampling station, a shielded germanium radiation detector for

isotopic analysis, signal processing and power supply panels,

in-line analyzers for boron, pH, and dissolved gas (hydrogen and

oxygen) analyses, and valve operators. These components are

. . _ _ . _

_ .

.

7

all physically located in the Solid Waste Handling Room on the 45-

foot level of the Auxiliary Building.

From the control panel a chemistry technician can remotely control

and monitor in-line analysis of post-accident reactor coolant, and

quantify the sample with respect to hydrogen, oxygen, boron and pH.

Additionally, in-line radioisotopic analysis is provided to qualify

and quantify various radioisotopes in the reactor coolant. Chloride

determinations are made from a diluted grab sample provided by the

CE-PASS.

The signal processing panel contains the electronic circuitry needed

to process the germanium detector signals for subsequent interpreta-

tion at the chemistry laboratory's multi-channel analyzer. The power

supply panels contain devices used to provide regulated power to the

detector and signal processing circuits.

Figure 1, " Simplified Drawing-CE-PASS." depicts the general arrange-

ment of the system. It is designed to collect and to analyze reactor

coolant samples from:

a. the hot leg directly, at operating pressure;

b. the hot leg via low Pressure Safety Injection (LPSI), at low

pressure; and

c. the containment sump, via the LPSI.

At the time of this inspection, the licensee considered backup sampling

and analysis capability required by NUREG-0737 to be provided by each

unit's Nuclear Steam Supply System (NSSS) sample sink and associated

Post-Accident Sampling Apparatus (PASA) previously used to fulfill

interim post-accident sampling capability requirements specified in

NUREG-0578. By this method about 28 ml of undiluted reactor coolant

is collected in a 1/2" lead shielded sample bomb at the NSSS sink

(See Figure 2). The bomb is then transported to the chemistry labora-

tory and attached to the PASA (See Figure 3) for sample extraction.

The sample then is diluted as necessary to permit chemical and iso-

topic analyses via normal laboratory procedures.

The Containment Atmosphere Sampling System (depicted in Figure 1) is

a unshielded sample bomb assembly that is in-line with each unit's

hydrogen / oxygen analyzer; and is located in the vicinity of

the NSSS sink for each unit. A grab sample is expected to be taken

from the septum of the bomb via syringes. The syringes are then

l transported to the chemistry laboratory for hydrogen / oxygen analyses

via gas chromatography and isotopic analyses using normal isotopic

analysis procedures.

1

_ _ . _ __ _

._, . _ _ .__ _ __

,. _

.

8

5.3 CE-PASS System Status

The CE-PASS operability requirement is specified by Technical Specifi-

cation 3/4.7.13 " Post Accident Sampling" (See Attachment 2).

Following issuance of amended Technical Specifications on

February 22, 1985, the licensee declared the system inopertble on

March 5, 1985, and submitted a Special Report to the NRC to that

effect on March 29, 1985. This report stated that the preplanned

alternate method of processing samples was in effect, and estimated

that the system would be returned to operation by April 15, 1985.

On June 6, the licensee submitted another Special Report to the NRC

which indicated the CE-PASS continued to be inoperable due to com-

ponent failure, that the preplanned alternate sampling method was in

effect, and estimated that~the system would be returned to service by

July 3, 1985.

(Note: At the time of this inspection, the CE-PASS was still con-

sidered inoperable due to component failure. Following this inspec-

tion, the licensee submitted another Special Report on July 22, 1985,

indicating that additional problems with CE-PASS components resulted

in continued -inoperability, and estimated that the system would be

returned to service by July 31, 1985.)

The NRC's " Order Confirming Licensee Commitments on Post-TMI Related

Issues", dated March 14, 1983, documented that the CE-PASS had been

taken out o' service for vendor-recommended modifications and im-

provements; that all of the PASS sampling functions were expected to

be restored by June 1, 1983; and that as an interim measure, the

licensee would use the post-accident sampling system (NSSS sink /PASA

method) which was in use prior to installation of the CE-PASS.

The Order further confirmed, per the licensee's submittals referenced

'

in Section III of the Order, that the licensee would implement and

maintain post-accident sampling via an upgraded post accident samp-

lirg capability (CE-PASS) by June 1, 1983.

Examination of the Facility Change Request documentation affecting

the construction and establishment of the CE-PASS, and records of the

systems preoperational testing failed to indicate that the system was

ever verified by the licensee to be completely operational and able

to perform as indicated in the submittals to the NRC. Further, these

submittals were in reference only to the CE-PASS capabilities and did

not infer or reference any intended use of the NSSS sink /PASA method

as a backup or alternate sampling technique upon establishment of the

CE-PASS as the post-accident sampling system on June 1, 1983. Such

submittals were the bases of NRC's safety evaluation of the system,

which concluded on February 12, 1985, that the CE-PASS system met all

of the criteria

1

. _ . _ -

.

9

associated with NUREG-0737, Item II.B.3, " Post Accident Sampling".

Subsequently, on February 22, 1985, Technical Specifications relative

to the system were issued.

