ML20154F950

App to SALP Repts 50-348/88-04 & 50-364/88-04.Viewgraphs Encl

ML20154F950
Person / Time
Site:	Farley
Issue date:	08/26/1988
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20154F508	List: ML20154F505 ML20154F950
References
50-348-88-04, 50-348-88-4, 50-364-88-04, 50-364-88-4, NUDOCS 8809200192
Download: ML20154F950 (35)

See also: IR 05000348/1988004

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August 26, 1988

ENCLOSURE 1

APPENDIX TO ALABAMA POWER COMPANY

FARLEY NUCLEAR PLANT

SALP BOARD REPORT NOS. 50-348/88-04; 50-364/88-04

(DATED JUNE 8, 1988)

8809200192 000907

gDR ADOCK 050 g 8

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August 26, 1988

I. Meeting Summary

A. A meeting was held on July 7,1988, at the Farley site to discuss the

SALP- Board Report for the Farley facility.

B. Licensee Attendees

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B. M. Guthrie. Executive Vice-President

R. P. Mcdonald, Executive Vice President

W. G. Hairston, III, Senior Vice President - Nuclear Operations

J. D. Woodard, Vice President - Nuclear Generation

D. N. Morey, General Manager - Nuclear Plant

G. W. Shipman, Assistant General Plant.Manaccr

J.W.McGowan, Manager,SafetyAuditEngineeringReview(SAER)

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J. E. Garlington, Manager Engineering & Licensing

C. D. Nesbitt, Technical Manager

S. Fulmer, Supervisor SAER

R. B. Wiggins, Supervisor of Operator Training

. J. K. Osterholtz, Manager - Operations

T. D. Arute, Shift Supervisor

R. L. Swif t. Shift Supervisor

C. NRC Attendees

M. L. Ernst Deputy Regional Administrator, Region 11

j A. F. Gibson, Director, Division of Reactor Safety

C. W. Hehl, Deputy Director, Division of Reactor Projects (DRP)

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H. C. Dance, Chief, Reactor Projects Section 18, DRP

, E. G. Adensam, Project Director, Project Directorate 11-1, l

Office of Nuclear Reactor Regulation (NRR) I

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E. A. Reeves, Senior Project Manager, PD 11-1, NRR j

j F. Herr, Deputy Director, OIA '

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N. Perkins, Auditor, OIA

j W. H. Bradford, Senior Resident Inspector, Farley

W. H. Miller, Resident Inspector, Farley l

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II. Errata' Sheet - Farley SALP

Page Line Now Reads Should Read

11 8 .... received a violation for .... received an apparent

for failure ....for counting ' violation for failure ....

gaseous samples for counting gaseous

samples. Subsequent to

the issuance of the report

the licensee has denied

this violation. The NRC

is reviewing this matter. ,

Basis for change: To give a more accurate description of the apparent violation r

which is being contested.

12 c However, g(neral corrosion However, hundreds of pounds

. .. . carbor stee , piping of iron and ....

(See corrected page)

Basis for ch:nt.- To properl/ addrecs the facts in regard to steam generator and <

secondary . tide chendcal treatment.

29 6 Category: .? Category: 1

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Basis for change: To correct aoministra;ive error ,

Unit 1 Unit 2 Unit 1 Unit 2

37 31 8 6

37 32 7 3 6 2

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Basis for change: To correct administrative error.

III. Licensee Comments

Licensee comments submitted in response to the SALP Board report are attached.

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A confirmatory measurements inspection indicated that the

licensee's counting results met the established HRC criterion

for comparing counting results except that a negative bia was

observed for a 14cc vial gas sample from the waste gas ecay

tank. This bias was attributed to sample preparation te niques

because the bias was consistent for the four detector for all

isotopes. During an inspection in March 1988, the icensee

received a violation for failure to make attenuation corrections

for self absorption of gamma photons in a solid pol mer standard

which was used for calibrating the detectors r counting

gaseous samples. Count room equipment was, in general, not

state-of-the-art since it was procured in th early 1970s.

