ML18150A044

Requests Response to Questions Re Insp Repts 50-280/86-42 & 50-281/86-42 Concerning Failed Feedwater Line,Including Identification of Codes,Stds,Specs & Regulatory Requirements Applied to Line

ML18150A044
Person / Time
Site:	Surry
Issue date:	03/16/1987
From:	Sharp P HOUSE OF REP., ENERGY & COMMERCE
To:	Ward D Advisory Committee on Reactor Safeguards
Shared Package
ML18150A040	List: ML18150A041 ML18150A044
References
NUDOCS 8704270042
Download: ML18150A044 (22)

See also: IR 05000280/1986042

Text

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P(NNSY\.VANIA

CAIILOS J. MOOIIHtAD, CAUPOIIHIA

Al. SW1", WASHINGTON

WILLIAM E. DANNEMEVEII.

CAUFOIINIA

It.&. J,ouse of l\eprestntatibes

11111(( SYNA!l OKLAHOMA JACK FIELDS, TtX.0.8 W.J. ,1ur TAUZIN, LOUISIANA

MICHAEL G OKLEY. OHIO IIU IIICHAIIOSON.

NEW MUIICO MICHAEL IILIIIAKIS, PLOIIIDA C:ommittn

on (nrru anb C:ommrru JOHN IIIYANT. TIXAS DAN SCHAlFEII.

COLOIIAOO

TIIIIIY 111\/CL IWNOII JOE IAIITON. 1tlCAI ll>WAN> J. MAlllttY.

SONNY CAUAHAN. AI.AIAMA MASSACMUSml

NOIIMAN F. UNT. NEW YOM SUBCOMMITIEE

ON ENERGY AND POWER IIICUY l.n,UIO. 1'EXAI (lX OFFICIOI IIOII WYOlfl. OIIEGON IIAI.PH II. NALL. TEXAI WAYNE OOWOY. IIISSIIIIPPI

MastJington, me 20515 JOHN D. OIHGEU. MICHIGAN !EX OfflCIOI Mr. David A. Ward, Chairman Advisory Committee

on Reactor Safeguards

1717 R Street Washington,~

20555 Dear Mr. Ward: March 16, 1987 The SubcOtIDDittee

on Energy and Power is investigating

the implications

for the safety of nuclear power plants of the recent Surry accident.

In lar, we are concerned

that (1) despite the designation

of the failed feedwater

line as "a nonsafety

related system," a similar failure in a Boiling Water Reactor could result in the release of radioactive

steam outside the ment structure;

and (2) standards

established

for new nuclear power plants and inspection

procedures

for operational

plants may not adequately

take into account the possibility

of deterioration

of materials.

We are requesting

your response to the following

questions:

1. The NRC Augmented

Inspection

Team Reports Nos. 50-280/86-42

and 50-281/86-42 (NRC team reports) indicate that the failure at the Surry Station was caused by service induced deterioration

of the feedwater

suction line between the condenser

and the feedwater

pump. (a) What codes, standards, specifications

and regulatory

requirements

are applied to the failed f eedwater line and associated

equipment (condenser, feedwater

pumps, steam turbine, pipelines

and components)?

Are these systems classified

as nuclear or non-nuclear?

Are they classified

as safety or nonsafety

related systems? (b) Are these requirements

different

than those applicable

to other tions of the feedwater

and steam lines that are closer to the steam erators and reactor vessel? If so, why are they, and do you think this distinction

is appropriate

in view of what occurred in the Surry Plant accident?

What is the safety justification

for the differences?

8704270042

870417 PDR COMMS NRCC 1 CORRESPONDENCE

PDR

t '* ... '\ * * Mr. David A. Ward -2-March 16, 1987 Cc) If a failure in the feedwater

piping occurred at a similar location, e.g., between the condenser

and feedwater

piping i~ a Boiling Water Reactor nuclear power plant, could radioactive

material be released outside the containment?

Ci) If so, bow much could be released and what would be the consequences

to the surrounding

area? (ii) Row are these areas of the feedwater

and steam lines classified

in Boiling Water Reactors? (iii) In view of the Surry accident, do you think that the

tions of these areas of the power plant Cincluding

the steam turbine, condenser

and feedwater

pumps) are appropriate? (d) What additional

requirements

could be applied to the feedwater

lines, steam lines, steam turbine, feedwater

pumps, condenser

and related ment to improve the safety of nuclear plant operation?