Based on a review of the licensee's maintenance records and dis-

cussions with cognizant personnel, it was observed that there have

been continual problems with the system including leaking valves,

valves which have been inoperative under system pressure, inoperable

system components, and erroneous instrument indications. As a re-

sult, the system has not been available for developing and verifying

procedures, and subsequent training of personnel.

No fully integrated or complete test of the system's capability to

collect and analyze samples had been conducted. For example:

1. The Technical Specification 3/4.7.13 require that the system be

able to collect a reactor coolant sample from the hot leg and

the Low Pressure Safety Injection (LPSI); and a containment sump

sample via the LPSI. Based on the data presented, only sample

acquisition from the Unit 2 hot leg had been tested on June 10,

1985. At the time of this inspection documented tests to

establish that samples can be collected from the LPSI at low

reactor coolant pressure had not been performed.

2. The system's analytical instruments have not been tested using

the NRC recommended standard test matrix solution to determine

possible chemical interference that might affect analytical

capability.

3. The system had not been tested to demonstrate the effectiveness

of the method of sample dilution. Sample dilution capability

was found on June 10, 1985 to be insufficient to permit reactor

coolant chemical and isotopic analyses within the tolerances

specified in licensae submittals.

4. Only one test on June 10, 1985 has been documented to establish

the accuracy of the in-line analysis of radionuclides. This

test revealed that just prior to the inspection, an anyalsis

error of a factor of 80 for some radionuclides could be

expected, and the NUREG-0737 required tactor of 2 could not be

generally achieved.

During the inspection, it was also noted that the germanium detector

efficiencies provided by the vendor have not been verified for

accuracy, and that the radiation background at post-accident

conditions has not been considered in the functioning of the

detector.

l

_ _ _ . _ _ _ _ __ _

.= _ _ - __ - . . - .

.

'

!

,-

"

10

l

These items will be reviewed in a subsequent inspection

! (317/85-16-01; 318/85-14-01).

i

5.4 CE-PASS Capability Demonstration

On June 26, 1985, an exercise of the CE-PASS was attempted, but could

not be performed, in that:

1. The Emergency Procedure (ERPIP 4.4.7.6) provided for th'e

operation of the system had not been updated to agree with the

current design configuration. As a result, it could not be

used. Subsequently, the technicians were required to rely on

the surveillance procedure (RCP-1-407) in an attempt to operate

the system.

2. The licensee was unable to sample from the Unit 2 Hot Leg. It

was later found that the jacking bolt to the Unit 2 Hot Leg

sample acquisition valve, 2-CV-5105, had been previously

tightened down on the valve in an effort to repair a chronic

seat leakage problem. The valve was not tested following this

" repair". The licensee's subsequent evaluation indicated that

the jacking bolt had effectively locked the valve in a closed

position, preventing any operation.

3. Further testing of the system was prevented when CE-PASS sample

exhaust valves 1-SV-6529 and 2-SV-6529 failed to function,

preventing sample flow.

These items will be reviewed in a subsequent inspection

(317/85-16-02; 318/85-14-02).

5.5 Preplanned Alternate Sampling Method Status

With the operability of the post-accident sampling system, CE-PASS,

less than the Limiting Condition of Operation specified in Technical

Specification 3.7.13, the licensee is required to initiate the

" preplanned alternate method of processing specified samples" within

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

The licensee's Special Reports to the NRC (Regional Administrator,

Region I) dated March 29 and June 6, 1985 indicated that such method

was in effect due to CE-PASS inoperability.

, The preplanned alternate backup sampling method as described and

understood from submittals to the NRC dated November 30, 1982,

'

involved the use of the CE-PASS to provide diluted grab samples for

laboratory analysis, with a provision for offsite analysis if

necessary. The licensee never addressed the use of the NSSS sink

I

with reference to the requirements of NUREG-0737 and did not indicate ,

in submittals that the NSSS sink /PASA method would be employed as a '

i backup to the CE-PASS. Consequently, this method was not analyzed

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

.

11

pursuant to the criteria of NUREG-0737, Item II.B.3; was not subject

.

^

to safety evaluation, and consequently could not be considered as the

" preplanned alternate method" referenced in the Technical

Specifications.

However, at least since March 5, 1985, (when the licensee first

declared the system inoperable pursuant to technical specification

requirements) and probably earlier considering maintenance history,

the NSSS sink /PASA assembly provided the licensee's only post ac-

cident sampling capability. Until July 22, 1985, it was considered

by the licensee to be the " preplanned alternate method of processing

specified samples" pursuant to Technical Specification 3.7.13.

The NSSS sink /PASA niethod was reviewed during this inspection effort

to determine if the system could be considered as meeting the

specification of NUREG-0737, Item II.B.3. The following was noted:

1. No approved procedure existed for the operation of the system in

the present configuration.

2. No personnel have been formally trained in the NSSS sink /PASA

method.

3. During a test of the NSSS sink /PASA method the operator

experienced a considerable difficulty in the extraction of the

sample from the bomb. Undiluted reactor coolant was forced out

of the top of a burette in the PASA assembly, with a consequent

loss of sample. This could have resulted in significant

personnel exposure, and facility and equipment contamination in

an actual post accident condition.

