JAFP-89-0689, Responds to NRC Re Violations Noted in Rept 50-333/89-80.Corrective Actions:Util Established Formal Preventive Maint Program in 1984 Which Includes 600-volt Class 1E Circuit Breakers,Motor Starters & Transformers
| ML20248A368 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 09/21/1989 |
| From: | Fernandez W POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20248A375 | List: |
| References | |
| JAFP-89-0689, JAFP-89-689, NUDOCS 8910020193 | |
| Download: ML20248A368 (25) | |
Text
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2 James A.FitrPatrick
>e s Nucteer Power Plant . :1 P.O. Box 41 !
Lycoming. New York 93093 '
315 342-3840 1
4 L #D NewWrkPbwer L
g gg William Fernandez ll Resident Manager ;
i September 21, 1989 I JAFP 89-0689 Urlite.d- States Nuclear Regulatory Commission Mail Station F1-137 Washington, D.C. 20555 ATTN: Document Control Desk
SUBJECT:
JAMES A FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333-INSPECTION NO. 89-80
Enclosures:
- 1) Response to Notice of Violation 1.0 )
- 2) Response to Notice of Violation 2,0
- 3) Response to Notice of Violation 3.0
- 4) Response to Notice of Violation 4.0
- 5) Response to Notice of Violation 5.0 l Gentlemen:
In'accordance with the provisions of 10 CFR 2.201, we are submitting our response to' Appendix A Notice of Violation i transmitted by your letter dated AugiIst 22, 1W9. This refers to -
the Safety System Functional Inspection (SSFI)' performed by I personnel of your office during the period of May 1, through May 26, 1989 at the James A. FitzPatrick Nuclear Power Plant.
I' Paragraph 4.7.2 of the SSFI report identified a concern regarding the absence of a formal preventive maintenance program for the 600V Class IE systems, components, and structures. The inspection report states that this concern was identified as a result of discussions with plant personnel. However, this concern was not identified at the inspection exit meeting nor was it identified to the Resident Manager during the course of the inspection. The specifics identified in Notice of Violation 4.1 were the only preventive maintenance issues discussed.
The Authority established a formal preventive maintenance program in 1984, which includes 600V Class IE circuit breakers, motor i starters, disconnect switches, transformers, switchgear housings, M i 1 I
8910020193 890921 PDR 0 ADOCK 05000333 PNU .
a-United States Nuclear-Regulatory Commission September 21, 1989 ATTN: Document Control Desk JAFP 89-0689
SUBJECT:
JAMES A FITZPATRICK NUCLEAR Page POWER PLANT DOCKET NO. 50-333 INSPECTION NO. 89-80 i
L and buswork. This program tracks, records, verifies, and clearly
' defines the maintenance intervals and required activities per the
-applicable sections of the relevant plant procedures. Since 1984, over.500 preventive maintenance work activities have been performed on the 600V equipment under this program.
Very truly yours, WILLIAM FERN D Z II WF:DAR:bnr DISTRIBUTION:
Records Management - WPO Director of BWR Licensing NRC Resident Inspector NRCI No. 89-80 File Document Control Center NRC Region 1 Office Attn: Bruce A. Roger, Acting Director Division of Reactor Safety i
CERTIFIED MAIL - RETURN RECEIPT REQUESTED l
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ENCLOSURE (1) TO JAFP-89-0689 L Response to Notice of Violation 1.0 i NOTICE OF VIOLATION As a result of the inspection conducted on May 1 through May 26, 1989, and in accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions," 10 CFR Part 2, Appendix C, 53 Fed. Reg..40019 (October 13, 1988),
the following violation was identified:
- 1. a. 10CFR 50, Appendix A, Criterion 10 (1967 issue),
Containment, requires that "The containment structure shall be designed to sustain the initial effect of gross equipment failure, such as a loss of coolant boundary break, without loss of required integrity and, together with the other engineered safety features as may be necessary, to retain'for as long as the situation requires the functional capability to protect the public."
b.- 10CFR 50, Appendix A, Criterion 40 (1967 issue),
Missile Protection, requires that " Protection for engineered safety features (such as the containment and its appurtenances) shall be provided against dynamic effects and missiles that might result from plant equipment failures."
- c. 10CFR 50, Appendix A, Criterion 42 (1967 issue),
Engineered Safety. Features Performance Capability, requires that " Engineered safety features shall be designed so that the capability of each component and system to perform its required function is not impaired by the effects of a loss-of-coolant-accident."
- d. 10CFR 50, Appendix A, Criterion 56 (issued as amended October 27, 1978), Primary Containment Isolation, requires that "Each line that connects ;
directly to the containment atmosphere (at any time, particularly during a LOCA) and penetrates primary reactor containment shall be provided with containment isolation valves. Isolation valves outside containment shall be located as close to the containment as practical and upon loss of actuating power, automatic isolation valves shall be designed to take the position that provides greater safety."
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m ENCLOSURE,(1) TO JAFP-89-0689 Response to' Notice of Violation 11.0..
(Continued)
- The reactor building closed loop cooling water'systemLpiping
inside containment.and the-components supplied byLthis-
- piping form aLelosed system inside: containment during. normal
. operation that is an extension of the~ reactor containment boundary. Contrary to'the requirements of a', b,'and c o .above~, the original plant ~ design did not provide for the protection of this' piping and. equipment from the dynamic effects of a' loss-of-coolant-accident. Therefore, the system may not remain; closed followingia. loss-of-coolant.
accident.