I However, the licensee has ordered new equip ent and expects

l onsite delivery by the latter half of 1988.

A simulated liquid waste sample which c tained H-3, Sr-89,

Sr-90 and Fe-55 was provided to Alabama ower Company in May

1987 by the NRC. The licensee's result compared favorably with

the NRC established criterion for com ring analytical results.

Liquid and gaseous radioactive e fluents were within the

Technical Specification limits and in compliance with 40 CFR 190

limits for radiation dose and r dioactivity concentration in

effluents. Fission and activ ion products in the gaseous

l effluents for 1987 were 35% 1 wer than in 1986. Also, 1987

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values for gaseous iodines a particulates were 75% lower than

1986 values. In general, seous ef fluents for Farley Unit I

have been steadily decli44 9 since 1982 when Farley experienced

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problems with failed 1. Radioactivity in the liquid

effluents was 47% lower S n 1987 as compared to 1986. Tritium in

liquid effluents has, .ained essentially constant for the past

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three years. Gross pha radioactivity in the liquid ef fluent

I was essentially b cY round, 2E-5 curies (C1) per year. Annual

effluent releaseg mmaries for 1985-1987 can be found in

Section V.K. l

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The licensee ported a total of five non-routine releases

(three liquid releases and two gaseous releases) during 1987.

The gaseous eleases occurred on Unit 2 and totalled 8.7 E-6 Ci.  !

These mont red, planned releases were caused by steam generator l

pressure 1se cleaning and steam generator helium leak testing.

The non- outine liquid releases occurred on Unit 1, and a total

i of 4.65 E-5 Ci were released. Two of the releases were due to a

l Refuel ng Water Storage Tank barrier penetration leak, and a

l thir release was caused by a leak in the pumping equipment on l

l the eactor Makeup Water Storage Tank. j

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11 50-348;364/88-04

Corrected 8/26/88

A confirmatory measurements inspection indicated that the licensee's

counting results met the established NRC criterion for comparing counting

results except that a negative bias was observed for a 14cc vial gas

sample from the waste gas decay tank. This bias was attributed to sample

preparation techniques because the bias was consistent for the four

detectors for all isotopes. During an inspection in March 1988, the

licensee received an apparent violation for failure to make attenuation

corrections for self absorption of gama photons in a solid polymer

standard which was used for calibrating the detectors for counting gaseous

samples. Subsequent to the issuance of the report, the licensee has

denied this violation. The NRC is reviewing tt is matter. Count room

equipment was, in general, not state-of-the-art since it was procured in

the early 1970s. However, the licensee has ordered new equipment and

expects onsite delivery by the latter half of 1988.

A simulated liquid waste sample which contained H-3, Sr-89, Sr-90 and

Fe-55 was provided to Alabama Power Company in May 1987 by the NRC. The

licensee's results compared favorably with the NRC established criterion

for comparing analytical results.

Liquid and gaseous radioactive effluents were within the Technical

Specification limits and in compliance with 40 CFR 190 limits for

radiation dose and radioactivity concentration in effluents. Fission and

activation products in the gaseous effluents for 1987 were 35% lower than

in 1986. Also, 1987 values for gaseous iodines and particulates were 75%

lower than 1986 values. In general, gaseous effluents for Farley Unit 1

have been steadily declining since 1982 when Farley experienced problems

with failed fuel, Radioactivity in the liquid effluents was 47% lower in

1987 as compared to 1986. Tritium in liquid effluents has remained

essentially constant for the past three years. Gross alpha radioactivity

in the liquid effluent was essentially background, 2E-5 curies (Ci) per

year. Annual effluent release sumaries for 1985-1987 can be found in

Section V.K.