4. The PASA assembly was not structurally sound. The operator had

to hold the assembly in place in order to apply enough force to

attach the sample bomb. Additionally, all of the sample

containing components of the PASA assembly, excepting tubing,

were laboratory glassware. Such glassware could be subject to

breakage in a post accident condition rendering the system

inoperable and subjecting personnel, facilities and equipment to

significant contamination.

5. The NSSS sink /PASA assembly was not provided with any shielding

to reduce personnel exposure (excepting 1/2" of lead on the

sample bomb). Additionally, the system has no provisions for

remote operation or handling of components or processes; and

required considerable direct personnel contact with components

and valves that contained undiluted reactor coolant. With the

exception of a shield wall and wide view mirror assembly to

observe coolant flow in the NSSS sink, and a cart with lead

bricks to transport the sample bomb to the radiochemistry

laboratory, the method did not employ provisions to reduce

personnel exposure. Though the procedure (ERPIP 4.4.7.4) did

,

_

.

12

.

specify the use of lead-line gloves and aprons, such equipment

was not available.

6. A time and motion study to demonstrated that a sample could be

collected and analyzed without exceeding GDC 19 dose criteria

has not been conducted. It is unlikely that, without

significant upgrading of personnel training, procedures, and

ALARA considerations, the method would provide a realistic and

workable option in the post accident condition.

7. The analysis procedure (ERPIP 4.4.7.4) did not contain pro-

visions for the analysis of hydrogen and pH for post-accident

samples.

8. The position indicator for the sample line isolation valve 5467

produced erratic indication. Further, the valve could not be

closed after sample acquisition at the NSSS sink.

The NSSS sink /PASA method did not appear to provide a realistic and

practicable option in the post accident condition; and the licensee's

capability did not appear to be " preplanned" with respect to Tech-

nical Specification 3/4.7.13, in that:

1. the method may require personnel exposure beyond the dose limits

of GDC 19;

2. the PASA is not structurally sound and is highly susceptible to

breakage and inadvertent sample spillage;

3. the method relies upon the acceptance of high risk of radio-

logical controls (i.e., personnel exposure, and facility and

equipment contamination) and a low probability of success,

particularly in successive sampling operations;

4. the method compromises the use of the NSSS sink for other samp-

ling activities due to resultant high dose rates, since there is

no sample line purge ability.

5. adequate procedure development and personnel training have not

been performed.

In summary, it appears that the licensee had failed to implement and

maintain the CE-PASS as required by the NRC Confirmatory Order, dated

March 14, 1983; and did not implement a preplanned alternate method

of sample processing sufficient for the requirements of Technical

Specifications 3/4.7.13 (317/85-16-03; 318/85-14-03).

i

l

._ _ , _ _

,_ _ . _ . _

_ _ _ _

,

.

..

.

13

5.6 Containment Air Sampling and Analysis Capability and Status

On June 26, 1985, a containment air sample was collected. However,

the following problems and concerns were noted:

1. A single common key was specified for opening two of the

isolation valves. The valves could not be opened consecutively

, because the switches had been earlier replaced with key capture

switches which prevented the use of a single key for two

switches at the same time.

Although, this change had been made about two months previous to

this inspection, the information had not been incorporated in

procedures or transmitted for personnel training and

information.

It was also indicated by the technician that the same set of

keys should have been capable of opening valves for Unit 1 and

2; however, only one of the two valves could be opened for

Unit 1 with the keys provided.

2. A technical basis for the purge times used in the procedure was

not developed.

'

3. A flow-rate indicator has not been installed to verify gas flow

in the system. A pressure indicator was relied on to indicate

flow-rate. However, since pressure can exist without flow, the

device did not provide positive flow indication. This was con-

firmed when the operator improperly aligned the valves which

resulted in a pressure indication at the collection point with

no flow.

4. Although remote handling tools, lead gloves and a lead-lined

apron were specified for use in the procedure ERPIP 4.4.7.2,

this equipment was not provided or available for use.

5. The procedure required the extraction of a gas sample from the

collection bomb with a syringe that was not rated for the

expected sample pressure.

6. The sample was extracted from a dead leg portion of the sample

rig, which may not be representative of actual containment

conditions.

7. A time and motion study was not performed sufficient to assure

that the sample could be collected and analyzed within GDC-19

dose limits. The sample rig is located in the normal sampling

room, which may be subject to excessive radiation levels after

an accident if the NSSS sink was used for post accident coolant

sampling.

'

'e

.

14

8. The operator appeared unfamiliar with the control panel. This

required him to expend time searching for valve switches. At

one point during the test an incorrect valve was opened.

Although a containment air sample was collected, the activity was too

low to. permit a valid test of the licensee's analytic capability.

> The following items were noted with reference to Core Damage

Assessment ERPIPs:

1. Containment iodine activity is required by procedure to estimate

core damage; however, the containment atmosphere sampling system

design has not been evaluated to determine if a representative

iodine sample can be collected via this mode.