. Contrary to criterion 56-(1978), in the modifications made t'o'the system in 1984, the isolation valves located ~outside containment were not made automatics-the isolation valves located outside containment were.not located as close to the containment as practical; and'the isolation valves'were'not-designed to take the position that provides greater safety (closed) upon loss of actuating power. .
This~is a. Severity Level IV violation'(Supplement 1).
' RESPONSE TO NOTICE OF VIOLATION The" Authority disagrees with the Notice of Violation. The
. prevalent. interpretation of a " closed system inside containment" when FitzPatrick was designed,' constructed, and licensed was' simply a' continuous physical boundary that separated the process fluid from the' containment atmosphere, and that missile protection of that boundary was not-implicit in this definition. Quotations from FitzPatrick's PSAR, FSAR, SER, correspondence between the. Authority and NRC, and'the Statement of Consideration that accompanied the 1971 GDC define the NRC's, the Authority's and the nuclear industty's interpretation of GDC 55, 56 and 57'in the period leading up to, and forming the basis for, the issuance of the FitzPatrick operating license (OL).
This violation should be withdrawn for the following reasons:
- 1. The 1967 draft of the General Design Criteria (GDC) quoted in parts la, Ib and Ic of the hotice of Violation were never incorporated into Title 10 of the Code of Federal Regulations (CFR). The presence of draft guidance in the public record is not an adequate basis for the fasuance of a violation. The draft GDC are act a legally binding requirement. Furthermore, the Authority did not. commit to use the GDC in the design of Page _ _ _ _ - _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ - - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _
ENCLOSURE (1) TO JAFP-89-0689 Response to Notice of Violation 1.0 (Continued)
FitzPatrick. Comments and suggestions on the proposed rule-making, and subsequent developments in the technology and licensing process resulted in a revised GDC which was not published in the Federal Register until February of 1971.
The December 18, 1972 letter referred to in the inspection report does not bear on this issue since that letter limits itself to pipe breaks outside containment. The violation was based on a hypothetical pipe break inside containment.
ii. The Federal Register Notice accompanying the 1971 Appendix A clearly states that the principal design criteria (based on the GDC in Appendix A),
established by the applicant and approved by the Commission will be incorporated by reference in the Construction Permit (CP). The 1971 Appendix A to 10 CFR 50 was not effective until almost one year after the FitzPatrick CP was issued by the NRC and is not entirely applicable to FitzPatrick.
In any event, the draft GDC, which are the basis of the Notice of Violation, are not applicable to FitzPatrick, iii. The interpretations of GDCs 10, 40, 42 and 56 delineated in the Notice of Violation are new interpretations that significantly differ from the FitzPatrick design basis. As evidenced by correspondence between the Authority and AEC, and by statements in the SARs and SER, dynamic effects such as pipe whip and missile protection were considered to the extent practical given that the issue of missile protection was in its earliest stages of development. The missile protection criteria, unique to FitzPatrick, were agreed to by the AEC staff during the OL review, as evidenced by provisions in the staff's SERs (see discussion below). These criteria r2present the appropriate principal design criterie against which any violations should be judged. The Statements of Consideration accompanying the 1971 GDC make the developmental nature of this issue very clear calling missile protection an important safety consideration "not sufficiently developed and uniformly applied in the licensing process to warrant inclusion in the criteria at this time."
A footnote to the Appendix A definition of loss of Page _ _ - _ - ___ _______ -_ ______ ________ -_----__-__ _-_ _ _ -_-_-_-__ - -
L ENCLOSURE (1) TO JAFP-89-0689 Response to Notice of Violation 1.0 (Continued) coolant-accident states "Further details relating to'the type, size, and orientation of postulated breaks in specific components of the reactor coolant pressure boundary are under-development."
For FitzPatrick, compensatory measures include an augmented inservice inspection program for high-stress piping joints. Technical Specifi-cations require 100%, rather than 25%, inspection during ecch period. These measures were accepted by the staff as satisfactorily addressing dynamic effects, iv. With respect to GDC 42, 1967 version (Engineered Safety Features Performance Capability), the term
" engineered safety feature" was consciously eliminated from the approved 1971 issuance'of Appendix A and the requirements for engineered safety features " incorporated in the criteria for individual systems" (36 FR 3255). The 1971 GDC did not include specific criteria for systems thet provide containment cooling. Therefore, this GDC and the term ESF was inappropriately applied to the Reactor Building Closed Loop Cooling Water System (RBCLCWS).
- v. The leakage paths postulated in the inspection report are highly improbable. For the hypothetical leakage paths to develop, the combined probabilities of a LOCA, pipe ship severing RBCLCW piping, and failure of control room operators to recognize the existence of this pathway and close the RBCLCW valves make this an incredible scenario. Instrumentation currently installed in the plant would allow operators to determine that RBCLCW was the source of the release and isolate the system using the existing remote manual valves. (See discussion of how operators can detect a RBCLCW/ESW line break inside containment for further details.)
vi. In Appendix H of Supplement No. I to the FitzPatrick FSAR, the Authority compared the FitzPatrick design basis with the six groups of 1971 GDC. Appendix H, Table H-5, Note V-4, clarifies how GDC 56 was applied to FitzPatrick.
The quote is reprinted below.
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ENCLOSURE-(1) TO JAFP-89-0689 an . .