Ce) Do you think the NRC should make any changes in its regulatory ments for Surry or other nuclear power plants in order to implement

lessons learned from the Surry accident?

• 2. The NRC team reports cited erosion/corrosion

induced thinning of pipe metal as the cause of the failure at the Surry Station. Do the design, construction, maintenance

or integrity

monitoring

codes, standards, or other regulations

applied to nuclear power plants adequately

provide for finding or make allowances

for deterioration

of plant components

and piping in service? If not, what regulatory

changes should the NRC make to incorporate

these factors in plant design, inspection

and maintenance

requirements?

3. The two Surry Station nuclear units are very similar in design, nuclear reactor system and age. The units also "share" some support and auxiliary

functions. (a) In view of this dependency, does it seem appropriate

that Unit 1 was not shut down immediately

when the failure occurred in Unit 2? (b) Should the NRC issue any new regulatory

guidance for such situations?

4. Changes in the control room ventilation

system were being implemented

while the plant was running and at the time of the accident.

The NRC inspection

team reports conclude that the modification

work resulted in the control room being flooded with potentially

lethal carbon dioxide gas.

1. "-.. .., .._, "I ,. I * .. \ * * Mr. David A. Ward -3-March 16, 1987 Ca) Are NRC regulations

adequate for modifications

being performed

while plants are operating?

Were these regulations

being observed at the time of the accident? (b) Do you feel that different

procedures

should have been used? Should the NRC make any regulatory

changes to prevent ongoing modification

work from compromising

operational

safety? 5. The NRC inspection

team reports indicate the accident was initiated

by an improperly

maintained

valve. (a) Does it seem appropriate

that the plant was allowed to operate with this valve not functioning

properly?

Are there adequate requirements

for inspections

of such valves? {b) Should the NRC make any regulatory

changes as a result of the maintenance

deficiencies

discovered

during the investigation

of this accident?

6. What actions independent

of NRC regulatory

requirements

should the industry take to implement

lessons learned from the Surry accident?

Thank you for your assistance

with this investigation.

We would appreciate

having your response no later than April 10. PRS:bh q;y, ~JJtfup ChaLrman

' . \ }ACK H FERGUSON President

and Chief Executive

Officer April 9, 1987 * The Honorable

Philip R. Sharp. Chairman, Subcommittee

on Energy and Power Committee

on Energy and Commerce U. S. House of/Representatives

Washington, D. C. 20515 Dear Repr~sentative

Sharp: * Post Office Box 26666 Richmond, Virginia 23261 804. 77J.j271 * VIRGINIA POWER On Marsh 16, 1987,-you informed us of your intent to.investigate

the implications

of* the December 9, 1986 Surry 2 feedwate~

• • pipe rupture. You requested -that we assist you in that investigation

by providing

responses

to six questions

contained

in your letter. Our responses

are attached * . __ ,,, .. , ... As indicat.ed

in my March 20, 1987 letter, we would be happy to discuss our responses

with you or the '*subcommittee

staff -in a meeting that would facilitate

the most complete understanding

of this information.

Very *truly yours, J. H. Ferguson Attachment

cc: Mr. L. W. Zech, Chairman U. S._Nuclear

Regulatory

Commission

Mr. W. H. Owen, Chairman NUMARC Steering Committee

Mr. Z. T. Pate, President

Institute

of Nuclear Power Operations

Mr. J. J. Taylor, Vice President

Electric Power Research Institute

• * Attachment

Question *1(a) The NRC Augmented

Inspection'

Team Reports .Nos. 50-280/86-42

and 50-281/86-42 (NRC team reports) indicate*

that the failure at the Surry Station was caused by service induced deterioration

of the feedwater

suction line between the condenser

and the feedwater

pump. ' . What codes, standards, specifications

and regulatory

requirements

are applied to the failed feedwater

line and associated

equipment (condenser, feedwater

pumps, steam turbine, pipelines

and components)?

Are these systems classified

as nuclear or non-nuclear?

Are they classified

as safety or nonsafety

related system~?--Response / The codes, standards, and specifications

to which the feedwater/condensate

piping was designed and built are: 0 UnHecL .... States of * America Standard Code for Pressure.