2. The core damage procedure corrects for containment pressure and

temperature. However, based on the method used to collect and

analyze the sample, corrections may not be required.

This area will be examined in a subsequent inspection

(317/85-16-04; 318/85-14-04).

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

6.1 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

design basis range of monitors for individual release pathways, power

supply, calibration 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 4.0

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

procedures as listed in Attachment 1.B.

The licensee's performance relative to these criteria was determined

by interviewing the principal persons associated with the design,

testing, installation and surveillance of the high range gas

monitoring systems, reviewing associated procedures and

documentation, examining personnel qualifications and direct

observation of the system design and operation.

6.2 Findings

Within the scope of this review the following were identified:

.

15

6.2.1 Description and Capability

The licensee has installed a GA Technology's, Wide Range Gas

Monitor (WRGM) system for each unit to monitor noble gas

releases through the main vent stacks. This system was added to

compliment the pre-existing plant main vent monitor and to

increase detection range. The WRGM system operationally

replaces the existing noble gas detectors, however, the existing

detectors are still maintained.

Separate WRGM systems are provided for monitoring the Unit I and

2 main vent stacks. Each system provides 3 channels of varying

sensitivity to provide coverage of the desired dynamic range.

The low range channel consists of an isokinetic sampling head

connected by heat traced tubing to a sample conditioning module

containing particulate and iodine filters. The sample then

passes to the sample detection module which includes a 2 cfm

pump and a plastic scintillator radiation detector. The

intermediate /high range detectors have a separate sampling

system sized for isokinetic sampling at 0.6 cfm, including heat

traced lines and shielded iodine and particulate filters. The

detectors used for this portion of the system are CdTe(C1)

directly coupled to 30 cm3 and .02 cm3 gas volumes.

Collectively, the detectors for the low range and

intermediate /high range sample channels monitor activity

concentrations from 10 7 to 105 pCi/cm 3 . They provide at least

one decade of overlap between ranges.

A microprocessor is included in each WRGM system to control the

sample flow rate, which filter and detector channels are used;

and to compute and display release information.

A simplified diagram of the WRGM system is shown in Figure 4.

The Main Steam Effluent Radiation Monitor System (MSERM) will be

used to monitor potential noble gas releases to the atmosphere

from the main steam line. Two separate radiation monitors are

included in the monitoring system for each Unit, providing one

monitor for each steam generator. The radiation monitors are

situated in the MSIV room of the Auxiliary Building to view the

effluent activity in each main steam line between the steam

generator and the turbine in the piping section preceeding the

safety relief valves.

i

6.2.2 Operational Status

At the time of this inspection, the Unit 1 WRGM system was not

! operational due to a faulty RM-23 readout monitor in the control

! room. An inventory of spare parts was not available to allow

!

i

_ _ _ _

_

'

,

.

16

rapid replacement of the RM-23 board. The Unit 2 WRGM system

was found to be operational.

The licensee has identified problems with the main vent stack

flow detection instrumentation and consequently is using a

default flow rate value for both WRGM systems as an input to the

microprocessor for calculation of release activity. The conse-

, quences of using this default valve with respect to the system's

isokinetic sampling capability requires further investigation.

This will be reviewed during a subsequent inspection (317/85-16-

<

05;-318/85-14-05).

The MSERM System had not been declared operational by the licensee

at the time of this inspection. The detectors and ratemeters

have been installed, but the system requires in place calibration

and the development of alarm setpoints and procedures. Addi-

tionally, review of the licensee's documentation for this system

identified the following concerns:

-

Calibration data showing detector response to noble gas

activity rather than dose rate was not available. This

data will be required to relate monitor readout (mr/hr) to

main steam activity.

-

Information was not available during the inspection demon-

strating that the attenuation of low-energy gammas by the

main steam line piping had been considered in determining

monitor response.

These items will be reviewed in a subsequent inspection (317/

85-15-06; 318/85-14-06).

6.2.3 Procedures and Training

Review of licensee procedures and associated training involving

the WRGM system indicated that procedures and training are in an

early stage of development. Specifically, the following was

identified. ;

!

-

The majority of necessary Surveillance Test Procedures and

associated Preventive Maintenance procedures for the WRGM

system have not been developed.

-

The licensee's Emergency procedures do not specifically

reference the use of the WRGM system as an input method for

obtaining stack release rate. Additionally, a study

evaluating WRGM detector response to varying isotope mixes

predicted for varying time intervals after the accident was

not evaluated by Emergency Planning as to effect on offsite

dose calculations.

.

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

.

't

17

,

.

-

The individual displaying WRGM centrol room readouts during

this inspection could not interrogate the system.

i

Additionally, no formalized procedures or training had been

developed to control activities associated with the operation or

"

maintenance of the MSERM and WRGM systems. Licensee development

of procedores and training in this area will be reviewed

in a subsequent inspection (317/85-16-07; 318/85-14-07).