. Resp'onse-to Noticelof. Violation'1.0 j ." (Continued)_
f V-4. "The-phrase andLupo'n: loss of-actuating "E
m power,LautomaticMisolation~ valves 1shall be designed to takeithe~ position that provides greater safety' shouldcbe. changed to read and no single failure of. actuating power shall prevent the intended' safety function of the valve.' This more clearly. states the apparent intent of these' criteria."
.vii. This same note appears in Table 116.6-5-(note V-4) of.the Updated FitzPatrick FSAR. The existing remote-manual, air-operated, non-automatic RBCLCW
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containment isolation valves satisfy this-alternate criteria. The valve operators are equipped with~ air accumulators to assure that-they can 3e reliably operated in the event of a loss of actuating power.
The position that valve closure' represents "the.
position that provides greater safety" is based on engineering judgement and- is not supported by l analysis. The existing design considers thet operation (both opening and closing) to be the-appropriate "intendedJsafety. function," rather than closure alone. For any accident other than a design basis LOCA, the availability of-these lines to cool.the containment and equipment inside the drywell would:significantly mitigate the effects of the accident.
iix. The concept of." leak before break" has been accepted by the Commission ^and recently incorporated in a change to GDC.4 (52 FR 41288, October 27, 1987). FitzPatrick's' design basis for postulated high energy line breaks is partially based on this concept and-is conservative, ix. The NRC was fully aware-of the RBCLCW containment isolation valw;s arrangement and-agreed that a single remote-manual isolation valve satisfies the FitzPatrick design basis as delineated in the FSAR. RBCLCW containment isolation valves were upgraded during the 1983 refueling outage in response to an NRC Notice of Deviation (N0D) issued with Inspection 50-333/81-07. The NOD states:
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. ENCLOSURE-(1) TO;JAFP-S9-0689 L
b . Response to Notice of' Violation 1.0'
~
p.
,(Continued) e g " Contrary to . the : licensee's commitments in J Sections 5.2.3'.5 and 7.3.4.3 of'the FSAR'the five Reactor Building: Closed Loop Cooling' p/"* '
. Water Systemfcontainment effluent lines.have-
'only single manual isolation. valves;out' side.
' containment-instead of one valve closing automatically by arocess. action (i.e.,
. reverse flow) or iy remote manual operation-
-from-the control room." (emphasis added)
In it's-August 24, 1981 response,.the' Authority committed-to modify.the plant.to agree-with the h FSAR. In' Inspection. 50-333/85-19, the associated modification was reviewed.and the deviation closed
'without questions.
4 Attached are. copies of excerpts from the: documents referenced in this response. .Many of.these same excerpts-were provided:to.the NRC inspectors during the inspection.
All werecprovided to the inspection team'during or' shortly '
after-the' inspection.
Many additional documents from the FitzPatrick.docketiand'
.public1recordiare available to: support the' positions-delineated'above. .
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-ENCLOSURE : (1) 1TO JAFP-89-0689
,g. Response'to Notice'of Violation-1.0-Fly i (Continued)
- 'i F.
-ArBrief History of theLDevelopment of FitzPatrick' L -
-Design Basis for Missile Protection j
_The following. summaries are offered to support the-position Lthat missile protection criteria, analysis. techniques, and L : postulated missiles were just developing.during the-FitzPatrick CP review and_that any comparison to current staff! criteria'is inappropriate. o0f-particular significance ris~the' fact that the AEC staff, in the FitzPatrick SER, i
- accepted'as compensatory actions, 100% inspection of piping !
weld.in lieu of main steam and feedwater line pipe' whip restraints.-
- 1. AEC FitzPatrick Safety Evaluation Report, Supplement 2, dated October 1974' ;
Section 4.4 states:
, i "In Supplement 20 to the FSAR, PASNY presented the ~
l results of.a study to determine the. effects of missile generation-from recirculation pump overspeed following- 1 a LOCA... The staff concludes that~the probability of this'potentially damaging event is sufficiently 1 We conclude that the applicant's 'l
-small...
not'to add pipe restraints is acceptable." proposal... J
- 2. ;AEC.FitzPatrick Safety Evaluation Report, dated !
November 1972 i Section'5.2.1, " Primary Containment - Design," p. 5-1 of the FitzPatrick Safety Evaluation Report states:
"... critical penetrations have been provided'with ]
restraints and auxiliary stops to limit pi7e movement j i
and prevent tailure of the containment... 3ased on our review of the information contained in this application and similar designs we conclude that the primary-containment design basis is acceptable." (emphasis added) _j Section 5.2.2, " Primary Containment - Missile and Pipe Whip Protection," p. 5-4, states: ;
"Several locations on the main steam lines and feedwater lines are not restrained to prevent pipe whip 1 in the event of pipe failure at these locations. The 1 applicant has stated that the physical layout within i the drywell precludes restraints at these points...
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@ . ' ENCLOSURE.(1)'T0"JAFP-89-0689 l ResponseEto: Notice of Violation 1.0 l (Continued). ~
1 1
Y The applicant-has identified the unrestrained'high 1
-stress areas in these lines where breaks could result
, , ai .in pipe whip such that the pipe could impact the 1 ". '
primary containment. wall...- In addition, he has agreed to perform augmented. inservice' inspection of these weld locations during each inspection period... The . .
augmented inspection will beiset forth in-the Technical i Specifications and will call for 100% rather than 25%-
inspection during each period. ,
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The applicant has.also considered the' effects of pipe whipson.the~ emergency core cooling systems'. The systems are" redundant'and physically separated such thatLa ruptured pipe could impact and affect only one 1 redundant ECCS... ;
The applicant has' considered the effect of missiles... i ahd concludes that no missile would have sufficient energy.to penetrate the containment.- !