Piping USAS B31.l.O Power Piping, 1967 Edition, plus* all applicable

code .cases 0 ASME Boiler and Pressure Vessel Code 0 ASTM Specifications

0 Manufacturers

Standardization

Society of the Valve ana Fitting Industry 0 Section IX Welding Qualification

of* *the* ASME Boiler and Presssure

Vessel Code 0 American Welding So.ciety Specifications

0* Pipe F~bricators

Institute

.. .,,.*..:,;*., The equipment

associated

with the feedwater/cond~nsate

piping was designed and built to equipment

manufacturers

standards

at the time of procurement (circa 1968). F-0r example, the condenser

and feedwater

heaters were built to .Heat *Exchange

Institute (HEI) standards.

The feedwater

heaters were also built in accordance

with Section VIII of the ASME Boiler and Pressure Vessel Code.

I.',** .. ... -. : .. . : :,.: * 2 * the systems. jssociated

with the failed feedwate!/condensat~

piping are not classified

as "nuclear" as defined by USAS B31.l.O Code Case Nl, and are considered

c_onventional

piping. The* condensate

piping systems are classified

as nonsafety-related

except for . ) the emergency

condensate

storage tanks and. the piping systems from these tanks to the suction side of the auxiliary

feedwater

pumps. These c;omponents

0 are classified

as safety-related

and are seismically.supported.

The fe.edwater

system pipi_ng is classified

as . nonsafety-related

except. for pipiri!f, * valves, and -supports from the steam generators

to and including

the f.irst isolation (check) valve outside containment;

auxilia.ry

feedwater

pumps; and-the piping, valves, and supports from the auxiliary

feedwater

pumps to *-. the main feedwater

lines. These compone_nts

are classified

as safety-related

and are seismically

su*pported.

The feedwater

regulator

valves are classified

-as safety-related

but are .not designated

as seismically

supported

components

• . .,: -* .

• 3 * Question l(b) Are these requirements

different

than those applicable

to other portions of the feedwater -and steam lines that are closer to the steam generators

and reactor vessel? If so, why are they, and do you think this distinction

is appropriate

in .view of what occurred in the Surry Plant accident? -What is the_ safety justification

for the differences?

Response . _ .... -,,,__ :-Yes, construction

requirements

for the safety-related

portions of the feedwater

and main steam lines were more stringent.

-The feedwater

piping between the steam generators

and the first isolation (check) valve outside containment

and for the main steam piping from the steam generators

to the non-return

valves were subjected

to additional

inspections;

i.e., all welds in these piping systems were 1oor radiograpbed (x..:rayed).

These additional

inspection

requirements

were e*stablished

to insure weld integrity

and supplement

the verification

of quality workmanship

in implementing

the piping system design. Imposing the additional

safety-related

piping weld inspection_

requirements

would not ha',[e prevented

the piping rupture event at.,,Surry

Unit _2. The event was caused by a flow-induced

erosion/corrosion

phenomenon

unrelated

to the weld integrity

• of the piping. Even if current weld inspection

criteria had .been used in the design and construction

of the feedwater/condensate

piping, the erosion/corrosion

phenomenon

at Surry_would

not have been. prevented.

The design criteri*a

required by USAS B3l. l.O for calculating

the piping minimum wall thickness (pressure

boundary)

and the materials

u_sed for the feedwater/condensate

piping are identical

for the safety arid

related portions of the piping. :*":., ..... _

e 4 * Regarding

the question on differing

requir~ments

for safety and

related *systems or components, the distinction

is justified

to assure that public health and safety is protected

and that there is no undue risk from operation

of a nuclear plant. The.,industry, and. regulators, require very *high standards

of performance*

for those systems and components

necessary

for nuclear safety. We place special emphasis on the systems, components

and structures

needed to prevent or mitigate the consequences

of postulated

radiological

accidents, and to shut down or maintain the unit in a safe shutdown condition.

Nevertheless, portions of the plant not associated

with nuclear -safety, for example, power productio~

or turbine support systems, are also held to high performance

and industrial

safety standards

established

within the electric utility industry.

e 5 * Question l(c) If a failure in the feedwater

piping occurred at*a similar location, e.g., between the condenser

and feedwater

piping in a Boiling Water Reactor nuclear power plant, could radioactive

material be released outside the containment?

.-~.*-. .-~*** (i) If so, how much could be released and what would be the consequences

to the surrounding

area? (ii) How are these areas of the feedwater

and steam lines classified

in Boiling Water Reactors? (iii) In view of the Surry accident, do you think that the classifications

of these areas of the power plant (including

the steam turbine, condenser

and feedwater

pumps) are appropriate?