6.3 Acceptability

The licensee's WRGM system provisionally fulfills the noble gas

monitoring requirements of NUREG-0737, II.F.1-1, pending correction

of the discrepancies identified above. Due to the non-operational

status of the MSERM System, the licensee has yet to demonstrate

capability for monitoring noble gas releases through the steam

-line.

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

, 7.1 Position

NUREG-0737, Item II.F.1-2, requires the provision of a capability for

the collection, transport, and measurement of representative samples

of radioactive iodines and particulates that may accompany gaseous

effluents following an accident. Such activities must be performable

without exceeding specified dose limits to the individuals involved.

The criteria including the design basis shielding envelope, sampling

- media,- sampling considerations, and analysis considerations are set

forth in NUREG-0737, Table II.F.1-2.

Documents Reviewed

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

and analysis system and procedures were reviewed against the criteria

identified in Section 4.0 and in regard to licensee letters, memo-

randa, drawings and station procedures as Itsted in Attachment 1.B.

I

The licensee's performance relative to these criteria was determined

'

. by interviewing the principal persons associated with the design,

testing, installation, and surveillance of the systems for sampling

and analysis of high activity radiotodine and particulate effluents,

by reviewing associated procedures and documentation, by examining *

personnel qualifications, and by direct observation of the system 5

design and operation.

,

-= , -r--,,,.v- - - - , - * y-- e---.--e-, - - - , - - ,-m m.7-, .c . - - - - - - .-,,,,, ---

---n . ---, -- y- - .c_-

.

18

7.2 Findings

Within the scope of this review, the following items were identified:

7.2.1 Descriptions and Capability

The licensee is intending to use the grab-sample capability of

the Wide Range Gas Monitoring (WRGM) system to meet the radio-

iodine and particulate sampling requirements of NUREG-0737, Item

II.F.1-2. As described previously in section 6.0, the licensee

maintains two independent WRGM systems for independent

monitoring of each unit's main vent stack. Each system samples

the main vent stack with two separate sampling pathways; a high

flow, low range pathway and a low flow, intermediate /high range

pathway.

Both the low range and the intermediate /high range sample

pathways of the WRGM system feature a grab-sample capability

with quick disconnect particulate and iodine (silver zeolite)

filters. Filters on the intermediate /high range pathway are

enclosed in shielded casks which can be removed from the

sampling skid for transport. Diversion of the effluent sample

through the grab-sample filters and timing of the grab-sample

duration can be controlled remotely from the control room.

The high flow (2 scfm) low range sampling pathway line for each

WRGM system is made of 3/4 inch outer diameter stainless steel.

The low flow (0.06 scfm), intermediate /high range sampling path-

way is made of 1/4 inch outer diameter stainless steel. These

lines are nearly 200 feet in length and contain about a dozen

right angle bends. The sampling lines are heat traced from each

stack to the sampling filter skids.

In light of the sampling line lengths and pathways, as de-

scribed, significant radioiodine and particulate line losses due

to plateout and deposition can be expected for this sampling

configuration. However, the licensee has not performed a line

loss evaluation to quantify the loss and determine appropriate

correction factors. Currently, the licensee assumes 100% trans-

mittal efficiency through the sampling lines. This item will be

reviewed in a subsequent inspection (317/85-16-08; 318/85-14-

08).

An operability surveillance requirement for the main vent iodine

and particulate sampling capability was established as an

'

amendment to the Technical Specifications on February 22, 1985.

Technical Specification Section 4.3.3.8. requires that, "the main

vent iodine and particulate sampler shall be demonstrated

operable by comparing samples independently drawn from the main

vent at least once per month." The inspector determined by

discussion with licensee chemistry personnel that the required

l

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

.

r

19

>

comparison has never been performed to demonstrate operability I

of either the Unit 1 or Unit 2 sampling capability. This

finding appears to constitute a violation of NRC requirements

-

,

(317/85-16-9;318/85-14-9).

]

The licensee is currently relying on the installed heat tracing

on the sampling lines to prevent condensation and subsequent

sampling problems with the sample lines. However, no evaluation

has been performed to demonstrate the capability of the heat

tracing to provide adequate heating to the sample lines under

all ambient temperature conditions. Additionally, at the time

of this inspection the heat tracing for the Unit 1 WRGM system

was found to be non-operational. The licensee indicated by sub-

sequent telephone conversation on July 29, 1985 that this had

been corrected.

7.2.2 Capability Demonstration i

A demonstration of grab sample acquisition, retrieval, and ,

analysis was performed during this inspection using procedure

RCP 1-405. Two chemistry technicians were assigned the task of

retrieval of the grab sample filter casks. The casks were then

transported to the chemistry laboratory for analysis. The

inspector identified the following problems during the filter

retrieval and analysis operation.

,

The technicians had not received formal training in the !

procedure describing grab sample retrieval.

The procedure did not specify tools or equipment that might

be required. One technician had to exit the area during

the operation to obtain a wrench. This may cause addi-

tional personnel exposure during accident conditions.

The procedure did not provide guidance relative to

personnel exposure considerations.

Filter cask removal from the skid was difficult to perform

and required two technicians approximately 25 minutes.