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_ We-conclude that'the applicant has provided adequate !
- measures'to protect against'the occurrence-and consequences of missiles and pipe whip." (emphasis j added) j In accordance~with these statements, Technical Specification'4.6.F.2, " Structural Integrity," requires an augmented ISI program for high-stress piping joints in-the main steam and feedwater lines. All welds on lines greater than 4 inches in diameter and ,
-unrestrained against pipe whip are inspected during i every inspection interval. 1 3.- Section 5.3.2, " Containment Isolation System," p. 5.3-1 j of the FitzPatrick PSAR-states: :
" Lines penetrating and entering a closed system within
~the containment, such as the closed cooling water ,
system lines,.are provided with at least one check l
valve located outside the containment. On the exit
-lines for this type system, a manually operated gate valve is required for isolation and is located outside !
the containment." (emphasis added) i .
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k ENCLOSURE (1.) T0.JAFP-89-0689 Response to Notice of Violation 1.0 (Continued) 4.' NYPA' letter, J.P. Bayne to T.A. Ippolitto dated January 7, 1982 (JPN-82-005) regarding NUREG-0737 item II.E.4.2, Cont.ainment Isolation Dependability.
In'this letter, the Authority described modifications to install power-operated isolation
-valves with. remote manual actuation on the RBCLCWS l- system. In that report, the Authority stated "The existing isolation capability of the. systems
-is as follows... The Reactor Building Closed Cooling Water lines do not communicate with the reactor coolant system or the containment atmosphere, are Seismic Class I inside containment,..."
- 5. NRC Inspection 50-333/81-07, dated July 31, 1981 NRC inspection 50-333/81-07 recognized that the RBCLCW system did not fully meet the 1971 GDC. The inspector reviewed the Authority's FSAR commitments on this matter. No Notice of Violation or open item was issued as a result of this inspection.
Inspection 81-07 did identify a deviation from the FSAR regarding RBCLCW effluent isolation valves. In its response, (R.J. Pasternak to B.H. Grier, dated August 24, 1981, JAFP-81-0871) the Authority committed to install remote manual containment isolation valves on the RBCLCW system influent and effluent lines, stating that these valves are " desirable for operational considerations and for additional safety margin."
During NRC inspection 50-333/85-19, dated August 12, 1985, the modification package prepared by the Authority for the installation of remote manual valves on RBCLCW lines was reviewed. The inspector closed the open item having no questions.
- 6. Letter, Joseph M. Hendrie (ACRS) to Glenn T. Seaborg (USAEC Chairman) dated January 27, 1970 Mr. Hendrie notes that the Authority "...is studying design modifications to prevent violation of the containment by pipe whipping and generation of missiles in the enlikely event of a failure of the primary system piping. The Committee believes that such design modifications as are practical should be implemented..." (emphasis added)
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ENCLOSUREL(1) TOIJAFP-89-0689 Response'to Notice ~of Violation-1~.0:
s o
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' (Continued) kl>
@ ' 7. LLetter, Peter AJ Norris..(AEC, Division of-Reactor Licensing) to;Mr. Asa GeorgeL(NYPA) dated October 19, i > - ' 1970.-
J -
Mr.:Nbrris'summarizesLa meeting held September ~24, 1970
.~ to discuss design. criteria for pipe whip protection.
.Mr. Norris' presents;aistaff position. stating:
'"You indicated lthat your criteria forJpipe whip
. protet tion. were/ stated. in Supplement?l3 - to your i application;...LUnless'the;results of a rigorous .
analysis, including appropriate experimental.
justification,.can.be provided-to support your.
' statement,7your proposed criteria for pipe whip k jprotection.for engineered safety features-should be
, revised.'with regard _to the emergency core cooling h _ systems:which are to be protected from'the effects of
- pipe whip."-
4 .The' design criteria added as part of Supplement.13 are
. quoted.in Section 16.3.2.2.1 of the Updated FitzPatrick FSAR.
8.- Amendment ~No. 13 to Application,-LeBoeuf, Lamb, Leiby &
MacRae to' Peter-A. Norris (AEC) dated January 23, 1970.
The' Authority responded to an AEC' question: " Identify and: discuss the supplementary . protection means such as the. inservice inspection-program, the use of energy absorbing structures,'and other,' to. assure that plant
> , : safety will not be' adversely affected by breaks at unrestrained high stress areas..."
In its response, the representatives for'the Authority stated: "
Since there crc no areas at which the specified criteria are exceeded, plant safety would not be compromised even if there were no means of supplementary protection. However, the lines iu E question will be subject to an inservice inspection program in accordance.with the provision of Appendix F of the F3AR."
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l ~ n.- , ENCLOSURE ~ (1) TO JAFP 0689
. Response"toLNoticeDof Violation 1.0 (Continued) !
