ML20039C797
| ML20039C797 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 12/08/1981 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | Rich Smith VERMONT YANKEE NUCLEAR POWER CORP. |
| References | |
| TASK-2.E.4.2, TASK-TM IEB-80-06, IEB-80-6, NUDOCS 8112300239 | |
| Download: ML20039C797 (33) | |
Text
y-Docket No. 50-271 DFC 8 1981 m
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f Mr. Robert L. Smith I
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Licensing Engineer 7
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Vermont Yankee Nuclear Power L;
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%a Framingham, MA 01701 3
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Dear Mr. Smith:
V In our letter of November 29, 1978 we identified the generic concerns of purging and venting of containments to all operatinn reactor licensees and requested your response to these conccens. Our review of your response was interrupted by the TMI accident and its demands on staff resources. Con-sequently, as you know, an Interim Position on containment purging and venting was transmitted to you on November 19, 1979. You were requested to imple-ment short-tem corrective actions to remain in effect pending completion of our longer-tem review of your response to our November 29, 1978 letter.
Over the past several nonths we and our contractors have been reviewing the responses to our November 1978 letter to close out our long-tem review of this rather complex issue. The components of this review are as follows:
1.
Conformance to Standard Review Plan Section 6.2.4 Revision 1 and Br_anch Technical Position CSB 6-4 Revision 1 These documents were provided as enclosures to our November 1978 letter.
2.
Valve Operability Although the Interim Position allowed blocking of the valves at partial-open positions, this is indeed an interim position. Earlier we requested a program demonstrating operability of the valves in accordince with our " Guidelines for Demonstrative Operability o.-
Purge and Vent Valves." These Guidelines were sent to you in our letter of September 27, 1979. There is an acceptable alternative which you may wish to consider in lieu of completing the valve qualification program for the large butterfly-type valves. This would be the installation of a fully qualified mini-purge system with valves 8 inches or smaller to bypass the larger valves.
Such a system change might prove more timely and more cost-effective.
The system would meet BTP CSB 6-4 item B.l.c.
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3.
Safety Actuation Sional Override This involves the review of safety actuation signal circuits to ensure that overriding of one safety actuation signal does not also cause the bypass of any other safety actuation signal.
4.
Containment Leakage Due to Seal Deterioration Position B.4 of the BTP CSB 6-4 requires that provisions be made to test the availability of the isolation function and the leaMage rate of the isolation valves in the vent and purge lines, individually, during reactor operations. But CSB 6-4 does not explain when or how these tests are to be performed. Enclosure 1 is an amplifica-tion of Position B.4 concerning these tests.
The status of our long-term review of the above items for the Vermont Yankee Nuclear Power Station is as follows:
1.
Conformance to Standard Review Plan Section 6.2.4 Revision 1 and Branch Technical Position CSB 6-4 Revision 1 This item is still under review. Since it appears that there may be some Aisunderstanding regarding the use of containment purge / vent valves, a restatement of salient features of the position as interpreted by the staff is provided in Enclosure 2 to assist you in understanding subsequent correspondence on this item from the staff. Additional information which we need to complete our review is shown in Enclosure 3., which is a draft SER for containment purging and venting during operation, is provided for your information. Please respond within 45 days of receipt of this letter.
2.
Valve Operability On March 3,1981 we transmitted questions concerning your subnittal on valve operability. Your May 21, 1981 response to our letter is still under review.
3.
Safety Actuation Signal Override We have completed our review. Enclosure 5 is our Safety Evaluation Report (SER) for this item. We believe your design should be modified to fully annunciate the overridden status of safety systems, as described in the enclosed SER (Enclosure S). Please inform us of your plans with respect to annunciation within 30 days of receipt of this letter. With this SER and your modification of ar:nunciation, the electrical override aspects of our long-term review of this generic task will be completed.
It is noted that a somewhat parallel review of engineered safety features reset is being carried nut in enninneti nn with 175 Rulletin 80-06.
That review will be tiled se_
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4.
Containment LeaNage Due to Seal Deterioration Provide a response to question 1 of Enclosure 3 within 45 days of receipt ef this letter.
In closing, you may have noted the similarity of this long-term generic issue with Item II.E.4.2 of NUREG-0737, THI Action Plan.
Except for Positions 5, 6 & 7 of Item II.E.4.2, the review of the remaining outstanding positions of Item II.E.4.2 will be completed by this purge and vent review. Our schedule of the purge and vent review agrees with the schedule for Item II.E.4.2.
Thus, your assistance in completing the outstanding purge and vent items, noted above, is necessary to complete Item II.E.4.2.
Although the Technical Specifications necessary to finalize the purge and vent part of Item II.E;4.2 are not completely finalized, a recently developed sample Technical Specifica-tion is provided for your consideration as Enclosure 6.