During this period, the technicians were in close proximity

to the grab sample cask and other scrubbing filters on the

l sampling skid.

< * No procedures were available to control handling of the

,

shield assembly or analysis of the filters in the

laboratory.

No remote manipulating tools were available in the i

chemistry laboratory to handle the filters. !

'

A letter from D. G. Weiss (General Atomics) to H. B. Wylie ,

[ (BG&E), dated 4/26/82, indicated exposure from retrieval of the

- - - -. .

- . - _ - .

.-

r

t

.

i

!

20

,

grab sample cask would be less than 5 rem whole body. Weiss

assumed work at a distance of one meter from a single cask. The

sample was 30 minutes in duration and activity levels were

assumed to be 2 x 102 uCi/cm 3

. The flow rate was 0.06 scfm.

The Weiss study did not account for 1) the time and motion

involved in separating the cask from the skid, 2) the proximity

of the person to the surface of three casks on the skid, and 3)

the activity of radiciodine on the scrubbing filters.

Concerns listed above, that were identified during the

licensee's demonstration of their grab sample capability will be

reviewed in a subsequent inspection (317/85-16-10;

318/85-14-10).

Currently, the grab sample capability of the WRGM system is less

than adequate in fulfilling the requirements of NUREG-0737, Item

II.F.1-2. This system had not been demonstrated to provide

representative samples, surveillance requirements have not been

implemented, procedures and personnel training are insufficient,

and personnel exposure considerations are not complete.

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

8.1 Position

NUREG-0737, Item II.F.1-3 requires the installation of high range

radiation monitors capable of detecting and measuring radiation

levels within the reactor containment during and following an acci-

dent. Specific requirements are set forth in NUREG-0737, Table

II.F.1, Attachment 3.

Documents Reviewed

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

high range radiation monitoring system was reviewed against criteria

identified in Section 4.0 and in regard to licensee letters, memo-

randa, drawings and station procedures as listed in Attachment 1.C.

8.2 Description

The licensee has installed two General Atomics, Incorporated (GA),

Model RD-23 Gamma Radiation Detectors (gamma ionization chambers) in

each reactor unit. Model RP-2C Readout Modules are installed in each

unit's control room. The RD-23 is capable of detecting radiation

from 10' to 10' R/hr. As a safety monitor, it satisfied Class 1E

requirements and is qualified under LOCA conditions per IEEE 323-1974. The detectors are encased in stainless steel to protect

them from containment sprays and high temperatures.

__

.

21

The following specifications apply:

Parameter Description

Range 10 to 10' R/hr

Sensitivity ~ 1x10 ' amp /R/hr

Max. Temperature 350 F

Max. Pressure 70 psig

Humidity 0-100*.' (saturated steam)

Seismic Qualification Per IEEE 344-1975

Adequate vendor calibration sufficient per instrument type certifi-

cation was verified. Though the system is not yet subject to tech-

nical specifications, the licensee conducts monthly surveillance

tests to verify operability, i.e., STP-0-98-1[2], " Containment High

Range Monitor Monthly Test." Vital Class IE power sources are used

for each instrument channel.

Startup and operation of the system is described in 01-35, " Radiation

Monitoring System", which is supported by formalized lessons plans

for personnel training. Calibration test procedures are performed at

refueling outages in accord with M-562-1, " Containment High Range

Monitor Alignment Check" and M-563-1, " Containment High Range Monitor

Source Check." The source check procedure subjects the instrument to

two dose rates within the 10 R/hr range.

Two independent monitors are located in each containment at the 73'

elevation. The monitors are on No. 12 & 22 Steam Generator cubicles,

and on each unit's Pressurizer cubicle. Sufficient view and volume

appears to be monitored by this configuration.

8.3 Findings

Submittals made to the NRC with respect to this monitoring require-

ment and incorporated in the NRC Confirmatory Order dated March 14,

1983, indicated that the installation was as prescribed by

NUREG-0737, i.e., developed and qualified to function in an accident

environment. To this end, each instrument channel is expected to be

comprised of an instrument assembly (i.e., detector,

detector-to-cable connector, cable, and cable-to penetration

connector) that is environmentally qualified for LOCA conditions.

On June 26 and 27, 1985, it was verified that the cable-to--

penetration connectors for each channel in Unit I were not sleeved

with any RAYCHEM sleeving material. Such sleeving was necessary to

assure that the Amphenol cable connector (part #82-816), which was

,

used in lieu of the originally specified part identified in Field

'

Change Request (FCR) 79-1057, would be environmentally qualifled.

l The Field Equipment Change (FEC) 79-1057-6(p) failed to specify the

l sleeving, though required for environmental qualification. The

l

result of this oversight is that Unit 1 installation was not

.

22

sufficient to assure the capability of detecting and measuring

radiation levels within the Unit I reactor containment during and

following an accident, for the period between March 14, 1983 (the

date of the NRC Confirmatory Order), and June 26, 1985.

The licensee later confirmed to NRC management that the proper sleev-

ing was originally installed properly on the Unit 2 cable-to pene-

tration connectors; and the Unit 1 deficiency had been corrected.