The licensee reported a total of five non-routine releases (three liquid

releases and two gaseous releases) during 1987. The gaseous releases l

occurred on Unit 2 and totalled 8.7 E-6 Cl. These monitored, planned  ;

releases were caused by steam generator pressure pulse cleaning and steam

generator helium leak testing. The non-routine liquid releases occurred

on Unit 1, and a total of 8.65 E-5 Ci were released. Two of the releases

were due to a Refueling Water Storage Tank barrier penetration leak, and a

third release was caused by a leak in the pumping equipment on the Reactor

Makeup Water Storage Tank.

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Radiation doses to the maximally expnsed offsite individual from

liquid and gaseous effluents for 1987 were calcula ed to be

0.16 mrem to the whole body and 0.17 mrem to the er ical organ.

These values were consistent with previous annual se estimates

and below 40 CFR 190 limits.

The licensee continued to meet the criteria f good chemistry

control established by the Steam Generator ners Group and

Westinghouse. However, general corrosion o carbon steel pipe

throughout the secondary coolant system co inued to result in

hundreds of pounds of "sludge" being tra ported to the steam

generators. Since this sludge had al ady initiated tube

denting, the licensee continued to add oric acid as well as AVT

control chemicals (ammonta and hydra ne) to the feedwater.

This action, in turn, complicated e pH control needed to

prevent general corrosion and pipe inning. Consequently, the

licensee planned to take two majo steps to provide additional

protection to the steam generato s. During refueling outages

(October 1987 and April 1988) t steam generators were cleaned

by a pressure pulse technique in an effort to remove solid

iron-copper oxides from tube- ube sheet crevices and from the

secondary sides of the st am generator tubes. Secondly,

beginning in the next fuel cycles, morpholine will be substi-

tuted for ammonia for pH ntrol in an effort to maintain higher

pH conditions in the gr n steel piping.

Six violations were ntified as follows:

a. Severity Le III violation with three examples:

(1) failureg adequately control access to a high

radiation q (2) failure to follow procedures, and

(3) failudp)r a,to adequately instruct individuals working in

or freque ing a restricted area (348, 364/88-02). '

b. Severit Level IV violation for failure to assure that a

recipi nt was authorized to receive radioactive material

(348,364/86-26). l

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c. Se rity Level IV violation for failure to comply with 00T l

r ulations applicable to the transportation of radioactive '

aterial (348, 364/86-26).

d. Severity Level IV violation for failure to follow the

requirements of a radiation work permit (348, 364/87-28).

e. Severity Level IV violation for failure to maintain records

of survey when local instrumentation was out of service

(364/87-29).

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12 50-348;364/88-04

Corrected 8/26/88

Radiation doses to the maximally exposed offsite individual from liquid

and gaseous effluents for 1987 were calculated to be 0.16 mrem to the

whole body and 0.17 mrem to the critical organ. These values were

consistent with previous annual dose estimates and below 40 CFR 190

limits.

The licensee continued to meet the criteria for good chemistry control

established by the Steam Generator Owners Group and Westinghouse.

However, hundreds of pounds of iron and copper oxide ' sludge' have been

transported to the steam generators each fuel cycle as the result of

general corrosion of carbon steel pipe throughout the secondary coolant

system. Also, iron oxide deposit have been formed in the tube-tube support

regions of the steam generators and indications of cracks have been

observed in tubes, in Unit 2, at these elevations. During the last

refueling outage the steam generators were subjected to a

' pressure-pulse' cleaning in an effort to remove these restriction. The

licensee continued to add boric acid to the secondary coolant to prevent

tube denting. This is consistent with the Owners Group guidelines. The

licensee planned to augment AVT chemistry control by also adding morpho-

line in an effort to establish less acidic conditions throughout the

secondary coolant system and thereby reduce erosion / corrosion.

