ML20206G270
| ML20206G270 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 06/18/1986 |
| From: | Stahle C Office of Nuclear Reactor Regulation |
| To: | Milhoan J Office of Nuclear Reactor Regulation |
| References | |
| TAC-60409, TAC-60410, NUDOCS 8606250145 | |
| Download: ML20206G270 (1) | |
See also: IR 05000327/1986020
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MEMORANDUM FOR: James Milhoan, Chief 18 JUN 1586
Plant Systems Branch
Division of PWR Licensing-A
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FROM: Carl R. Stahle, Project Manager
Project Directorate #4
Division of PWR Licensing-A ,
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SUBJECT: REQUEST FOR REVIEW 0F TENNESSEE _VAlIEY AUTHORITY
SUBMITTAL ON CONTAINMENT ~ ISOLATION
Enclosed is a letter dated May 30, 1986, in which the Tennessee Valley Authority
(TVA) provided additional infonnation on the containment isolation design per-
taining to the chemical volume and control system. Although this concern was
originally identified in inspection reports 50-327/86-20-09 and 50-328/86-20-09,
subsequent review of the issue determined that resolution should be transferred
from Region II to the Office of Nuclear Reactor Regulation. Therefore, the
purpose of this memorandum is to request that the Plant Systems Branch review
the enclosure and provide its findings by July 11, 1986. All time expended
towards this effort should be charged to TAC 60409 for Sequoyah Unit 1 and TAC
60410 for Unit 2. If you have any additional infonnation please contact myself
at X27385 or Joe Holonich at X27270. It should be noted that Mr. James Shapaker
of your staff has been involved in a conference call on this issue.
\M
Carl R. Stahle, Project Manager
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Project Directorate #4
Division of PWR Licensing-A
Enclosure:
As stated
DISTRIBUTION: *SEE PREVIOUS CONCURRENCE
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PDR ADOCK 05000327
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TENNESSEE VALLEY AUTHORITY
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CHATTANOOGA. TENNESSEE 374o1
SN 157B Lookout Place *
May 30, 1986
Director of Nuclear Reactor Regulation
Attention: Mr. B. J. Youngblood, Project Director
i PWR Project Directorate No. 4
Division of Pressurized Water Reactors (PWR)
Licensing A
U.S. Nuclear Regulatory Conuaission
Washington, D.C. 20555
Dear Mr. Youngblood:
In the Matter of ) Docket Nos. 50-327
Tennessee Valley Authority ) 50-328
i Please refer to J. A. Olshinski's letter to S. A. White dated April 23, 1986
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transmitting IE Inspection Report Nos. 50-327/86-20 and 50-328/86-20 for the
Sequoyah Nuclear Plant (SQN). The reference inspection report contains
-i unresolved item 50-327/86-20-09 and 50-328/86-20-09, containment Isolation
ij Design Pertaining to the Chemical and Volume Control System. Inclosed is our
' response to NRC's request for additional information made during a telephone
conference call on May 15, 1986 among your staff, Region II, IE-Operating
Reactor Programs Branch, and TVA. TVA would appreciate bringing this
unresolved item to its final conclusion with all parties as soon as possible.
If you have any questions, please get in touch with R. E. Alsup at FTS
858-2725.
Very truly yours,
TENNESSEE VALLEY AUTHORITY
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' R. L. ridley, irector
Nuclear Safety nd Licensing
Enclosure
ec: See page 2
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860530
ADOCK 05000327
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An Equal Opportunity Employer
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Mr. B. J. Youngblood May 30, 1986
cc (Enclosure):
U.S. Nuclear Regulatory Conunission
Region II
Attn: Dr. J. Nelson Crace, Regional Administrator
101 Marietta Street, NW, Suite 2900
Atlanta, Georgia 30323
Mr. James Taylor, Director
Office of Inspection and Enforcement
U.S. Nuclear Regulatory Commission
Washington, D.C. 20555
Mr. Carl Stahle
- . Sequoyah Project Manager
U.S. Nuclear Regulatory Contaission
1 7920 Norfolk Avenue
Bethesda, Maryland 20814
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ENCLOSURE
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RESPONSE - NRC-CIE INSPECTION REPORT
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NOS. 50-327/86-20 AND 50-328/86-20
JOHN A. OLSHINSKI'S LETTER TO S. A. WHITE .