}
Supplement ~13 1 also'added the,following statements to-g 1 ,
'the;FitzPatrickLPSAR:
"The Authority recognizes the~AEC. concern over 4 /
- postulated pipe-ruptures and subsequent pipe-movement, within the primary containment. To-meetuthis concern
>:the following design' criteria-were established for the
- FitzPatrick Planti
- 1. . Where: p'/acticable, considering available
, structurally _ adequate supports,-available-space,-
the'need for inservice inspection and.the need for 1, Maintenance, pipes will be restrained against pipe movement-resulting from postulated. pipe rupture so
-thattthe integrity of.the primary. containment will
'be maintained. ;
i 2; . . Wherever it is not practicable to prevent pipe 1
-movement by' restraints, the integrity-of the- 'l primary containment'will be maintained with an !
energy-absorbi'g n material' applied to the primary containment at areas subject to_the impact of the ruptured pipe.
3.- Pipes.will'be restrained, if shown necessary by {
analysis to prevent the loss of the entire j
, emergency cooling function. Single active .. j component failure will be included in the analysis
. . in. addition to those failures resulting from i excessive pipe movement. )
At least one core spray subsystem or one LPCI' loop !
utilizing two RHR pumps for core' cooling will be j available for the limitation of core temperature '
rise. In any event, thereafter, sufficient flow from one RHR pump, in conjunction with one RHR heat exchanger and two RHR service water pumps, will-be available for long term containment
, cooling."
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. ENCLOSURE 1-('l)'TO JAFP-89-0689
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Response-to Notice'of' Violation 1.0-
.(Continued)
- 9. - M.C' FitzPatrick Safety Evaluation Report, ' dated Novemberi1972 Section 5.2.3, " Containment' Isolation'," p. 5-5 of the FitzPatrick Safety Evaluation states:
"The. numbers, types and locations:of these-(containment isolation)' valves depend on the manner in'which the line penetrate the' reactor vessel and containment.
WhereLnecessary,.the valves are equipped with operators and,close-~ automatically when sensors detect certain:.
accident or fault conditions.
The' consequences'of postulate'd pipe failures both inside and outside.of:the containment-have been evaluated and are described in Section.10. -The isolation valves and their control systems'have been reviewed'to assure that no single-failure can. result in a loss of containment integrity."
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ENCLOSURE (1) TO JAFP-89-0689 e Response to Notice of Violation 1.0 (Continued)
How Operators Can Detect a RBCLCW/ESW Line Break Inside Containment Control Room Operators would be able to detect an RBCLCW/ESW
-line break inside containment with the existing instrumentation.
Existing RBCLCW instrumentation is described in Section 9.5.3 of the updated FitzPatrick Final Safety Analysis Report:
"A radiation monitor is located in the closed loop cooling water return header to detect high radiation levels in the closed loop...
The following conditions alarm in the Control Room:
- 1. Head tank low level
- 2. Head tank-high level
- 3. Pump discharge head low pressure
- 4. High radiation level
- 5. High and low temperature at RBCLCW heat exchanger discharge."
Either a RBCLCWS effluent low-flow alarm or a high rrdiation alarm could be indicative of a breach of the RBCLCW system boundary. Low head tank level alarms could identify a potential leakage pathway through the head tank vent. High heat exchanger discharge temperature could also be indicative of a RBCLCW line break.
During a design basis LOCA, elevated radiation levels in the circulating water system discharge canal could also be indicative of a break. Only two fluid systems, potentially radioactive because they interface with the reactor coolant or drywell atmosphere, operate following a postulated LOCA:
the Residual Heat Removal Service Water System (RHRSW), and the Emergency Service Water (ESW) or RBCLCW System. Both reject heat to the circulating water system and Lake Ontario. Any increase in radiation level would have to be the result of an RBCLCW/ESW or RHRSW System break.
Using indicators, operators can identify an RBCLCW line break and isolate the RBCLCW using the existing remote-manual containment isolation valves from the control room.
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ENCLOSURE (1) TO JAFP-89-0689 Responseito Notice of Violation 1.0 (Continued)
CONCLUSION The basie for the alleged violation are new requirements (backfits as defined in 10CFR 50.109) that go significantly beyond thu. original FitzPatrick design basis. Specifically the violation is based on the application of draft versions of the General Design Criteria that are not part of the
' design basis of the fac'lity.
Furthermore, the prevalent interpretation of a " closed system inside containment" when FitzPatrick was designed, constructed, and licensed was a continuous physical boundary that separated the process fluid from the containment atmosphere. Missile protection of that boundary was not implicit in this definition. Quotations from FitzPatrick's PSAR, FSAR, SER, correspondence between the Authority and NRC, and the Statement of Consideration that accompanied the 197.1 GDC define the NRC's, the Authority's and the nuclear industry's interpretation of GDC 55, 56, and 57 in the period leading up to, and forming the basis for, the issuance of the FitzPatrick 01 .
Imposition of criteria or requirements beyond the plant's design basis is an inappropriate basis for enforcement action and would constitute a backfit under the provisions of 10CFR 50.109.
Accordingly, the violation should be withdrawn. The Authority further believes that this issue should have been
, identified as backfit in accordance with NRC Staff Manual Chapter 0514, Section 042, before being transmitted to the Authority.
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POWER AUTHORITY OF THE STATE OF NEW YORK l 10 COLUMous CIRCLE NEW YORK. N. Y. loo 19 n 3 I212) 397-6200 c E oa c,E, p ,7 ,
OPE R A14NG OF F IC$ m TaUSTEES JOHN W. DOSTON JOHN $ DYSON- NaNIor noctownse
"""'" **'"' ""'"C' [,
. JOSEPH a, SCHMIEDER l GEoaot L .NGALLs VIC E C.wa s s ee & N P El of 7 CMIEF
. 8CH Aa D M. . FLYNN ao.E.T i MiLLoNu - January 7, 1982 LL*.'E^E.....,
ratoEasex a cLAnx- JPN-82-5 ' r*0'l. ""'"'
THOM AS a FaEY
.".'L" L ;"/ *!'"'
Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Mr. Thomas A. Ippolito, Chief Operating Reactors Branch No. 2 Division of Licensing
Subject:
James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Containment Isolation Dependability NUREG-0737, Item II.E.4.2
References:
- 1. NRC letter, D.G. Eisenhut to all Licensees q of Operating Plants, dated October 31, 1980
- 3. PASNY letter, R.J. Pasternak to Boyce H. Grier
,,s (JAFP-81-0871) dated August 24, 1981 b- 4. PASNY letter, J.P. Bayne to T.A. Ippolito (JPN-81-49) dated July 7, 1981
Dear Sir:
The Power Authority has completed a comprehensive review of the containment isolation dependability of the James A.
FitzPatrick Nuclear Power Plant, as requested in Reference 1.
The enclosed report documents this review and is submitted in accordance with the commitment made by the Authority in Reference 2.
The review identified the following systems which do not fully meet the NUREG-0737 acceptance criteria: Reactor Water Cleanup; Traversing Incore Probes; Recirculation Pump Mini-Purge; Leak Rate Analyzer; Reactor Building Closed Loop Cooling Water; and Contain-The Power Authority intends to achieve full ment Vent and Purge.
compliance by modifying these systems as described below.
The Reactor Water Cleanup, Traversing Incore Probes and Recirculation Pump Mini-Purge Systems will be modified by the addition of'The automatic Authorityisolation will perform valves an with diverse evaluation engineering actuation and signals.
an assessment of equipment availability and develop a schedule for the completion of these modifications. This schedule will be submitted by February 2, 1982.
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The Reactor Building Closed Loop Cooling Water System has f several return linesInfrom the containment accordance with the which havecommitments Authority's one manual isolation valve.
in Reference 3, a schedule for the installation of power operated valves with remote manual actuation will be submitted by February 2, 1982.
The Containment Vent and Purge Valves do not isolate on high radiation in the drywell. The Power Authority intends to provide a drywell high radiation isolation signal to these valves in ac-cordance with the commitment in Reference 4. Dependent upon material availability, this modification will be completed during the current refueling outage.
The Leak Rate Analyzer System does not comply with the NUREG-0737 criteria which does not permit ganged reopening of isolation valves. Dependent upon material availability, the Power Author-ity will modify this system to eliminate ganged reopening, during the current refueling outage.
"1 The Power Authority considers theseHowever, modifications an addition the Authority con-
to the margin of safety in the plant.
I siders the existing design to be adequate to assure safe operation
, ;;; until the modifications are complete. Each of these systems has 4L- an existing isolation capability, which will be described below.
Any leakage from these systems into the Reactor Building would be In addition,
+ processed by the Standby Gas Treatment System.
reactor coolant leaking past the existing isoJation valves can be replaced by the HPCI system.
The existing isolation capability of the systems is as fol-lows:
A. The Reactor Water Cleanup return is isolated by re-dundant check valves. In addition a motor operated valve capable of remote manual operation from the control room provides further isolation capability.
B. The Traversing Incore Probe purge line does not com-municate with either the reactor coolant system or the con-tainment atmosphere and is isolated by a check valve.
C. The Recirculation Pump Mini-Purge lines are isolated by two 3/4 inch check valves in series and are comparable in size to instrument linec.
a D. The Reactor Building Closed Cooling Water lines do not communicate with the reactor coolant system or the contain-ment atmosphere, are Seismic Class 1 inside In containment, and c-,
lA_) addition, the
~ are equiped with manual isolation valves.
reactor operators will be reminded that these lines do not automatically isolate.
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The enclosed report supercedes the Power Authority's pre-vious submittals in response to NUREG-0737 Item II.E.4.2.
If you have any further questions, please do not hesitate to contact us.
Very truly yours,
.h
- 3. A a nior Vice esident Nuclear Generation cc: Mr. J. Linville Resident Inspector U. S. Nuclear Reactor Regulation (j P. O. Box 136 Lycoming, New York 13093 Mr. Ron Barton
~~' United Engineers & Constructors, Inc.
30 S. 17th Street Philadelphia, PA 19101
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.....- 3_3 with Section III of the ASME Boiler and Pressure Vessel Code, maximum . :$ r dryvell pressures up to 62 psig are permissible for this ' design.- [
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. Combinations of live, dead, and seismic loads in conjunction with l
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- thermal streases have been considered in the design analysis. ' ,i
. design also considered the jet ' f orces that might. act on the contain- lJ*ff I;>
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' ment consequent to a pipe severance. Adequate strength has been
- 3 provided to prevent failure of the containment vall as a result of
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provided with restraints and auxiliary stops .to limit pipe movement !