Six violations were identified as follows:

a. Severity Level III violation with three examples: (1) failure to

adequately control access to a high radiation area, (2) failure to

follow procedures, and (3) failure to adequately instruct individuals

working in or frequenting a restricted area (348, 364/88-02).

b. Severity Level IV violation for failure to assure that a recipient

was authorized to receive radioactive material (348, 364/86-26).

c. Severity level IV violation for failure to comply with DOT

regulations applicable to the transportation of radioactive material

(348,364/86-26).

d. Severity Level IV violation for failure to follow the requirements of

a radiation work permit (348, 364/87-28).

e. Severity Level IV violation for failure to maintain records of survey

when local instrumentation was out of service (364/87-29).

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licensing activity. Thus, many operation's relate questions

from the NRC staff for information surveys or for information

related to event occurrences are answered without n additional-

burden to the plant operations staff.

2. , Conclusions

Category: 2

3. Recommendations

None

K. Training and Qualification Effectiveness

1. Analysis

During the assessment period, spections were conducted by the

resident and regional staf . Inspections included two

licensing examination site isits and one requalifications

program evaluation. Assess ent of training effectiveness were

also made during the OPA n ted previously.

The resident inspectors have had numerous occasions to ect

the training received y licensed and non-licensed per el.

The inspectors h e observed simulator training and ave

reviewed the 1 ed operator requalification training

material. The i ectors have observed and reviewed certain

hands-on train at the training center and have reviewed

instruction mp ial for non-licensed personnel. The training

center is st M of-the-art. The instructors are considered to

be very pro (f ient and well qualified in their positions. The

training pr rams which are prescribed for each craf t are a

required a continuing training evolution. Each program is an

indepth c erage of all required work evolutions. Each training

phase re ired craftsmen to successfully complete an examination

on that ortion of the training. .The observed training has been

profes ional, comprehensive and well received by personnel .

Addi onally, the ten program areas of training for plant

per nnel have been accreditsd by INPO. I

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T e majority of the operators interviewed during the OPA

ndicated that both initial and requalification training were

adequate and had improved substantially over the last two years.

Interviews also indicated that the practice of operating crews

attending roqualification and simulator training as a crew

enhanced the interface and teamwork within the crew. Simulator

training was highly praised and operators indicated that plant

specific events and emergency operating procedures (EOPs) were

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Corrected 8/26/88

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licensing activity. Thus, many operation's related questions

from the NRC staff for information surveys or for information

related to event occurrences are answered without an additional

burden to the plant operations staff.

2. . Conclusions

Category: 1

3. Recommendations

None

K. Training and Qualification Effectiveness

1. Analysis

During the assessment period, inspections were conducted by the

resident and regional staffs. Inspections included two

licensing examination site visits and one requalifications

program evaluation. Assessment of training ef fectiveness were

also made during the OPA noted previously.

The resident inspectors have had numerous occasions to inspect

the training received by licensed and non-licensed personnel.

The inspectors have observed simulator training and have

reviewed the licensed operator requalification training

material. The inspectors have observed and reviewed certain

hands-on training at the training center and have reviewed

instruction material for non-licensed personnel. The training

center is state-of-the-art. The instructors are considered to

be very proficient and well qualified in their positions. The

training programs which are prescribed for each craf t are a

required and continuing training evolution. Each program is an

indepth coverage of all required work evolutions. Each training

phase required craftsmen to successfully complete an examination

on that portion of the training. .The observed training has been

prafessional, comprehensive and well received by personnel.

Additionally, the ten program areas of training for plant

personnel have been accredited by INPO. l

The majority of the operators interviewed during the CPA

indicated that both initial and requalification training were i

adequate and had improved substantially over the last two years.

Interviews also indicated that the practice of operating crews

attending requalification and simulator training as a crew

enhanced the interface .nd teamwork within the crew. Simulator

training was highly praised and operators indicated that plant

specific events and emergency operating procedures (EOPs) were

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detailed, well written and easy to understand. The narrative

sections typically included specific details of the event such as

valve identification numbert, model numbers, number of operable

redundant systems, the date of completion of repair .to provide a

good understanding of the event.