Response North Anna and Surry Power Stations use Westinghouse-design

pressurized

water reactors which Virginia Electric and Power Company (Virginia

Power) is licensed by the NRC to operate. We are fully qualified

to address questions

regarding

their design, 'construction

and operation.

However, we have no practical

experience

with boiling water reactors and thus do not consider ourselves

qualified

to* r~~po~d to questions

regarding

such designs. :, *. ,.,::. -; . ~--,., *'

• ... ,.., .. e 6 * Question l(d) What additional

requirements

could be applied to the feedwater.

lines, steam lines, steam turbine, feedwater

pumps, condenser

and related equipment

to improve the safety of nuclear plant operations?

Response We have considered

the question of."safety" from three perspectives:

nuclear (radiological)

safety, potential

system interactions

between safety-related*

and nons.afety-related

systems, and finally, industrial (or non....;radiological)

safety._.

From the nuclear safety p~;;pective, no additfonal

requirements

should be applied. The regulatory

requirements

for periodic testing and inspection

programs currently

in place for safety--related

systems provide adequate assurance

that t*hey wil_l perform their intended safety functions.

We also b~1.ieve that the distinction

between safety-related

and nonsafety-related

systems is appropriate

for the reasons cited in response to Question l.b. The issue of system interaction

in nuclear power plants* is currently

• being examined by the NRC (designated

as Unresol~ed

Safety Issue A-17) in concert with industry groups and several nuclear utilities.

The objective-

of this effort is to identify where the current design, analysis, and review procedures

may not adequately

account for potentially

adverse systems interactions

and to recommend

action to rectify deficien~ies.

The current ...... NRC position, pending the completion

of this effort, is that* existing regulatory.

requirements-

and procedures

provide an*adequate

degree of public health and safety assurance.

I 7 * As described

in the NRC team report, certain system interactions

did occur during the Surry event (i.e., inadvertent

fire protection

systems actuation, -security system degradation).

However, these interactions

did not result in a reduction

in nuclear safety. Proper operator/security

force actions and -the use of appropriate

emergency

systems (e.g., control room *emergency

ventilation)

fully mitigated

any system interaction

effects. Regarding

industriat.safety, we deeply _regret the loss of four lives as a result of the Surry* 2 accident.

The activities_

currently

underway within the industr~ (described

in our response to Question 6) should assure that the lessons learned from the Surry 2 event are appropriately

implemented

at all power plants. Although this event occurred*

at a nucl~ar plant, it was not a nuclear accident (-i.e., involving .radioactive

materials)

but rather an industrial

accident.

Other industrial

facilities (e.g., industrial

plants using heated, pressurized

water or fossil-fuel

power plants) could be susceptible

to the erosion/corrosion

phenomenon

experienced

at.Surry.

On -February

10, 1987, we conducted

presentations

across the country to disseminate

information

regarding

the Surry 2 event. A number of major utilities

with fossil-fuel

plants attended.

In addition, we are working with the Electric Power Research Institute (EPRI) and other industry groups to assure the broadest distribution

and understanding

of irformation

related to the single phase liquid erosion/corrosion

phenomenon.

e 8 e Question l{e) Do you think the NRG should make any changes in its regulatory

requirements

for Surry or other* nuclear power plants in order to implement

lessons learned from the Surry accident?

Response -No. As nuclear industry groups address the Surry event, utilities

will be receiving

both the information

and the technology

necessary

to correct the problem. No changes in regulatory

requirements

are necessary.

The nuclear industry's

ability to learn the lessons has improved significantly

since the March 1979 accident at Three Mile Island. The creation of the Institute

of Nuclear Power Operations (INPO) was the first of several steps toward that improvement.

Part of INPO' s mission is to "analyze events* that occur in construction, testing, and operation

of nu~lear plants worldwide

to identify possible precursors

of more serious events; disseminate

the lessons iearned.11 -Utility groups, such as Nuclear Utility Management

and Resources

Committee (NUMARC) ., vendor owners groups, and industry groups such as the Electric Power Research Institute (EPRI),-and the Atomic Industrial

Forum (AIF) represent

other mechanis_lllS

by which lessons learned have, been shared. These groups are currently

being folded under the umbrella of the Utility Nuclear Power Oversight

Committee (UNPOC) to further improve industry's

p_erformance

and enable it to work even more effectively

with the Nuclear Regulatory

Commission (NRG).