DATED APRIL 23, 1986
Unresolved Item 50-327/86-20-09 and 50-328/86-20-09
BACKCROUND
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IE Inspection Report Nos. 50-327/86-20 and 50-328/86-20 identified an
unresolved item (URI) concerning five (5) chemical and volume control system
(CVCS) containment penetrations. The penetrations involved are penetration
I-16, the normal charging supply, and penetrations I-43A, -43B, -43C, and
-43D, the four reactor coolant pump (RCP) seal injection lines. The URI was
, identified during an Operational Readiness inspection. The issue involves
lack of conformance to the explicit requirements of 10 CFR 50 Appendix A
General Design Criteria (CDC) for containment isolation.
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Subsequent to a letter dated March 25, 1986, from John A. Olshinski
identifying the URI, a telephone conference call was held on April 14, 1986,
among TVA-Licensing / Design Nuclear Engineering (DNE)/Sequoyah (SQN),
Westinghouse, NRR-Containment Systems Branch, IE-Operating Reactor Programs
Branch, and Region II to discuss the URI and determine a course of action
toward resolution. The Westinghouse design bases for the five penetrations
and the FSAR section 6.2.4 and table 6.2.4-1 were discussed at length. It
was determined that a written submittal, answers to remaining NRC questions j
and concerns, would be required. Region II was to coordinate NRC resolution
to the URI. With the assistance of the Region, a list of 'six (6) questions
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was provided by the SQN resident inspector. The inspection report itself,
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received following the April 14, 1986 conference call, principally
j identified misclassifications in FSAR table 6.2.4-1 and stated that
- exceptions should have been taken for the five subject penetrations since r
i their isolation schemes differed from the explicit schemes identified in the
! 10 CFR 50 GDCs. This report also identified resolution to the URI as a
startup item.
Before answers to the six questions could be transmitted to the resident
inspector, TVA Licensing was informed that responsibility for resolution had
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been transferred from Region II to NRR. TVA initiated a second conference
call to reestablish the course for resolution and to define a schedule for
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such resolution since this URI had been identified as a startup item. The
teleconference was held on May 15, 1986 among TVA-Licensing /SQN/DNE.
NRR-Containment Systems Branch, IE-Operating Reactor programs Branch, and
Region II. During this second conference call, TVA presented the design
, bases and requirements for the SQN containment isolation system as a whole
and then specifically for the five lines in question. Technical questions
were asked and answered for the five subject penetrations; the majority of
the conference centered on whether NRC correctly understood TVA's position
regarding the CDCs at the time of licensing, i.e., the actual design for SQN
relative to the then existing 10 CFR 50 CDCs. As a result of the subsequent
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discussions, NRC requested that TVA provide a written summary of the TVA-DNE
presentation made during the conference call regarding the design bases for
the SQN containment isolation system and full information for all
penetrations which have isolation schemes differing from the explicit
schemes identified in the GDCs, i.e. , alternate isolation schemes deemed
acceptable on 'other defined bases.' Clarification of TVA's position at the
time of licensing regarding the SQN design relative to current GDCs was also
requested.
TVA RESPONSE
The two attachments provide the information requested. Attachment 1
provides the design bases for the SQN containment isolation system and
specifically for the five subject penetrations. This includes references to
applicable criteria, standards, supporting documents, system descriptions,
etc. Explanation is also provided of TVA's position at the time of
licensing regarding the SQN design relative to current 10 CFR 50 GDC.
Attachment 2 provides a tabulated listing of pertinent information for SQN
containment penetrations. Specifically identified are the isolation
configurations and bases for penetrations which employ alternate isolation
schemes from those specifically identified in the CDCs. The penetration
isolation configurations--available to the NRC through the FSAR narrative,
tables and flow / control /logie diagrams--are being concisely described and
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summarized herein due to identified typographical errors and apparent
confusion regarding the FSAR narrative and table 6.2.4-1 format.
SUMMARY
TVA believes the existing designs for the five subject penetrations satisfy
all applicable design criteria, provide isolation redundancy consistent with
the intent of current criteria, and are, therefore, fully adequate to ensure
protection of the health and safety of the public. The questions regarding
the technical presentation and format of section 6.2.4 and table 6.2.4-1 of
the FSAR will be resolved by a rewrite of these sections to be submitted
with the next annual FSAR update scheduled for April 15, 1987.