LERs presented the event information in an organ ed pattern with

separating headings and specific information in e h section that led

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to a clear understanding of the event informati . Previous s' .ar

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occurrences were properly referenced in the LE as applicable

The licensee updated some LERs during the sessment period. The

l updated LERs provided new information and e portion of the report

l that was revised was denoted by a vertica line in the right hand

i margin so the new information could eas y be determined by the

reader.

The licensee submitted several report and updates on a voluntary

basis during the assessment period, s stated on page 10 of NUREG-

1022, licensees are encouraged to re ort any event that does not meet

reporting criteria if the licensee Jelieves that the event might be

of safety significance, might be o generic interest or concern, or

contains a lesson to be learned.

l A review of LERs does not i eneral indicate any trend that the

plants are subject to rec ng problems. Recently the licensee

has developed a program rend personnel errors and repetitive

equipment failures. The team noted that all corrective actions

taken were not listed i .he LER and therefore, were not always

correct. Licensee eva ions did not always show that the root

cause was trended or p Jed.

l The distribution of th events analyzed by cause by the licensee were

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Cause Unit 1 Unit 2

Component Failur 8 5

Design / 7 3

Construction, brication, or

Installatio 6 2

Personnel

- Operat4n etivity 5 1

- Maintena ce Activity 4 4

- Test /Ca ibration Activity 3 4

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Other 6 -

Out of alibration - -

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Other 3 -

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Si TOTAL 56 , l

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Corrected 8/26/88

detailed, well written and easy to understand. The narrative

sections typically included specific details of the event such as

valve identification numbers, model numbers, number of operable

redundant systems, the date of completion of repairs to provide a

good understanding of the event.

LERs presented the event information in an organized pattern with

separating headings and specific information in each section that led

to a clear understanding of the event information. Previous similar

occurrences were properly referenced in the LERs as applicable.

The licensee updated some LERs during the assessment period. The

updated LERs provided new information and the portion of the report

that was revised was denoted by a vertical line in the right hand

margin so the new information could easily be determined by the

reader.

The licensee submitted several reports and updates on a voluntary

basis during the assessment period. As stated on page 10 of NUREG-

1022, licensees are encouraged to report any event that does not meet

reporting criteria if the licensee believes that the event might be

of safety significance, might be of generic interest or concern, or

contains a lesson to be learned.

A review of LERs does not in general indicate any trend that the

plants are subject to recurring problems. Recently the licensee

has developed a program to trend personnel errors <ind repetitive

equipment failures. The OPA team noted that all cor.aective actions

taken were not listed in the LER and therefore, were not always

correct. Licensee evaluations did not always show that the root

cause was trended or pursued.

The distribution of the events analyzed by cause by the licensee were

as follows:

Cause Unit 1 Unit 2

Component Failure 6 5

Design 6 2

Construction, Fabrication, or

Installation 6 2

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- Operating Activity 5 0 l

- Maintenance Activity 4 4

- Test / Calibration Activity 3 4 ,

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Other 6 -

Out of Calibration - - *

Other 3 -

TOTAL 39 17

SITE TOTAL 56

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EXITED STATES

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SYSTEMATIC ASSESSMENT

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LICENSE 3 PERFORMANCE

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COMPANY

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AUGUST 1,1986 - MARCH 31,1988

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JULY 7,1986

DOTHAN, ALABAMA

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DIVISION OF REACTOR PROJECTS .

ORGANIZATION .