Other aspects of this area appear to be in accord with the require-

ments specified in NUREG-0737.

This item appears to constitute a violation of NRC requirements

(317/85-16-11; 318/85-14-11).

9.0 III.D.3.3 Improved Inplant Iodine Instrumentation Under

Accident Conditions ,

9.1 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.

Technical Specifications 6.15, " Iodine Monitoring" requires the

licensee to implement a program which will ensure the capability to

accurately determine the airborne iodine concentration in vital areas

under accident conditions. This program requires the following:

1. Training of personnel,

2. Procedures for monitoring, and

3. Provisions for maintenance of sampling and analysis equipment.

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

iodine monitoring under accident conditions was reviewed against the

criteria in Section 4.0 and in regard to the following documents:

--

Procedure No. ERPIP 4.1.5, Revision 9, " Radiation Protection

Director."

--

Procedure No. ERPIP 4.1.6, Revision ll, "Offsite Monitoring

Team".

--

Procedure No. ERPIP 4.1.7, Revision 9, "Onsite Monitoring Team".

t

l

--

Training Instruction 5, " Emergency Response Training Program".

--

Lesson Plan No. ER-2-6, " SPEC PR0".

!

.

23

9.2 Findings !

An onsite monitoring team is designated to perform in plant radio-

logical surveys, including radiotodine monitoring, in post-accident

conditions. Twelve technicians have been selected to perform the

duty as part of the licensee's program for emergency response.

& A portion of the licensee's program to ensure the capability to

accurately determine iodine concentration in post-accident conditions

is defined by Training Instruction 5, which requires yearly training

for personnel assigned to the onsite monitoring team.

From review of personnel training records it was noted that the last

training to be performed in this area was conducted during February,

1984; and at that time only seven of twelve designated personnel were

provided with such training. This appears to be contrary to the

licensee's program requirements of Technical Specification 6.15, and

associated implementing procedures (317/85-16-12/ 318/85-14-12).

The licensee indicated that ERPIP 4.1.7 defines the scope of acti-

vities for the Onsite Monitoring Team. Review of this procedure

revealed that 1) in plant radioiodine monitoring was not addressed in

the procedure, and 2) the procedure did not detail any in plant moni-

toring activities. This item will be reviewed in a subsequent

inspection (317/85-16-13; 318/85-14-13).

Relative to other areas affecting in plant radiofodine monitoring, it

was noted that the licensee had provisions for the maintenance of

sampling and analysis equipment. Such equipment appeared to be of

sufficient type, quality and quantity to assure an adequate capabi-

lity.

Silver zeolite cartridges are used for iodine sampling during

post-accident conditions, and are subject to clean air purging prior

to radioanalysis.

10.0 Exit Interview

The inspector met with the licensee management representatives (denoted in

Section 1.0) at the conclusion of this inspection on June 28, 1985, to

discuss the scope and findings of the inspection as detailed in this

report. At that time, the licensee's representatives were informed that a

management meeting to discuss these findings and the licensee's corrective

actions would be held on July 11, 1985, at NRC Region I.

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

. .

...

.

Attachment I.A.

Documentation for NUREG-0737, II.B.3

Correspondence.

--

A. E. Lundvall, Jr. , VP BG&E, to Robert A. Clark, Chief, ORB #3, DOL,

dated May 21, 1980.

--

A. E. Lundvall, Jr. , VP BG&E, to D. G. Eisenhut, Dir. DOL, dated

December 15, 1980.

--

A. E. Lundvall, Jr. , VP. BG&E, to D. G. Eisenhut, Dir. DOL, dated July 7,

1981.

--

A. E. Lundvall, Jr. , VP BG&E, to D. G. Eisenhut, Dir. , D0L, dated November

23, 1981.

--

R. A. Clark, Chief ORB #3, DOL to A. E. Lundvall, Jr. VP BG&E, dated

March 15, 1982.

--

A. E. Lundvall, Jr. , VP BG&E, to D. G. Eisenhut, Dir. D0L, dated March 26,

1982.

--

A. E. Lundvall, Jr. , VP BG&E, to D. G. Eisenhut, Dir. DOL, dated April 19,

1982.

--

R. A. Clark, Chief ORB #3, DOL, to A. E. Lundvall, Jr. VP BG&E, dated

June 30, 1982.

' --

A. E. Lundvall, Jr. , VP BG&E, to R. A. Clark, Chief ORB #3, DOL, dated

August 6, 1982.

--

A. E. Lundvall, Jr. , VP BG&E, to R. A. Clark, Chief ORB #3, 00L, dated

November 30, 1982.

--

'R. W. Starostecki, NPE O!, to A. E. Lundvall, Jr. , VP BG&E, dated June 25,

1984.

--

A. E. Lundvall,. . , 3G&E, to J. R. Miller, ORB #3, D0L, dated June 29,

1984.

i

--

A. E. Lundvall, Jr. , VP BG&E, to J. R. Miller, ORB #3, DOL, dated

October 30, 1984.