Typographical ereces will be corrected and the presentation of the design
bases and system configurations clarified. While the current 10 CFR 50 CDC
54, 55, 56, and 57 are not the applicable criteria for the design of the
containment isolation system at SQN, review of the existing design relative
to the CDCs indicates the criteria are satisfied in the SQN design, either
directly or by alternate isolation schemes which are acceptable and meet the
intent of the GDCs on 'other defined bases.'
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ATTACHMENT 1
RESPONSE TO NRC QUESTIONS CONCERNING THE DESIGN BASES AND
CRITERIA FOR THE CONTAINMENT ISOLATION SYSTEM FOR ,
SEQUOYAH NUCLEAR PLANT
The following provides a brief summary of the design bases for the SQN
containment isolation system as a whole and specifically for five (5)
chemical and volume control system (CVCS) penetrations of recent concern to
the NRC---the normal charging line and the four reactor coolant pump (RCP)
seal injection lines. A discussion of the design bases relative to current
10 CFR 50 Appendix A General Design Criteria (CDC) is also provided.
SQN is a late 1960s design vintage plant, and was designed, along with other
plants of the same vintage, according to Westinghouse " Systems Standard i
Design Criteria Nuclear Steam Supply System Containment Isolation," 1.14
Revisions 0 and 1. Westinghouse stated before licensing that this is tha
only design standard applicable to SQN; this position was recently
reconfirmed in the attached letter to TVA from Westinghouse dated May 7,
1986. They also stated that they have provided a design for SQN which is in
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keeping with the requirements for a plant of its vintage and consider the
design adequate. It is Westinghouse's position that no backfits or
modifications were required, l
Revisions 0 and 1 of this Westinghouse standard satisfy the requirements of
Criterion 53 of the Atomic Energy Commission (AEC) July 1967, " Proposed
General Design Criteria for Nuclear Power Plant Construction Permits," which
was the applicable regulatory requirement at that time. The July 1967 AEC
Design criterion 53, " Containment Isolation valve," states: " Penetrations
that require closure for the containment function shall be protected by
redundant valving and associated apparatus." The SQN design meets this
requirement, utilizing the previously referenced Westinghouse system
standard. This Westinghouse system standard design criteria includes the
following general containment isolation criteria.
" Piping which penetrates the containment and is not required to function
following a loss-of-coolant accident must be provided with two automatic '
isolation barriers; one barrier must be located outside the containment and
one must be inside the containment.
The definition of an automatic barrier is either a closed system, trip
valve, or check valve. Using this definition, four general classifications
are derived: !
A. Two closed systems - one inside, one outside, no isolation valves
required.
B. No closed systems - one valve inside and one valve outside required.
C. Closed system inside - no valve inside, valve required outside.
D. Closed system outside - no valve outside, valve required inside.
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NOTE:
1. The same criteria apply to lines which are used after an accident except
that manual isolation outside containment is acceptable.
2. For lines 1-inch nominal pipe size and larger which penetrate the
containment and which are connected to the Reactor Coolant System, at
least two valves shall be provided inside the containment. The valves
shall be normally closed or shall have automatic closure. For incoming
lines check valves are permitted and are considered as automatic.
The requirements for a closed system include the following:
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Inside Containment
1 1. Does not conmunicate with either the reactor coolant system or the
- . reactor containment interior.
2. Safety classification same as for engineered safety system
3. Must withstand external pressure and temperature equal to containment
design pressure and temperature
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Must withstand accident transient and environment
5. Must be missile protected
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Outside Containment
1. Does not communicate with the atmosphere outside the containment
2. Safety classification same as for engineered safety systems
3. Internal design pressure and terperature must be at least equal to
containment design pressure and temperature"
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The five lines specifically in question a're the normal charging line and the
four seal water injection lines to the reactor coolant pumps (RCPs). A
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check valve inside containment provides the first isolation barrier for the !
l normal charging line; the second automatic isolation barrier is provided by
a closed system outside containment. Additionally, two automatic system {'
gate valves are available which isolate on a safety injection (SI) signal.
The CVCS seal water injection lines to the RCPs contain check valves inside
containment and are connected to a closed system outside containment which
serves as the second automatie isolation barrier. Additionally, manual
needle valves outside containment may be closed to provide long-term ;
isolation. Within the design basis operation of the plant - normal, '
transient, and accident conditions - loss of seal flow from the charging
i pump (s) is not anticipated, i.e., at least one charging pump remains in
operation. The designs for these penetrations are identified in the
previously referenced Westinghouse system standard. I
The design basis for the normal charging line provides capability for
containment isolation and system isolation to prevent boron injection tank
(BIT) bypass of ECCS injection flow from the centrifugal charging Pumps.