We request that you review existing Technical Specifications (TS) against the samole provided herein.
For any areas in which your existing TS needs expansion, you are requested to provide a TS change request within 30 days of receipt of this letter.
The request for infomation and changes to the Technical Specifications con-tained in this letter are specific to the Vemont Yankee Nuclear Power Station.
Therefore, OMB clearance is not required under P.L.96-511.
Please contact your NRC Project Manager should you have any questions.
Sincerely, wuuu.a sw.so g Thomas A. Ippolito, Chief Operating Reactors Branch #2 Division of Licensing
Enclosures:
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Mr. W. F. Conway Public Service Board President & Chief Operating Officer State of Vermont Vermont Yankee Nuclear Power Corp.
120 State Street 411 Western Avenue Montpelier, Vermont 05602 Drawer 2 West Brattleboro, Vermont 05301 W. P. Murphy, Plant Superintendent
-Vermont Yankee Nuclear Power Corp.
Mr. Louis H. Heider, V.P.
P.O. Box 157 Ve'rmont Yankee Nuclear Power Corpo.
Vernon, Vermont 05354 25 Research Drive
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Westboro, Massachusetts 01 581 Vermont Yankee' Decommissioning Alliance John A. Rit.scher, Esquire 5 State Street Rope.& Gray Box 1117 226 Frankline Street Montpelier, Vennont 05602 Boston, Massachusetts 02110 Brooks Memorial Library Honorable John J. Easton 224 Main Street Attorney General Brattleboro, Vermont 05301 State of Vermont 109 State Street Resident Inspector Montpelier, Vermont 05602
.c/o. U.S. NRC P.O. Box 176 Vermont Yankee Decommissioning Vernon, Vermont 05453 Alliance 53 Frost Street Brattleboro, Vermont 05301 Mr. E. W. Jackson Manager of Operations Vermont Yankee Nuclear Power Corp.
411 Western Avenue Drawer 2 West Brattleboro, Vermont 05301
' ' 'Raymond N. McCandless Vermont Div'ision of Occupational
& Radiological Health Administration Building 10 Baldwin Street Montpelier, Vermont 05602 New England Coalition on Nuclear Pollution Hill and Dale Farm R.D. 2, Box 223 Putney, Vermont 05346 l
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Enclosure.1
. PURGE / VENT VALVE LEAKAGE TESTS
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The long term resolution of Generic Issue B-24, " Containment Purging During Normal Plant Operation," includes ~, in part, the implemer.tation of Item B.4 of Branch Technic.a1 Position (BTP) CSB 6-4.
Item B.4 s ecifies that provisions should be made for leakage rate testing o.f the (p purge / vent system) isolation valves,, individually, during reactor operation. Although..
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Item B.4 does not address the testing frequency, Appendix J to.10 CFR Part ~
50 specifies a maximum test interval of 2 years.
As.a resiJ1t' of the numerous reports.on unsatisfactory performance of the
" resilient seats foi the isolation valves in containment purge.and -vent if nes (addressed in OIE Circular 77-11,' dated September 6,1977), Generic Issue B-20,' " Containment Leakage Due to Seal Deterioration," was established to evaluate the matter and establish an appropriate testing frequency for the isolation valves. Excessive idakage past'the~ resilient seats ~of isolation valves in purge / vent line.s is typically caused by severe environmental. con-
. Consequently,' the leakage test ditions and/or year due-to frequent use.
frequency foc these valves should be keyed to the occurrence of severe e'nviron-mental conditions and the use of the valves, rather than the current require-ments of 10 CFR,50, Appendix J...
It is recommended that the following provision be added to the Technical Specifications for the leak testing of purge / vent line isolation valves:
" Leakage integrity tests shall be performed on the containment isolation valves with resilient material seals in (a) active purge / vent systems (i.e., those which may be. operated during plant operating Modes 1 through 3) at least once every three months and (b) passive purge systems (i.e., those which must be administrative 1y controlled closed during reactor operating Modes 1 through 3) at least once every six months."
By way of clarification, the above proposed surveillance specification is predicated on our expectation that a plant would have a need to go to cold st)utdown several times a year. To cover the possibility that this may not occur, a maximum test interval of 6 months is specified. However, it is not our intent to require a plant to shutdown just to conduct the valve leakage integrity tests.
If licensees anticipate long duration power oper-ations with infrequent shutdown, then installation of a leak test connection This that is accessible frpm outside containment may be appropriate.
It will not be-will permit simultaneous testing of the redundant valves.
possible to satisfy, explicitly the guidance of Item B.4 of BTP CSB 6-4 (which states that valves should be tested individually), but at least some testing of the valves during raactor operation will be possible.