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REACTOR PROJECTS TECHNICAL SUPPORT STAFF

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CHIEF K. LANDIS

DIR. L. REYES

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l REACTOR PROJECTS REACTOR PROJECTS REACTOR PROJECTS

BRANCH NO. 1 BRANCH NO. 2 BRANCH NO. 3

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j CHIEF D. VERRELLI CHIEF B. WILSON CHIEF V. BROWNLEE 5

PftOJECT3 SECTION PROJECT 3 SECTION PROJECTS SSSTION

NO.1A NO. 2A NO. 3A ,

i OHIEF P. FREDRICKSON -

CHIEF J. CRLENJAK CHIEF T. PEE 3tJS

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BRUNSWICK CRYSTAL RIVER CATAWBA

HARRIS ST. LUCIE McGUIRE

l ROSINSON TURKEY POINT OCONEE

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PROJ1iECTS SECTION PROJECT 5 SECTION PROJECTS BSCTION

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NO.1B _N O . 2 3, MA N

i CHIEF H. DANCE CHIEF F. CANTRELL CHIEF M. 30NKULZ

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J"" NORTH ANNA HATCH

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SURRY VOGTLE

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NRR ORGANIZATION .

j OFFICE OF

NUCLEAR REACTOR

REGULATION PROGRAM MANAGEMENT,

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POUCY DEVELOPMENT,

& ANAL. M ST M

DIR. THOM#.3 E. MURLEY ,

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ASSOC. DIRECTOR FOR

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ASSOC. DIRECTOR INSPECTION &

! FOR PROJECTS TECHNICAL ASSESSMENT

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i DIV. OF ENGINEERING

DEVISION OF AND SYSTEM

, MACTOR PROJECTS I/II TECHNOLOGY

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5. VARGA. DIR. I/Il

j E. ADENSAM. DIR 11-1 'i

i E. REEVES. PROJ. MGR. DiV. OF

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FN , OPERATIONAL EVENTS

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l ASSESSMENT

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DIV. OF REACTOR

1 INSPECTION AND

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AND 3rECIAL PROJEGTS

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PERFORMANCE ANALYSIS AREAS

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FOR OPERATING REACTORS

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1. PLANT OPERATIONS

2. RADIOLOGICAL CONTROLS

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3. MAINTENANCE

4. SURVEILLANCE

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5. FIRE PROTECTION

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6. EMERGENCY PREPARE 0 NESS

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7. SECURITY

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j 8. OUTAGES  !

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9. QUAUTY PROGRAMS

10. UCENSING ACTMTIES

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11. TRAINING

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12. ENGINEERING SUPPORT j

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CMEGORY 1 i

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REDUCED NRC ATTENTION MAY BE APPROPRIATE. j

LICENSEE MANACEMENT ATTENTION AND INVOLVEMENT

ARE AGGRESSIVE AND ORIENTED TOWARD NUCLEAR

SAFETY; LICENSEE RESOURCES ARE AMPLE AND

EFFECTIVELY USED SUCH THAT A HlGH LEVEL OF ,

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PERFORMANCE WITH RESPECT TO OPERATIONAL

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SAFETt OR CONSTRUCTION IS BEING ACHIEVED.

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AREA PERFORMANCE ~d-

CATEGORY 2

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NRC ATTENTION SHOULD BE MAINTAINED AT NORMAL

LEVELS. LICENSEE MANAGEMENT ATTENTION AND

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INVOLVEMENT ARE EVIDENT AND ARE CONCERNED WITH

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NUCLEAR SAFET(; LICENSEE RESOURCES ARE ADEQUATE

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AND ARE REASONABLY EFFECTIVE SUCH THAT SATISFACTORY

PERFORMANCE WITH RESPECT TO OPERATIONAL SAFET( OR l

CONSTRUCTION IS BEING ACHIEVED.