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p and prevent failure of the-containment. i The primary containment was designed to sustain the combination of loads resulting from the design basis loss-of-coolant accident, I earthquake, and the conventional live-and dead loads within the
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.- stress limits defined in Subsection III B of the ASME Boiler and Pressure Vessel Code 1968 and applicable addenda in effect r -
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as of June 1968. We find the design stress limits for the l primary containment system to be acceptable. I'
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Containment piping penetrations satisfy the criteria of the [
Lines connected '
ASME Boiler and Pressure Vessel Code noted above. ! J L
to the reactor coolant system incorporate a sleeve to extend the f E..
dryvell to the outer isolation valve, thus containing the effluent [
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. . . in the event of a pipe break. Hot lines which must sustain large thermal and mechanical stresses are designed with combinations of .
penetration sleeves and flued fittings. ..
t Based on our review of the information contained in this application and similar designs we conclude that the primary ;
containment design basis is acceptable. }
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S.2.2- Rissile and Pipe Whip Protection l
Several locations on the main steam lines and feedwater lines l are not restrained to prevent pipe whip in the event of pipe failure at these locations.- The applicant has stated that the physical. layout within the dryvell pr'ecludes restraints at these points. For all T
other lines and locations, restraints have been provided where ;
l a break could result in containment impact. The applicant has identified the unrestrained high stress areas in these lines where breaks could result in pipe, whip such that the pipe could impact the
'l . Primary containment vall. At those locations which are accessible !
I the applicant has provided 1-1/4 inch thick impact plates as sup- .
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plementary protection for the drywell. In addition, he has agreed i i
to perform augmented inservice inspection of these weld locations f during each inspection period. At the remainder of.these identified *
- I areas the physical layout precludes installation of impact places.
t Here, the applicant will perform augmented inservice inspection of
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N- ' augmented inspection will be set forth in the Technical Specifications
'.m' and will' eall f or 100% rather than 25% inspection during each period.
The applicant has also considered the ef fects of pipe whip on the
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emergency ' core cooling systems. The systems are redundant and
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physically separated such that a ruptured pipe could impact and v.;. .
' were shown to limit peak fuel clad temperature to 1370*F following the
.' .. most severe postulated break sequence.
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' size from nuts and bolts to valve bonnets, and concludee that no e;. (
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i mics11e would have sufficient energy to penetrate the containment.
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- in addition, where possible, components are arranged so that the
.'.3 direction of flight of potential missiles is away from the contain-
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ment well.
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'- The ef fects of pipe whip and steam jet impingement on the shield s
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and vessel support structure resulting from a LOCA occurring within
- the sacrificial shield area were analyzed and found to be acceptable.
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We conclude that the applicant has provided adequate measures to protect against the occurrence and consequences of missiles and pipe
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whip.
', 5.2.3 Containment Isolation The ability to isolete the primary containment provides the N;-
necessary integrity between the coolant system pressure boundary,
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or the containment atmosphere, and the environs in the event of 4 ,
Isolation is accidents or other non-nominal conditions. * .
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The numbers, types and locations accomplished by means of _ valves.
of these valves in the various lines depend on the manner in which Where the lines penetrate the reactor vessel and the containment.
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necessary, the valves are equipped with operators and close j- . or fault
< . automatically when sensors detect certain accident o
conditions.
The consequenc,es of postulated pipe failures both inside and outside of the containment have been evaluated and are described The isolation valves and their control systems _,
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have been reviewed to assure that no single failure can result '"
in a loss of containment integrity. An exception exists in the case of' instrument lines connecting to the reactor coolant
' system which penetrate the containment and dead-end at 4
Such lines are instruments located in the reactor building.
provided with manually operated isolation valves and excess flow A break check valves, both of which are outside the containment.
in the line between the containment and the oucer check valve would result in blowdown directly into the reactor building.
The applicant has installed 1/4 inch diamator orifices in each G
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', e o these lines'insido.the' primary containment to prevent over-
- : pressurization'of thc.'rcactor building and -limit off sito doses to substantially below thc 10 CFR Part 100 va uer, in the event c . ^
4 of the postulated instrument line. break. Based on'our review of
,' the design .we ' conc 1'ude that the provisions f or instrument 1. ines
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penetrating the primary containment are adequate. and satisfy the- '
supplement to Safety Guide 11.
Leakage through the closed main steam line isolation valves fo115 wing aLpostula'ted.LOCA presently relics on thc. low leakage j
characteristic ~of the vaives.- The acceptability of present
" t' leakage _ limits and the need for an auxiliary scaling system
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- are under study by the staff. :There is nothing in the existing
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design which would preclude incorporation;of an additional. sealing *
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- feature'if such is determined to be necessary.- The applicant t .
- will' continue to study developmen'es in this area.. .
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Based on'our review we conclude that'the primary contain-
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ment isolation provisions are adequate.
5.2.4 --Leakage Testing Program 3 .
Leakage testing of the reactor primary containment and E..
. associated systems is intended to provide initial and periodic
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. verification of the leaktight integrity of the containment.
the The applicant hes stated in Amendment Nos. 4 and 5 that
[h~ primary reactor containment and its components have been designed ,
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- at the calculated peak pressure.~ Penetrations,-including-t ,
personnel and equipment,hatthes.and airlocks,i and isolation valves, -
P have been designed with the capability of being individually ;
.c leak tested at calculated peak pressure. l
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- We conclude that the containment system will permit contain- .
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ment leakage rate testing in compliance with the AEC proposed I
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~ " Reactor Containment Leakage Testing for Water Cooled Power e i- ;
V . Reactors," 10 CFR 5 50.54(o), Appendix J, and therefore is acceptable.