--

M. D. Patterson, BG&E to J. R. Miller, ORB #3, dated January 15, 1985.

--

J. R. Miller, ORB #3 to A. E. Lundvall, Jr., VP BG&E, dated February 12,

1985.

. . .

.

2

--

D. H. Jahl, ORB #3 to A. E. Lundvall, Jr. , VP BG&E, dated February 22,

1985.

NRC Memoranda

--

W. V. Johnson, NRRET to G. C. Lainas, NRRLO, dated June 24, 1983.

l

--

D. M. Crutchfield NRRLS to G. C. Lainas, NRRLO, dated November 6,1984.

--

W. V. Johnson, NRRET to G. C. Lainas, NRRLO, dated January 17, 1985.

Procedures

--

RCP 2-102, " Operation of the Nuclear Data 6620 System", Rev. 10, dated

January 6, 1984.

--

RCP 2-103, " Operation of the Tracor Northern-11 System", Rev.10, dated

November 21, 1984.

--

RCP 2-105, " Operation of P.A.S.S. Gamma Detector System", Rev. 10, dated

June 20, 1985.

--

RCP 1-407, " Post Accident Sampling System Operation & Analysis", Rev. 10,

dated May 29, 1985.

--

TSP 104, " Flush, Hydrostatic, Pneumatic and Functional Testing of

Modifications to the PASS", Rev. O, dated June 29, 1983.

--

TSP 64, " PASS Flush, Hydro, and Valve Verification", Rev.1, dated May 17,

1982.

--

TSP 58, " Post Accident Sampling System Preliminary Testing", Rev. O, dated

August 11, 1982.

--

TSP 69, " Post Accident Sampling System (PASS) Preoperational Testing",

Rev. O, dated May 28, 1982.

--

ERPIP N0.: 4.4.7.3, " Post Accident Reactor Coolant Sampling", Rev. 9.

--

ERPIP NO.: 4.4.7.6, " Post Accident Sampling System and Analysis, Rev. 2.

--

ERPIP NO.: 4.4.7.1, " Containment RMS Reading Versus Time Following

Accidents", Rev. 9.

--

ERPIP NO.: 4.4.7.2, " Post-Accident Containment Atmosphere Sampling",

Rev. 10.

--

ERPIP N0.: 4.4.7.3, " Post-Accident Reactor Coolant Sampling, Rev. 10.

--

ERPIP N0.: 4.4.7.4, " Post-Accident Reactor Coolant Analysis, Rev. 9

- -

9

.

3

.

+ --

ERPIP NO.: 4.4.7.5, " Post-Accident Hydrogen Analysis", Rev.1.

--

ERPIP NO.: 4.4.5, " Initial Determination of Projected Whole Body Doses,

Rev. 8.

-

--

ERPIP N0.: 4.1.3.1, " Core Damage Assessment", Rev. 8.

,

--

ERPIP N0.: 4.1.3.2, " Core Damage Assessment", Rev. 8.

--

ERPIP NO.: 4.1.3.3, " Core Damage Assessment", Rev. 8.

--

ERPIP NO.: 4.1.3.4, " Core Damage Assessment", Rev. 8.

--

ERPIP NO.: 4.1.3.5, " Core Damage Assessment", Rev. 8.

Drawings

>

--

M-66, " Reactor, Coolant & Waste System Units 1 & 2", Sheets 1, 2 & 3

Rev. 15, dated October 23, 1984.

--

'M-66, "M-72, Reactor and Coolant System Unit No. 1", Rev. 17, dated

January 23, 1985.

--

M-74, " Safety Injection and Containment Spray Systems", Sheets 1, 2, & 3,

Rev. 36, dated December 10, 1984.

--

M-77, " Reactor, Coolant & Waste System Units 1 & 2", Sheets 1, 2, & 3,

Rev. 20, dated September 6, 1984.

--

M-463, " Gas Analyzing System Units 1 & 2", Sheets 1 & 2, Rev. 14, dated

October 15, 1984.

,

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

_

.- _ _. .

.. ,,

..

Attachment I.B.

Documentation for NUREG-0737, II.F.1-1,2

Calvert Cliffs Nuclear Power Plant Emergency Response Plan

--

ERP Pages 5-3, 5-4, 5-5, 5-6, 5-7, 5-8.

Calvert Cliffs Nuclear Power Plant Emergency Implementation Procedures

--

ERPIP 3.0, " Radioactivity Release Quick Estimate", Rev. 11.

--

ERPIP 4.4.6, " Initial Estimate of Fission Product Release Based on

-Environmental Measurements", Rev. 2, dated September 1, 1981.

--

ERPIP 4.4.3, " Initial Determination of Accident Radioactivity Release",

Rev. 8, dated October 28, 1981.

Calvert Cliffs Nuclear Power Plant Final Safety Analysis Report

--

11.2.3, " Radiation Monitoring", Rev. 2.

'Calvert Cliffs Nuclear Power Plant Operating Procedures

--

RCP 1-405, " Operation of the Main Vent Wide Range Noble Gas Monitors",

Rev. O.

ML20134A753

Text

Navigation menu

Search