-' e 9 * To that end, these industry organizations

are being restructured

into three broad areas: Regulation

and Technical

Support; Communication, Educational

and Technical

Services;

and Government

Affairs. The Regulation

and Technical

Support organization

is intended to be the primary interface

between the industry and NRC, although its scope will also include technical

issues. This organization

will encompass

the functions

of NUMARC primarily

the ability to present* a unified industry position on issues. A NUMARC working group has been formed to address the erosion/corrosion

phenomenon (see our response to Question 2). *ti:' '-

10 ** Question 2 The NRC team r~ports cited erosion/cor~osion

induced thinning of-pipe metal . as the cause of * the * failure at the Surry Station *. * Do * the design, construction,_

maintenance

or integrity

monitoring

codes, standards, or other regulations

applied to nuclear power plants adequately

provide for finding or make allowances

for deterioration

of plant components

and piping-in

service? . If not, what regulatory

changes should -the NRC make to incorporate

these factors in plant design, inspection

and maintenance

requirements?

Response_ ) . Yes, deteric.:,ration

in service is considered." The original construction

specifications

applicable

to this piping were in accordance

with USAS B31. l. 0. With r.espect to corrosior:i

and erosion, USAS B31. l. 0 states: "When corrosion

or erosion is expected, an increase in wall thickness

of the piping shall be provided over that required by other design requirements.

This , allowance

in the judgement

of the designer shall be consistent

with the expected life of the piping~" Our original design provided additional

pipe wall thickness

above that required for ** the, internal system pressure which would have accounted

for any expected corrosio?*

At that time, the complex phenomenon

of erosion/corrosion

was not gener~lly

recognized

in the industry as a problem ih single * phase flow * piping~ systems .and therefore

was not specifically

evaluated.

It is also -important

to recognize

that piping systems made of stainless

steel, or carbon steel containing

lqw temperature, high oxygen water are not susceptible

to this phenomenon.

In-service

testing requirements

for the safety-related

portions of the *** 1* systems are also impoi;ed by the plant's T.echnical

Specifications'

and Section XI of the ASME Boiler and Pressure Vessel Code for Inservice

Inspection.

In addition, -Virginia

Power is expanding

its augmen~ed

program to include / scheduled

inspection, testing, and maintenance*

for applicable

secondary-side

,._ ... p,iping~ .-, *:..-. *.:.. '*< *.l: .

• 11 Until the Surry pipe rupture event, the single phase liquid erosion/corrosion

phenomenon

was neithet widely understriod

nor expected in power plant piping systems. However, the nuclear industry, in conjunction

with EPRI, is developing

a comprehensive ,understanding

of the technical

elements of erosion/corrosion.

We can now discuss qualitatively

the important

variables

affecting

erosion/corrosion.

Reliabl~ nondestructjve

in~peetion

procedures

are available

so that utilities

can determine

the extent of erosion/corrosion

and measure its progression.

A NUMARC worki~g group, chaired by Mr. W. L. Stewart, Vice President-Nuclear

Operations, Virg:i.nia

Power, is coordinating, and evaluating

these industry-wide

inspection

results. They will determine

whether the scope of the concern justifies

additional

action by industry, and if so, what that action should be. We expect that this effort will identify factors in plant design, inspection, and maintenance

requirements

that may have to be modified.

Any regulatory

change, should it be necessary, should only come as a _result of a thorough examination

of the benefits and liabilities

associated

with the change. We are confident

that industry initiatives

will more than satisfy the concerns of regulators

and.that no regulation

to compel action will be required.

... < ' . -* 12 Question 3 The two Surry Station nuclear units are very similar in design, nuclear reactor*_system

and. age. The units also "share" some support and auxiliary

functions. (a) In view of this dependency, does it seem *appropriate

that Unit 1 was not shut down immediately

when the failure occurred in Unit 2? (b) Should the NRC issue any new regulatory

guidance for such , situations?

Response 3(a) Under the circumstances;

it was appropriate

that Unit 1 was not shut down immediately.

Had Unit 1 been adversely

affected, automatic

safety systems as well as trained operations

personnel

were fully capable of *-* shuttin_g

the unit down swiftly and safely. However, Unit 1 was judged by th~ onsite management

and operations.

staff to be -in a safe and stable . steady-state . operating

condition

and any precipitous

action was deemed / unwarranted

until the event was better understood.