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BOTH NRC AND LICENSEE ATTENTION SHOULD BE

INCREASED. UCENSEE MANAGEMENT ATTEi4Tl0N OR

INVOLVEMENT IS ACCEPTABLE AND CONSIDERS NUCLEAR

S/fETf, BUT WEAKNESSSES ARE EVIDENT; LICENSEE

RESOURCES APPEAR TO BE STRAINED OR NOT EFFECTIVELY

USED, SUCH THAT A MINIMALLY SATISFACTORY PERFORMANCE

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WITH RESPECT TO OPERATIONAL SAFET( OR CONSTRUCTION ,

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IS BEING ACHIEVED.  ;

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EVATLATION CRITERIA

1. MANAGEMENT INVOLVEMENT IN ASSURING QUALITY

2. APPROACH TO RESOLUTION OF TECHNICAL ISSUES

FROM A SAFETY STANDPOINT

3. RESPONSIVENESS TO NRC INITIATIVES

4. ENFORCEMENT HISTORY

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5. REPORTING AND ANALYSIS OF REPORTABLE EVENTS

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6. STAFFING (INCLUDING MANAGEMENT)

7. TRAINING EFFECflVENESS AND QUAUFICATION

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penetrations (86-25) was denied. NRC reviewing.

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CJTEGORY 2 AREAS -.

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! 1. MAINTENANCE

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2. SURVEILLANCE

3. EMERGENCY PREPAREDNESS

j 4. SECURITY  ;

) 5. QUALITY PRCGRAMS  :

l 6. TRAIN NG

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7. ENGINEER NG SUPPCRT

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NONE .

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• ENCLOSURE 2

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• Mabama Peer Company

600 North 18th Street '

Post O'f.ce Sci 2641

Birm.ngham. Mabama 35291 o400

Te ephone 205 2501837 .*T*

...J'

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W G. Heiteten.lH I'

sen<or vice Pres. dent

Nxlea? Operat.ons ,

the SOJhern egCfrC system

August 2, 1988 ,

.

Docket Nos. 50-348

50-364

U. S. Nuclear Regulatory Commission

ATTN: Document Control Desk

Vashington, DC 20555 ,

.

Gentlemen

Joseph M. Farley Nuclear Plant Units 1 and 2

NRC Inspection Report Nos. 50-348/88-04 and 50-364/88-04

By letter dated June 8, 1988, the NRC forwarded the results of the

Systematic Assessment of Licensee Performance (SALP) Board evaluation of

Farley Nuclear Plant for 1988. Alabama Power Company has reviewed this

report and provides comments in an attachment to this letter. ,

Alabama Power Company appreciates the opportunity to provide comments on the ,

SALP report and requests that these comments be considered in the NRC's  !

final conclusion. In addition to the attached comments. Alabama Power  ;

company requests that comments and discussions from the July 7, 1988 meeting i

be taken into consideration for final disposition of the SALP report.  ;

If you have any questions, please advise.

I

Respectfully submitted, ,

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(~. .

k V .,G. Hairston, III l

O

VGH,III/BHVidst-V8.3 /

Attachments

cci Mr. L. B. Long f

Dr. J. N. Grace /

Mr. E. A. Reeves

Mr. V. H. Bradford

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1988 SALP Comments

NRC Inspection Report No.

50-348/88-04 and 50-364/88-04

No. Reference Comment

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l. Page 12, 2nd 1 This paragraph contains several factual errors

(Section IV.B.1) in regard to steam generator and secondary

side chemical treatmentt

The report states, "Since this sludge

had already initiated tube denting,

...

" Sludge has not been shown to

cause tube denting. Crevice hideout

and the resultant crevice pH and

corrosion cause denting. Boric acid

soaks and online addition was

preventatively initiated on Unit 1 due

to support plate crevice corrosion (a

precursor of dentiaii. The same

treatment was initiated on Unit 2 due

to stress corrosion cracking occurring

at support plate intersections

(nondenting related). Tube deformation

has not been substantiated at FNP1 or

FNP2 (approximately 8 tubes in 2A steam

generator have que:tionable

indications. The ether tubes are not in

question). Neither FNP unit has had a

problem with eddy current test probe

passage due to restrictions which vould

be caused by denting.