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- In' addition to agreeing. to meet the requirements "of proposed
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- Appendix J, PASNY has agreed to perform a leak test of dryvell 3
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.' " to suppression-chamber piping, headers, downcomers and vacuum )
' breaker. valves at each refueling outage. They will also determine criteria and will,be
,' receptable bypass leakage limits and other test
(* required to perform frequent surveillance testing of the vacuum breakers. We have not' completed our review of the details of the i test and surveillance program. However, the applicant has indicated l y , j his intention to base it on the recently approved Browns Ferry leak s
check program. We find this commitment acceptable pending i
', completion of our review.
5.3 secondary Containment The reactor building, together with the Standby Gas
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Treatment System (SGTS) and the main stack, form the sero .dary .
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I ggMENT XSOLATION SYSTEM Bases l1 system is designed to '
design bases The containment isolation during integrity any ts.
containment all piping lines
'j, isolate The A
Provisions.are made toor open into the primary
' dependent containment.
upon
' penetrate.and mannerthe of ' isolation provided prepotential consequences ce, of as well- as p ice P[
failure. for L-ting parts which complete the closure j r-
- [ Communi ca pisary containment have an integrity at least equal'l to ifT j>
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. rr ti l Description _ l containment and l ,
'. . Pipes whichthe penetrate the primaryreactor primary on outlet system are provided ; i
? connect to and two valves i
'two i onevalves on inlet practicable linesvalve to is located the inside containment be the containm I outside as close as
~.: On inlet lines, one or both of the two valves mayIsolation
- i - ck valve. actuation
.u, ll signals and are capable of remote heat manualcontrol ro
+ the lines for core spray, residual because
_,be Provided on coolant injection l and high-pressure
- tion of thes6 systems is essential following a loss-of- ,1 l
t accident. into
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Lines which penetrate the containment and same open criteria f !'
dry well or suppression chamber have' thethat the the pipes or ducts i
tabove except series located outside suppression 4 Power-operated valves into the above are are Exceptions which A lainment. relief lines to c.t.mo sphere ,
r vacuum ided with one self-actuated and one power-operated valve
. series.
,' and entering a closed system E. '
penetrating cooling water g ye ,-
the closed 9'" Lines l the containment, such as f, J,
- provided with On the exit lines for at least one check valve located this type a
i ide.arethe containment. gate valve is required for a manually operated a
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containment, if isolation and is located outside the systems used for maintenance, such as service air, se g water, and breathing air, at least one manual valve that I \ normally closed is provided outside the containment. :
h Instrumentation piping connecting to the rea 'f8 i
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- primary systems which leaves the primary containment and '
y / dead-ended at instruments located in the reactor building provided with an excess-flow check valve and at least ' [l
' manual valve outside the containment.
Each traveling incore probe (TIP) system guide l is provided with an isolation valve which cl J
- " l automatically upon receipt of the proper signal and a the TIP cable and fission chamber have been retracted.I isolation valve, an additional or ba series with this isolation shear valve is included. Both valves are loca outside the dry well. The function of the shear valve is '
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assure integrity of the containment even in the unl event that the isolation valve should fail to close or '
chamber drive cable should f ail to retract if it shouLd contal extended in the guide tube during the time that isolation is required. This valve is designed to shear cable and seal the guide tube, if necessary, upon a man initiated signal. Valve position (full onen or full cle of the automatic closing valves is indicated in the control room. The shear valve is an explosive type val with monitoring of each actuating circuit provided, notive power for the valves on reactor pr, system lines which require two valves derives physically independent sources to provide a high probab _
that no single component failure could interrupt i power to both closure devices. -}
All remote manually actuated isolation valvecon 8 the provided with limit switches which indicate in i
room whether the valve is open or closed.
i 5.3.3 Evaluation Since a rupture of a large line penetrating containment and connectina to the reactor coolant sy5t** iSDI"
/ take place within the containment, one of the two valves for that line is located within the contal sec0DO.
Additional reliability is provided by the located outside tne containment and as close as pract1 If a failure involves one valve, the second valV8 be available to function as the containment barrief*
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i physically separating the two valves the reliabill
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s 4 PASIR m
~ 2. - i en t, gficantly enhanced. The two valves in series are servi pided with independent power sources.
that gg
... Q g The TIP system isolation valves are normally
.; ged. When the TIP system cable is inserted, the valve of rea De selected tube automatically opens and the chamber and and ja suble are inserted. If closure of the valve is required lding'Is turing calibration, the isolation signal causes the cable to east asis .nsj retracted and the valve closes automatically on
'ssupletion of cable withdrawal.
ide t i- It is not- necessary or desirable that every clos'es isslation valve close simultaneously with a common isolation ad aftat . For example, if a process pipe were to rupture in ed.-4 dry well, it would be important to close all lines which r ba @ M. o to- the dry well and some effluent process lines locab I IIIli' Pen as the main steam lines. Ilowever, under these is'i i tions, it is essential that containment and' core fika) , analing systems be operable. For this reason, specific or .. d l Signals will be utilized for isolation of the various
- ould) s and safeguards systems.
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Reactor primary system isolation valve closure
.an _'. are such that for any design basis break, the coolant clos '
will not result in excessive off-site doses.
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. Tests and Inspection
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Valves, sensors, and other automatic devices 2 11 - al to the isolation of the containment are provided me ans to test periodically the functional performance equipment. Such tests include demonstration of operation, correct set point of sensors, proper speed es nses, and operability of fail-safe features. tieans on ided for measurement of leak rate across individual
. valves,
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