In fact, ,, placing Unit 1 .:**:. in a transient

condition

similar to the one in progress on Unit 2* could have increased

risk. During the evening and night of December 9, 1986 we placed emphasis on initiating

a preliminary

investigation

of the Unit 2 event, establishing ,a quarantined

area to preserve evidence, bringing in needed specialists, working with regulators

and the media\*~ and . establishing

a recovery/investigation

organization.

Access to the Unit 1 Turbine Building was re!ftrict~d

to __ preclude personnel

injury iri the event of a similar occurrence -on the Unit.I side. . :;..; * . .; .... On December 10, following

preliminary

inspections

of the-Unit 2 pipe rupture, metallurgists.

had determined

that the probable cause of the pipe failure was thinning *of the pipe .wall* from the inner surface. Because the Unit* 1 feedwater

piping design was .similar, they recommended

inspection

of Unit 1

• 13 e piping. Virginia Power management

immediately

decided to shut Unit 1 down to inspect the wall thickness

of piping. Shutdown of Unit 1 on December 10 was initiated

as soon as Unit 2 was in a cold shutdown cgndition

and the full attention

of* station personnel

could be focused on * the orderly shutdown*

of the operating

unit. We beli~ve that these actions were responsible, well-considered, and, J considering

the circumstances, timely. We believe that it_ was appropriate

to delay th~ s_hutdown

of Unit 1 until we understood

the nature of the event that had occurred on Unit 2 arid were assured that the shutdown could proceed in a controlled

manner. 3(b) No new regulatory

guidance is needed. Because each potential

event is -unique, it is difficult

for us,,. to !:!nvision

regulatory

guidance that would provide information

on how to handle unique events such as the one that occurred at Surry. Rather, the *operating

license and technical

specificat~ons

.~lready provide adequate regulatory

guidance by defining the envelope within which the unit can be safety operated.

In addition, reliance should be placed,-as it is now, ori* a defense-in-depth

design philosophy, redundant

safety systems, highly ~.rained and motivated

personnel, and knowledgeable, responsible

responsible

actions are taken. management

-to assure that appropriate

and

• ,. \ . ( . Question 4 * 14 *Changes in.the coifrol room ventilation

system were being the plant was running and at the time of th~ accident, team reports conclude that the modification

work resulted being flooded.with*potentially

lethal carbon dioxide gas. implemented

while The NRC inspection

in the control room * (a) Are NRC re"gulations

• adequate for modifications

being performed

while piants are operating?

Were these regulations

being *observea

at the time of the accident?

' (b) Do you feel that different

procedures

should have been used?. . Response Should the NRC make any* regulatory

changes to prevent-ongoing modification

work from compromising

operationa,l

safety? As described

in thE: NRC's Augmented

Inspection

Team Report, 50-280/86-42

and 50-281/86-42, some carbon dioxide gas (CO 2) was present in the control room. However, the control room was not described:as "flooded" with carbon dioxide. Rather, it experienced

a mild ingress of CO/Halon.

• Personnel

in the control room were able to carry out their operational

duties safely,* The NRC report attributed

the co 2 to the open doors into the control room area and discussed

... l'modi.fication" work on a ventilation

fan as another P<;>ssible

source. The NRC reference

was to a general area ventilation

fan, l-VS-AC-4, which is nonsafety-related

equipment

outside the control room area boundary.

It supplies conditioned, fresh makeup air to several areas including

the control room . and is isolable by redundant, safety-relate~.'

motor-operated

dampers. At the time of the accident, 1-VS-AC-4

was removed from service due tp maintenance

work (not modifications)

and the isolation

dampers were _operable,*

• 15 * The control room has separate redundant

safety:-related

systems for emergency

air supply and *filtration*

which are described

in the Updated Final Safety Analysis Report (UFSAR) for Surry .Power Station. The control room personnel

turned on the emergency

supply fans for the Main Control Room to dispers*e

and dilute the co 2 , pr~vent its further infiltration, and supply fresh air to the control room. Additionally, two bottled air supply subsystems

were available

and rea*dy for use in conjunction

with the isolation

dampers had it been deemed necessary.

No modifications

were being made to control room ventilation

systems at the* time of the accident;

they were fully operable -at the time of the accident.

The ability to maintain a habitable

control room environment

under emergency.situations

was demonstrated.