S
a It is intended that tS above proposed surveillance specification be. app, lied ll as passive purge lines: 1.ee,.the "to the active purge / vent lines, as we purge lines that are administrative 1y controlled closed during reactor oper-ating modes 1-4.
The reason for including the passive purge lines is that B-20 is concerned wtih the potential adverse effect of seasonal weather con-ditions on'the integrity of the isolation valves. Consequently, passive purge lines' must also b.e included in' the surveillance program.
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-The purpose of the leakage. integrity tests of the isolation valves in the
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containment purge and ' vent lines is to identify exce'ssive' degradation of Therefore, they need not be. conducted
.the resilient seats for these valves.
with, th'e precision pequired. for the. Type C isolation :v'alye _t.e.sts in~ 10 CFR Part 50, Append'x J. - These tests would ;be performed -in additfon.io.tho
, quantitative Type C tests required by Appendix J and would In view of the wide variety of valve types and seating materials, the :
acceptance criteria for such tests should be developed on a plant-specific
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i Standard Review Plan 6.2.4 Clarification of Valve Usage i
Staff Position 1.
Purging / venting should be minimized during reactor ' operation
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because the plant is inherently safer with closed purge / vent valves 1
e n:s (containment) than with open lines which require valvegaction to provide; conthinment.
(Serious consideration is being given to ult'imately requiring that future plants be designed such that purging 7 venting is not required during operation).
2.
Some purging / venting on current plants will be permitted provided that:
a) purging is needed and justified for safety purposes, and b) valves are judged by the staff to be both operable and reliable, and c) the estimated amount of radioactivity released during the time required to close the valve (s) following a LOCA either 1.
does not cause the total dose to exceed the 10 CFR Part 100 Guidelines; then a goal should be established which
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represents a limit on the annual hours of purging expected through each particular valve, or ii. causes the total dose to exceed the guideline values; then purging / venting shall be limited to 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> / year.:
3.
Purging / venting should not b6 permitted when valves are being used that are known to be not operable or reliable under tra'nsient r
. or accide'nt conditions.
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ADDITIONAL INFORMATION REQUIREC TO EVALUATE VERMONT YANKEE NUCLEAR POWER STATION FOR CONFORMANCE TO GUIDELINES OF BTP CSB 6-4 REVISION 1, " CONTAINMENT PURGING DURING NORMAL OPERATION" 1.
It is necessary to determine the seat leakage of the butterfly type isolation valves in the containment purge and vent lines.
Enclosure l' provides guidance in this respect.
If the results of current and past surveillance is believed to demonstrate operability of those valves, provide this information as justification for not increasing the surveillance requirements.
Otherwise, please propose Technical Specifications for periodically testing these valves.
2.
We and our contractor have reviewed the information you have provided with regard to debris protection, including your May 21, 1981 letter.
You have not provided sufficient information concerning the provisions made to ensure that isolation valve closure will not be prevented by debris which could potentially become entrained in the escaping air and steam.
We request that you provide debris screens for the purge supply and exhaust lines to deal with this problem or some other alternative which you support adequately.
If debris screens are provided, the debris screens should be designed to seismic Category I criteria and installed about one pipe-diameter away from the inner side of each inboard isolation valve..
The piping between the debris screen and the isolation valve should also be designeo to seismic Category I.
3.
It is the staff's recommendation that you commit to limiting the use of the purge / vent system to a specified annual time commensurate with plant operational safety needs.
DRAFT SAFETY EVALUATION REPORT FOR CONTAINMENT PURGING AND VENTING DURING NORMAL OPERATION OF THE VERMONT YANKEE NUCLEAR POWER STATION (Docket No. 50-271)
INTRODUCTION A number of events have occurred over the past several years, which directly relate to the practice of containment purging and venting during normal plant operation. These events have raised concerns relative to potential failures affecting the purge penetrations which could lead to degradation in containment integrity and for PWRs & degradation in ECCS performance.
By letter dated November 28, 1978 the Commission (NRC) requested all licensees of operating reactors to respond to certain generic concerns about containment purging or verting during normal plant operation. The generic concerns were twofold:
(1) Events had uccurred where licensees overrode or bypassed the safety actuation isolation signals to the containment isolation valves. These events were determined to be abnormal occurrences and were so characterized in our report to Congress in January 1979.
(2)
Recent licensing reviews have required tests or analyses to show that containment purge or vent valves would shut without degrading containment integrity during the dynamic loads of a design basis loss of coolant accident (DBA-LOCA).
i The NRC position of the November 1978 letter requested licensees to cease purging (or venting) of containment or limit purging (or venting) to an absolute minimum. Licensees who elected to purge (or vent) the containment were requested to demonstrate that the containment purge (or vent) system design met the criteria outlined in the NRC Standard Review Plan (SRP) 6.2.4.. Revision 1 and the associated Branch Technical Position (BTP) CSB 6-4, Revision 1.