The design basis for the four seal injection lines provides containment l
isolation capability while maintaining injection flow to retain integrity of ,
the RCP reals and use of the RCps, if available, for post-accident
cooldown. Some limited reactor coolant system (RCS) makeup capability is
also provided through the seal flow. !
In susunary, the gQN containment isolation system was designed in accordance
with criteria and standards applicable to all plants of that vintage. The l
applicable criteria are criterion 53 of the 1967 CDC, and the Westinghouse '
systems design standard 1.14. Revisions 0 and 1. The later-issued 10 CFR 50
Appendix A GDC 55, 56, and 57 designate specific allowable isolation
schemes. CDC 55 and 56 acknowledge alternate schemes can be considered l
acceptable for applicable penetrations on 'other defined bases.' A review
of the containment isolation system for SQN relative to the later-issued 10
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CFR 50 CDCs was performed to verify that the design philosophy / intent of the
current GDCs was acceptably employed for SQN even though these design
criteria are not the applicable criteria for SQN. This review, performed by
TVA before licensing, concluded that the SQN containment isolation system
did in fact satisfy the requirements (meet the intent) of the CDCs, either
directly or on 'other defined bases.' Westinghouse review at the time of
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issuance of Revision 2 to their system standard 1.14 (prompted by issuance
of the new GDCs) resulted in their like conclusion that the Revision 0/1
designs were, as such, adequate and, therefore, did not require backfits or
modifications to predesigned plants. The isolation schemes for the five (5)
CVCS penetrations in question reflect standard Westinghouse-supplied and
NRC-reviewed designs and provide the redundant isolation capability required
by both earlier and current design criteria to ensure protection of the
health and safety of the public.
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Westinghouse Water Reactor Nuchar Technolcgy 0iosion
ElppMicfqgyugon DMslons ,, g 3
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TVA-86-565 i
FSD/CWBS-184 l
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May 7, 1986
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Re f. : 1. TVA Memo, 4/18/86
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Mr. J. A. Raulston . 3. TVA Ltr. 3085,
Chief Nuclear Engineer -
7/23/73
TENNESSEE VALLEY AUTHORITY 4. W Ltr FSA-II-13170,
400 West Summit Hill Drive, 11/26/73
W10C126
Knoxville, TN 5. W Ltr FSA-II-TV-15310
37902 ~9/5/74
6.jiLetterRFS-TV-11214
TENNESSEE VALLEY AUTHORITY
SEQUOYAH NUCLEAR PLANT
UNIT NUMBERS 1 AND 2
Containment Isolation Criteria
Dear Mr. Raulston:
Reference 1 transmitted a request from the Tennessee Valley Authority (TVA)
for Westinghouse support in their effort to satisfy NRC questions regarding
the adequacy of the Sequoyah Unit's Seal Water Injection Containment Isolation
design. Listed below is the documentation TVA has requested, followed by a
description of the information Westinghouse is providing in response.
1. AEC design criteria in existence at the time of Sequoyah's design.
The General Design Criteria for Nuclear Power Plant Construction
Permits published by the Atomic Energy Commission dated July 11, 1967.
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2. Draft of the NRC's general design criteria which was also in
existence at the time of Sequoyah's design.
This information is not available; documentation does not appear in
Westinghouse files.
3. Westinghouse design standards.(entire copies) - revisions 0, 1, and 2.
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Westinghouse Systems Standard Design Criteria Nuclear Steam Supply
System Containment Isolation - Revisions 0, 1, and 2. -
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J. A. Raulston 2 May 7, 1986 .
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4. Any NRC/ Westinghouse correspondence that may be pertinent to my case.
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This information is not available; documentation does not appear in
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Westinghouse files.
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5. Documentation supporting the Westinghouse position not to backfit
Sequoyah to the General Design Criteria. .
. Westinghouse /TVA correspondence confinning the adequacy of
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the Sequoyah Containment Isolation and verifying the 1967 AEC
.. . Criteria as the design bases. .