NRG regulations

governing

modification

activities

are adequate and compre-hensive. These regulations

govern modifications

to systems as described

in the UFSAR. Developed

to comply with NRG regulations, Virginia Power's design change * program subj ect_s * system modifications

• to strict administrative

controls with numerous safety, technical, management

and independent

organization

reviews. Iri _addition;

modification

and *maintenance

work on safety-related

systems such as the control room emergency

air supply systems is* subject to strict * operability

requirements

set forth in the_ Surry Power Station TechnicaLSpecifications.

.. *, { .. -------< * 16 e Question 5(a) The NRC Inspection

team reports indicated

the accident was initiated

by an.~-improperly

maintained

valve. Does it seem_ appropriate

that-the plant was allowed to operate with this valve. not* functioning

properly?

Are there adequate requirements

for inspections

of such valves? Response The* deficiencies

in the maintenance

procedure

did not affect the valve's ability to perform its intended safety function (i.e., to shut}-.. -Other *administrative

controls required that this capability

be demonstrated

successfully

prior to returning

the unit to operation.

However, as not.ed in the NRC team report, the maintenance

procedure

used to overhaul the yalve* lacked detailed instruct1ons, was not fully followed, and did not provide adequate *documentation.

These deficiencies

have been* corrected.

• Current requirements

assure that a quality maintenance

program be established

and implemented

for safety-related

valves. The main Steam trip valve maintenance

program is an ongoing program which provides adequate assurance

that periodic inspection

of these valves will be performed.

The referenced

maintenance

deficiency

applied to one particular

aspect of one specific procedure

and did not adv:er_sely

affect* the. valve's ability -to * perform its . . . -intended

safety. function.

We conclude th~.t_ adequate requirements

for valve inspections

are -already in place, that known deficiencies

have _ been corrected, and that plant operation

was. appropriate

because the valve's safety function had not been adversely

aff~cted.

We believe it is important

to note that improper valve maintenance

was not the cause of the Surry accident.

Rather, the pipe rupture was the result of ___ __ a chain of events:* a normal pressure transient

in the condensate

system *re*sulting

from a reactor trip t}lat caused the failure of" a portion of -piping that had been severely thinned due to erosion/ corrosion*.

. ... . ... l i!'I ( ;, ....... * 17 Question 5{b) Should the NRC make any regulatory

changes as a result of the maintenance

deficiencies

discovered

dul,"ing the investigation

of this accident?

Response Current, regulations

require that administrative

controls be in place to assure that maintenance

activities

are performed

in a quality manner. The maintenance

deficiencies

that occurred at Surry were not as a result of any programmatic

breakdown, but rather in our implementation

of a specific maintenance

.. procedure.

We don:' t believe that any regulatory

changes are necessary

as a result of this single, isolated occurrence.

In response to concerns from both regulators

and the nuclear industry about maintenance

performance, a NUMARC Work,in*g

Group was established

in late 1984':****_

Its objective

was to facilitate

and accelerate

industry-wide*

maintenance

improvement, assist with technology

transfer, and improve the confidence

that U.S. power stations are being properly maintained.

An industry assessment

of maintenance

programs has been 'completed.

Peer evaluations

are underway.

Event analyses have been conducted

to determine

influence

of maintenance

on plant significant

events .... The Workip,g.(}1:ro~~ _ has assisted INPO in upgrading

evaluation-

criteri~, developing

a guideline

docu1nent

and installing

a maintenanc~,-

trend indicator

program. The Work~ng Group h~s inte_rfaced

with the NRC staff and with Standards

committees

'in the maintenance

area. These, and.other

industry efforts, are expected to continue under the reorganized

irldustry

groups (se~-response to Question l.e.).

• • 18 * Question 6 What actions independent

of NRC regulatory

requirements

should the industry take to implement

lessons learned from the Surry accident:,?

Response Since the event _at Surry station, we have responded

fully to every good faitI::i inquiry related to it. We have sponsored

industry seminars throughout

the country to provide the widest possible dissemination

of information

about the phenomenon

that led to the pipe rupture. In addition, we have worked closely with industry groups to make them aware of the possibility

of piping deterioration.

We have cooperated

closely with INPO in issuing a Significant

Event Report and a Significant

Operating

Experience

Report. We have also helped establish cooperative

program at EPRI and a NUMARC workin~ group to develop a unified industry position and .determine

appropriate

action in response to the Surry event. We believe that these actions, rather than any regulatory

requirements, will be the most effective

means of implementing ,the lessons learned from the Surry event. *, -*. ;,*-~*-* ,,,.-. .. "..:-.:-..