III.
DISCUSSION AND EVALUATION The Purge / Ventilation System at Vermont Yankee utilizes various size isolation valves, the largest of which are 18, 8, and 6-inch butterfly-type valves.
The exhaust lines from the drywell and torus are connected to a common exhaust header which leads to the containment purge Exhaust Fan System and the Standby Gas Treatment (SBGT) system. These systems are individually isolated from the common exhaust header by 8-inch isolation valves.
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l An inch butterfly isolation valve, wtitch~ is bypassed bl a 3-inch line containing an isolation valve is provided to isolate the ex haust header from the drywell.
The exhaust header is isolated from the torus by in 18-inch butterfly isolation valve which 'is bypassed by a 3-inch line containing an isolation valve.
The purge supply lines to the drywell and torus are connected to a cocoon supply header which leads from the -
Reactor Building. An 18-inch butterfly isolation valve is provided for the j
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cocraon supply header.' An 18-inch butterfly isolation valve, which is by-
. assed by a,.1-inch line containing an isolation valve is provid.ed.to p
isolate the. supply-header from.the drywell.
The supply header is isolated from the torus by an 18-inch butterfly isolation valve which is bypassed by
'a 1-inch line containing an isolation valve..
7 _ _
1he licensee responded.to the NRC position letter;of November.
1978, by indicating that they planned to justify unlimited purging for only.
those valves / systems involved in the containment differential pressure o>eration, and that all other purge and, vent valves would be limited ~in t1eir use to 90. hours per year during power operation.
In a letter dated January 9,'1979. the licensee submitted a proposed Technical Specification change to reflect the incorporation of the necessary requirements to limit the use of selected containment vent.and purge valves during power opera-tion to 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per year.
The licensee indicated that normal on-line purging of the Vermont Yankee containment (drywell) is not a routine operation.
Typically, the only time on-line purging is conducted is just prior to an extended outage, e.g., refueling, when work is to ba done inside the drywell. This control-led purging allows drywell airborne activity levels to be reduced to ninimize radiation exposures and improve working conditions. The. necessity for Venwnt Yankce to continuously vent is related to the requirement to maintain a differential pressure between the drywell and the torus. Vernont Yankee has been operating with the torus /drywell differential pressure as a result of an " Order" issued by the KRC staff on February 13, 19,76.
It was concluded in that " Order" that maintaining a differential pressure of >1.7 psi between the drywell and torus. provides adequate assurance that contain-ment integrity will not be jeopardized by upward utions resulting from hydrodynamic forces from the most severe loss of coolant' accident.
The differential pressure control ' scheme, currently being employed by the plant, requires a positive pressure in the drywell (typically 1.8 to 1.1 psi) and essentially atmospheric pressure in ths torus.
In order to kee'p, the torus near atmospheric pressure, and thus maintain the required AP, a j
continuous vent is necessary to compensate for air in-leakage from the.
drywell.
The air in-leakage is primarily through the Containment Air Manitoring System and the torus vent header drain lines.
The continuous vent flew is directed through a 3. inch bypass valve (SB16-19-6B) and a downstream 8-inch isolation valve (5B16-19-6) to the SBGT system.
We have reviewed the licensee's justification for usage of the purge /vant sy. stem '
end find it to be acceptable.
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,In response to the ' issue of provisions to insure that isolation valve cloiure will not be prevented by debris which could potentially
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become entrained in escaping air and steam, the licensee stated that the
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prisiary containtnent atmosphere control piping penetrations are designed
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with a tail piece that projects into the containment.
The drpeell connec-tion is high up, far away from any inside equipment.
The torus connection is on top of the torus. No additional provisions
- were made in the oricinal design to prevent entrance of debris.
In. addition, the licensee concluded that in the small BWR containments, there is not much that could generate debris; therefore, it is not likely that debris would be carried into the The redundant back-up valves are located far from the containment, piping.
further reducing the probability of debris preventing containment isola tion.
Althouoh the licensee has provided some justification for not providing
' debris screens, se are not able to conclude that isolation valve closure will not be prevented by debris which could potentially become entrained in the escaping air and steam.
We therefore recommend that debris screens be provided for the purge supply and exhaust systems. The debris screens should be seismic Category I design and installed about one-pipe-diameter away from the inner side of each
. inboard isolation valve. The piping between the debris's.creen and the isolation valve should also be seismic Category I design.
III.
CONCLUSIONS r
We have reviewed the Vermont Yankee purge / vent system against the provisions of
~ BTP CSB 6-4 (Revision 1), " Containment Purging During Normal Plant Operations."
