ML18150A044
| ML18150A044 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 03/16/1987 |
| From: | SHARP P R HOUSE OF REP., ENERGY & COMMERCE |
| To: | WARD D A Advisory Committee on Reactor Safeguards |
| Shared Package | |
| ML18150A040 | List: |
| 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
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
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
pumps; and-the piping, valves, and supports from the auxiliary
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. *, -*. ;,*-~*-* ,,,.-. .. "..:-.:-..