The report states that the addition of

boric acid "complicated the pH control

needed to prevent general corrosion and

pipe thinning." Boron decreases

secondary pH slightly but does not

cause pH control problems.

The report states, "Consequently, the

licensee planned to take two anjor

steps to provide additional protection

to the steam generators. ...beginning

in the next fuel cycles, morpholine

vill be substituted for ammonia for pH

control in an effort to maintain higher

pH conditions in the carbon steel

piping."

The decision to add morpholine was not

based on inadequate or complicated pH

control but rather on the reduction of

erosion / corrosion and of steam

generator sludge loading that vould be

provided by using morpholine as a

secondary pH elevating additive.

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1988 SALP Comments

NRC Inspection Report No.

50-348/88-04 and 50-364/88-04

Page 2

,No . Reference Comment

2. Page 12, 2nd 1 Morpholine has been added at 4-10 ppm, not

(Section IV.B.1) substituted for ammonia, in Unit 2 Cycle 6 and

Unit 1 Cycle 9. Note that ammonia fr.om

decomposition of hydrazine is the dominant

determinant in steam generator pH control.

3. Page 15 2nd 1 The deviation for failure to control clams in

,

(Section IV.C.1) service water is not indicative of the

l progress that has been made since August 1,

1986. Extensive testing during the SALP

period has resulted in the development of an

,

effective methodology which is environmentally

I acceptable.

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l 4. Page 26, 2nd 1 In discussing problems identified in

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(Section IV.I.1) environmental qualification and procurement

control, the report states, "The licensee has

been slov to acknowledge and correct some of

these problems." APCo disagrees with this

conclusion. Where it could be demonstrated

that problems existed, APCo's corrective

action vas taken in a timely manner. It vould

appear that APCo's efforts to explore

inspection findings as tc their validity has

been interpreted as slov acknowledgment and

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corrective action.

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5. Page 26, 4th 1 The report states, "In January 1988. the

(Section IV.I.1) proposal to install a vent on the 2B charging

pump suction line van canceled." No proposal

was canceled. A design change was voided as a

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result of concerns over the adequacy of the

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proposed design to vent the accumulated

hydrogen and the fact that operational

practices had been adopted to prevent adverse

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l affects to the 2B pump.

The report further states, "The licensee had

been avare of this problem since 1979 but had

not instituted permanent corrective action

other than running or venting the pump."

Contrat u_ this assertion, APCo was not aware

of the(total problem since 1979. This

incorrec erception on the part of the Staff

I was discussed at length in the enforcement

conference.

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. l988 SALP C:mm:nts

NRC Inspection Report No.

50-348/88-04 and 50-364/88-04

Page 3

No. Reference Comment

6. Page 32, 1st & 2nd 1 The report draws conclusions regarding the

(Section IV.L.1) environmental qualification program which APCo

disagrees sith. Whereas the SALP is not an

appropriate forum to thoroughly discuss the

difference of opinions on environmental

qualification, the following concerns are

highlighted:

1. The report states that inspections

found the environmental qualification

program to be carginal during the

early development stages. To the

contrary, correspondence and the NRC

SER seem to indicate the

environmental qualification program

was satisfactory in the early

development stages.

2. The report states that inadequate

staffing was a contributor to

environmental qualification

deficiencies. APCo does not agree

that inadequate staffing was

provided.

3. The report cites "extensive use of

unqualified tseminal blocks in

instrument circuits inside

containment". The issue'on terminal

blocks has bewn thoroughly discussed.

APCo has maintained the blocks vere

qualified but the issue regarding

instrument inaccuracy could not be

resolved until the blocks vere

replaced with qualified splices,

4. It is inappropriate to cite the issue

of upgrade of equipment qualification

in accordance with 10 CFR 50.49(1) in

the SALP report. This issue resulted

from misunderstanding and

miscommunication on behalf of both

APCo and the NRC. It is not

indicative of a programmatic

breakdown in engineering support.

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