We conclude that the Verr.ont Yankee Plant containment purge / vent system is acceptable, subject to the licensee's inst'allation of debris screens as recommended above.
Also, as a result of numerous reports on the unsatisfactory performance of resilient seats in butterfly-type isolation valves due to seal deterioration,
' periodic leakage integrity tests of the 18, 8, and 6-inch butterfly isolation valves in the purge system are necessary.
Therefore, the licensee should also propose a Technic,al Specification for testing the valves in accordance with the following. testing frequency:
"The leakage integrity tests.of the isolation valves in the containment l
purge / vent lines shall Le conducted at least once every three months."
The purpose of the leiakage integrity tests ~ of the isolation valves in the containment purge lines is to identify excessive degradation' of the resilient seats for these valves.
Therefore, thay need not be conducted with the precision required for the Type C isolation valve tests
'in 10 CFR Part 50, Appendix J..
These tests would be performed in addition
. to the quantitative Type C tests required by Appendix J, and would not.
relieve the licensee of the respor.sibility to conform to the requirements of Appendix J.
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' Subject to successful ' implementation of.the above recomended actions, we find the purge / vent system design and operating practices for Vermont Yankee to be acceptable.
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Enclosur.e 5 SAFETY EVALUATION REPORT VERMONT YANKEE PLANT OVERRIDE OF CONTAINMENT PURGE ISOLATION AND OTHER ENGINEERED SAFE'TY FEATURES ACTUATION SIGNALS i
Introduction
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Instances have been reported at nuclear power plants where the intended j'
automatic closure of containment purge / ventilation valves during a postulated accident would not have occurred beqause the safety -
1 actuation signals were inadvertently overridden and/or blocked, due to. design deficiencies. These instances were determined to constitute 4
1 an Abnormal Occurrence (#78-5). As a followup action, NRR issued a generic letter requesting each licensee to take certain actions.
l Evaluation The enclosed report was prepared for us :by Franklin Resea'rch Center,' as part of our technical assistance contract program. The report provides their technical evaluation of the conformance of the pl6nt design to the l
NRC criteria established for this review.
Regarding the containment purge / vent isolation system, the contractor's report concludes that four of the six criteria are satisfied. Criterion 3 requires that annunciation be provided when the system is in an
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overridden state.
Criterion 4 requires diverse isolation signals, j
including containment high radiation. - These two items are discussed below.
Regarding other ESF systems that have manual override design features, the contractor's report concludes that:
(1) the 10 motor-operated valves associated with the containment spray function may not comply with criterion 1, which requires that the manual override features do not i
interfere with actuation signals that have not occurred; (2) the bypassing i
of the isolation signals to the oxygen analyzers sample lines would j
prevent these valves from responding to any isolation signal; and (3) the common inlet valves to the standby gas treatment system will change j
position directly upon !' resetting" of an actuation signal, a violation of criterion 6.
These three items are discussed below.
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Regarding criterion 3 for the containment purge / venti system, we concur that a visual alarm only is n'ot sufficient.
For various reasons, i
including a reliability of electrical power to the alarm and shift turnover of plant status information, we have required all plants to.
provide a full annunciation.
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Regarding criterion 4 for the containment purge / vent system, Appendix "A" to this SER provides our evaluation regarding the adequacy of the radiation monitoring as a diverse parameter for purge isolation. As discussed in '
Appendix "A", we have determined that the intent of this criterion is complied with.
The containment spray function in this plant design is an auxiliary
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function of the RHR system. Containment spray is not actuated auto-matically - only manually at the, option of the reactor operator..Since the containment spray system wou'ld divert water that may be needed for core cooling, a safety interlock is p~rovided so that containment spray may not be selected unless the core wner level is. above a pre-selected value. An emergency override of th.'s interlock is provided,in the design.
The fact that the manual selector switch prevents the containment spray
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from being actuated automatically in response to high drywell pressure is an inherent part of this manual-option system 'and is acceptable.
The emergency override feature satisfies the criteria for this review i
and is acceptable.
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The adequacy of isolation of the oxygen sample lines was recently '
addressed and apparently found to be acceptable as part of license i
amendment #58, issued November 3,1980, The change of position of various values upon " reset" of the safety actuation signal -is, in general, unacceptable. A comprehensive' review of this matter is being handled separately regarding I&E Bulletin 80-06.
Conclusions The electrical override aspects of ESE actuation signals at this plint have been reviewed against NRC criteria. ' Based upon our review of the contractor's report and our own review of these matters, we conclude:
1.
The design of the containment purge / vent isolation system conforms to five of the six criteria. To comply with Criterion #3, the license ~e should provide full annunciation whenever any actuation sig6al to the system is in an overridden. state.