The available documentation requested by TVA is attached for.your use and
information. A copy of this attachment was informally transmitted to
M. Chattin of TVA at the Sequoyah site for review on April 28, 1986.
Westinghouse has reviewed the attachment and considers the following
information pertinent to your case:
o Criterion 53 of the Atomic Energy Comission General Design Criteria
for Nuclear Power Plant Permits dated 1967 is the only criteria for
- containment isolation valves. Criterion 53 states that," Penetrations
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that require closure for the containment function shall be protected
by redundant valving and associated apparatus." (Reference 2)
e " Westinghouse stated that they have provided a design for -
Sequoyah which is in keeping with the requirements for a plant
of its vintage and consider the design adequate." (Reference 3)
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< The applicable Containment Isolation design criteria for the Sequoyah <
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Nuclear Plant are those contained in the July 1967 General Design !
. . Criteria for Nuclear Power Plant Permits. (Reference 4)
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- Westinghouse has the respons1bility to perform the design of the
several fluid systems included in the terms of the contract. These
systems do not include the Containment Isolation System, which must
therefore be in TVA's scope of design. The Westinghouse supplied
systems have incorporated into their design all provisions necessary
for isolation of certain lines penetrating the reactor containment.
(Reference 5) -
e The AEC General Design Criteria 55, 56 and 57 (which became effective
in July 1971) cannot be considdred applicable to the Sequoyah Nuclear
Plant. (Reference 6)
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TVA-86-565
J. A. Raulston 3 May 7, 1986
e Revision 2 of the Westinghouse document 1.14'was transmitted to TVA
for infomation cnly. "Use of the document to view corrections, minor
editorial changes and more detail in some, areas, in conjunction with
Revision 1 is recomended, however, changes of a najor nature due to
Westinghouse position to meeting the intent of the new AEC GCC criteria
55, 56 and 57 are not to be construed .as being required for the
licensing of the Sequoyah Plant." (Reference 6) .
Should you have any questions regarding the attachmer.t. or require additional
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information, please contact the undersigned.
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Very truly yes.s
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WES NGHOUSEEJCTfICCORPORATION
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. L. Williams, Manager
SSD Projects
88. / 4 tag .Mid South Area
/plw/C.A.Ilarmo/05025.0200.0 -
cc: H. L. Abercrombie
R. U. Mathieson
I.'R. Williamson -
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C. C. Mason
M. E. Chattin
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ATTACHMENT 2
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RESPONSE TO NRC QUESTION RECARDING SPECIFIC DESIGN , !
0F CONTAINMENT PENETRATIONS FOR
SEQUOYAH NUCLEAR PLANT j
TVA has again reviewed the SQN containment isolation design, penetration by
penetration, with regard to (1) initial design requirements, (2) current 10 [
CFR 50 CDC classifications and corresponding isolation schemes, and (3) CDC
intent, i.e., acceptability of alternate isolation schemes on 'other defined
bases.' The TVA view of acceptability (on 'other defined bases') is based
on consideration of 1967 CDC-Criterion 53, Westinghouse system design
standard 1.14 Revision 0/1, the general requirements of 10 CFR 50 CDC 54,
and the specific process line function (normal and accident conditions). It
is readily apparent that even for a new CDC plant, numerous classes of
penetrations cannot conform to the CDC specified isolation schemes, e.g., ;
ECCS lines, instrument lines, equipment hatches, airlocks, flanged I
penetrations, electrical penetrations, etc. Alternate schemes must be
employed to provide the redundant isolation capability consistent with
criteria 54 and still permit system operation. CDC 55, 56, and 57 provide ,
requirements for three general classifications of penetrations and are, !
therefore, somewhat subject to interpretation. Based on our interpretation
of these criteria and the current 10 CFR 50.2 definition (v.) for reactor
a coolant pressure boundary, we have identified for each penetration:
(1) penetration classification (CDC 55, 56, or 57), (2) physical '
configuration (barrier inside, barrier outside). (3) the applicable FSAR
figure (if available), and (4) the 'other defined basis' for acceptability
if the isolation scheme differs from 10CFR50 CDC identified schemes.
Supporting description, notes and/or references are provided as necessary.
This information is presented in the form of two tables. Table 2.1 lists
penetrations for which the design correlates with the explicit CDC
designated isolation schemes. Table 2.2 lists penetrations for which their
design correlates with the intent of the CDC and the design is determined
acceptable on 'other defined bases.' only principal process line isolation
barriers are identified; penetration branch takeoffs such as vent, drain and
test lines, and instrumentation sensing taps are not addressed within the
scope of this presentation,
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