2.
Based upon the audit of th.e design of other ESF systems that have manual override features, there is reasonable assurance that these systems conform to the.NRC criteria,'with only a single exception. The matter of valve motion upon " reset" of the safety' signal is being addressed as part of I&E Bulletin 80-06. '
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l APPENDIX "A" The Contractor's report identifies that the radiation monitors which automatically actuate PCIS, monitor the fuel pool area and the reactor building.
In discussing the details of this matter with the contractor, '
we were informed that the reactor building monitor is physically in ths
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exhaust plenum for the reactor building vent and that flow through the 1
containmant purge and vent valves exhau'sts through the same plenum.
Therefore, when the purge / vent valves are open, the reactor building monitor is capable of detecting high radiation from the drywell. The intent of the NRC Criterion #4 is to assure that, over the entire.
spectrum of break sizes, two diverse signals wjll be effective.to detect a loss of coolant accident and will automatically isolate the purge and vent valves.
In comparison with a PWR, the smaller free volume of the containment for a BWR (.i.e., the drywell) and the smaller margin between the nominal drywell pressure and its setpoint for actuat' ion of containment isolation (i.e., PCIS) both serve to provide a greater assurance that, if a SB LOCA inside containment were to occur when the purge / vent valves.are closed, the drywell high pressure setpoint will be reached without significant delay. Vessel low water level will also actuate PCIS. If -
the purge / vent valves should be open at the time of the accident, PCIS will be actuated by high radiation, as sensed by the monitor in the reactor building exhaust plenum, as well as by vessel low water level.
Therefore, signal diversity exists in the form of (1) reactor vessel low water level and (2) either drywell high pressure or exhaust radiation, depending upon the status of the porge/ vent valves for breaks inside containment with the exception of the stuck open relief valve.
There are some scenarios that may be postulated where signal diversity may not exist.
For breaks outside containment and small LOCAs caused by stuck open or partially stuck open relief valves (SORV) the high containment pressure signal may not be developed. When the purge / vent valves are closed (which is typically the case during power operation) the containment atmosphere is not being monitored for high radiation; therefore, the water level instrumentation' must be depended upon to generate the containment isolation signal.
If for analysis. purposes, one presumes that the water level instrumentation does not actuate the PCIS, one must address the possibility that the operator may later decide to open the containment, purge / vent valves. This is considered to be a remote possibility because several other instruments in the control room will alert the operator to the conditions. Vermont Yankee has a continuous air monitorTCAM) system that samples the drywel.1 or torus atmosphere with indication / recording and alarming in the control l
room. There is also an area radiation m'onitor (ARM) inJthe drywell that is indicated / recorded / alarmed in the ' control room. Furthermore, normal operating and emergency procedures require sampling of the containment atmosphere prior to venting / purging.
If level instrumentation is lost emergency procedures also require that the operator completely fill the vessel to preclude core uncovery and possible high radiation.in I
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the containment. The operator also has other indications for a SORV.
f Suppression pool level and temperature as well as direct position indication on the safety / relief valve will alert him to the SORV and he can take corrective action to prevent core uncovery.
For 's team j
line breaks outside containment ample indication and protective features exist. High steam flow, high steam line radiation, high steam-line l
tunnel temperature will initiate an MSIV closure and subsequent scram and alert the operator to take corrective action, if needed. Any.
will be monitored on the reactor building ventilation system area radiation in the reactor building due to breaks outside containment l
radiation monitor that will isolate the purge / vent lines.
While the radiation monitor in the reactor buildin~g exhaust plenum does not measure radiation levels inside the drywell under all conditions.
this radiation monitor taken in combination with other signals and operational procedures does provide the intended diversity for containment isolation and is therefore an acceptable alternative to satisfy criterion
- 4 of this review.
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TER-C5257-189 1.
INTRODUCTION Several instances have been reported at nuclear power plants in which the containment ventilation / purge valves would not'have automatically closed when
- required because the safety actuation signals were either overridden or blocked during normal plant operations due to procedural inadequacies, design-deficiencies, and lack of proper management cont'rols. These instances also brought into question the mechanical operability of.the containment isolation
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valves themselves. The U.S. Nuclear Regulatory Commission (NRC) determined these instances to be Abnormal Occurrences (178-5) which were, accordingly, reported to the U.S. Congress.
As a follow-up to these Abnormal Occurrences, the NRC staff is reviewing' the electrical override aspects and the mechanical operability aspects of.
4 containment purging for all operating power reactors.. On November 29, 1978, the NRC issued a letter entitled " Containment Purging During Normal Plant Operation" [1]
- to all boiling water reactor (bWR) and pressurize'd water reactor (PWR) licensees.
In a letter dated January 9, 1979 [2], the Vermont l
Yankee Nuclear Power Corporation (VYC) replied to the NRC generic letter. On i
VYC provided additional information pertaining to the December 27, 1979 [3],
NRC generic letter. On March 12, 1980 [4], the NRC requested that the I
Licensee provide additional information concerning electrical bypass 'and reset of engineered safety feature (ESF) signals for Vermont Yankee. On May 28, l
1980 [5), the Licensee provided a partial response to this request. This information was supplemented by VYC on January 15, 1981 [6]; this submittal 4
j provided an evaluation of NRC criteria for ESF equipment and the necessary information (electrical schematics, system diagrams, and electrical data) to support their evaluation.
This document addresses only the electrical, instrumentation, and control' (EI&C) design aspects of the containment ventilation isolation (CVI) ' system and other engineered safety features.
1
- Numbers in brackets refer to citations in the list of references, Section 4.
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2.
EVALUATION
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3 2.1 REVIEW CRITERIA The primary intent of this evaluation is Eo determine if th'e following,
s NRC staff criteria are met for the safety signals to all ESF_cquipment; o Criterion 1.
In keeping with the requirements of General' Design Criteria (GDC) 55 and 56, the overriding
- of one type of safety -
actuation signal (e.g., radiation) shall not cause the blockage of any other type of safety actuation signal (e.g. -, pressure) for those valves that serve containment isolation function 'only.-
o criterion 2.
Sufficient physical features (e.g., keylock switches).
shall be provided to facilitate adequate administrative controls.-
o Criterion 3.
For every safety system, a system-level annunciation shall be provided when any override is active.
(See NRC Regulatory Guide 1.47.)
Incidental to this review, the following additional NRC staff design criteria were used in the evaluation:
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o criterion 4.
Diverse signals shall be provided to initiate isolation of the containment ventilation system. Specif ically, containment _,high radiation, safety injection actuation, and contairasnt high pressure (where containment high pressure is not a portion of safety injection actuation) should automatically, initiate CVI.
i o Criterion 5.
The instrumentation and control systems provided to initiate the ESP shall be designed and qualified as safety-grade equipment.
o Criterion 6.
The overriding or resetting + of the ESF actuation signal shall not cause any valve or damper to change position.
In this review, Criterion 6 applies primarily to other related ESF systems, because implementation of this criterion for. containment isolation i
has been reviewed by the Iassons Learned Task Force, based on the recommendations'io NUREG-0578, Section 2.1.4.
Automatic valve repositioning
- Override:
The signal is still present, but it is blocked in order that
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a function contrary to the signal may be performed.
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and the circuit is being cleared in
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+)eset The. signal has come and gone, i
order to return it to the normal condition.
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s-TER-CS257-189 F,-
upon reset may be acceptable when containment isolation is not' involved. T,he '
acceptability of repositioning upon reset will be determined on a case-by-case I
basis. Acceptability will be dependent upon system function, design intent, I
and suitable operating procedures.
2.2 CONTAINMENT VENTILATION SYSTEM DESIGN DESCRIPTION
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2. 2.1.
Ge neralize-em Desian s
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The Licenset.,ca s indicated that the instrumentation and c.ontrol sys't.ersi were originally purchased and ins _talled as. safety grade equipment.
A' review
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of initiation logic and wiring diagrams has confirmed.that no. credible. single malfunc* ion will prevent Froper protective' action at the system _leyeliwhen.
required.
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2.2.2 Logic Circuits for Reset, Seal-in, and Trip At Vermont Yankee, conta-inment ventilation is accomplished through the primary containment isolation system (PCIS). This system is composed of two I
control circuitry subsystems (operating an inboard and outboard group of an'd two PCIS trip logic channels (A and B). The trip logic channels valves) supply isolation signals to the control circuitry subsystem and will interrupt power to valve pilot solenoids, shutting ventilation isolation valves under the following conditions:
1.
Automatic Isolation I
Outboard Group High drywell pressure or low reactor vessel water level (1-out-of-2 taken-twice)
Reactor building ventilation exhaust high radiation (1 of 2)
Reactor building (refueling floor zone) high radiation (1 of 2) g.
Reactor building ventilation exhaust low radiation (2 of 2) e' s
Reactor building (refueling floor zone) low radiation (2 of 2).
Inboard Group w
High drywell pressure or low reactor vessel' wa'ter level (1 of 2 taken-twice) i 1
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TER-C5257-189 Reactor building ventilation exhaust high radiation (1 of 2)
Reactor building (refueling floor zone) high radiation (1 of 2)
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Reactor building (refueling floor zone) low radiat, ion (2 of 2) g Reactor building ventilation exhaust low radiation (2 of 2) 2.
Manual Isolation 16 system level manual isolation is provided. Manual' isolation is accomplished through individual valve control rwitch's.
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e With the reactor protection system in the run mode, inboard and outboard valve group pilot solenoids receive power via a valve. control.lo.gic network containing parameter isolation contacts, seal-in contacts (K'23), and bypass switches (three-position, torus bypses-off-drywell bypass) a.2 shown in Figure 1.
I The parameter isolation contacts shown in Figure 1 are driven by slave relays in series with the detector (not shown). An "out of normal operating band signal" wil) cause the detector contacts to open, thus deenergizing the slave relays and opening the associated contacts in the valve control logic I
network. Power will then be, removed from the slave / seal-in relays (K23) and
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subsequently from the valve pilot solenoids. When the monitored parameter returns to the normal operating band (signal cleared), the isolation logic is I
reestablished. However, as a result of an open seal-in relay contact, power will not yet be restored to the seal-in relay or the valve group pilot
- solenoids. Power is restored to the seal-in relay through contacts K23A, I
which are closed when all local valve control switches (in series with the pilot solenoids) are positioned to deenergize ' the pilot solenoids (i.e., to l
close the valves). This will cause the " reset permissive string" local switch i
contacts to shut, energizing K23A. Once the K23A contacts are closed, the erset switch Kil is placed in the reset position, ' causing the reset contacts I
(Kil) to close. This will return power to the reset / seal-in relay (K23) and 6
l make power available to the valve pilot solenoids.
/
r Either reactor building ventilation exhaust or refueling floor zone high radiatic*n isolation signals can be bypassed by switches (Figure 1), one for 1
each electrical train, which close contacts paralleling the appropriate logic j
network parameter. isolation contacts. Activating any of these bypass switches E
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TER-C5257-189 is indicated by a single red light for each of the electrical trains. Since the contacts providing this bypass feature are incorporated in the trip logic
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s network, the actuation of these switches cannot in itself restore power-to the l
l pilot solenoids. Restoration of power to the valve solenoids, allowing valve i
. operation, is accomplished by the use of the bypass switch followed by the f
system reset as previously described.
I 2.2.3. Individual valve control Circuits Major isolation valves--nitrogen purge supply inlet, air purge supply inlet, drywell and torus vent to reactor, building exhaust (outboard group),
and drywell and tutus purge and vent valves (inboard group)-are normally not supplied with power when in the run mode.(K41 open, controlled by a relay and switch in series). Power can be supplied to the pilot solenoids for these vglees, allowing them to be opened by local switches and the use of permissive switches (16AS50) for the outboard and inbo'ard groups.
With the' reactor protection system mode switch in any positi~on other than "Run,'" the 2-in vent relief valves from the drywell or torus and the vent to t'
the emergency gas treatment system may be opened regardless of the condition of the val've control logic network. This bypass is accomplished by selection of " Torus" or "Drywell," as appropriate, on a spring return-to-normal I
keylocked bypass switch. This switch will operate contacts K42IK43 as appropriate and the use of this bypass is annunicated via the PCIS l.
annunciation system.
since the permissive circuit used to allow system reset requires that all i
local valve control switches be in the electrically open position ti.e., valves -
closed) in order to allow ' reset; the valves cannot cha6ge position upon reset and will remain closed until individually reopened.
f 2.3 CONTAIMiENT VENTILATION SYSTDi DESIGN EVALUATION No instances were found in which the overriding of one type of safety actuation signal (e.g., reactor building ventila-ion exh'aust high radiation) caused the blockage of any other type of safety actuation signal (e.g., high f
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TER-C5257-189 I
drywell pressure) for those valves that serve containment isolation function l
b only. Therefore, it was concluded that NRC staff Criterion 1 has been j
satisfied in the PCIS at Vermont; Yankee.
k Override switches provided are keylock-type switches and will support l
adequate admi~nistrative controls. Therefore, it was concluded that NRC staff l
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Criterion 2 has been satisfied in the PCIS at Vermont Yankee.
Each override switch provides one contact which energizes a red light in the control room to display the bypqss gondition to' the operat,or for each individual trip parameter when the switch is placed in the override position.
However, no audible alarm is provided, and consequently Criterion 3 is not satisfied.
1 The six isolation parameters listed in Section 2.2 will automatically initiate primary containment isolation. However, reactor building ventilation i
exhaust radiation (high and low) monitors indicate cantainment atmosphere l
during a venting operation only and do not sample' the containment atmosphere when the containment vent and purge system is isolated. Because this is not j
in complete compliance with Criterion 4, this situation has been identified
[
for NRC staff evaluation with respect to acceptability.
The Licensee has indicated that the instrumentation and control systems I
provided to iniciate the PCIS were purchased and installed as safety grade equipment.
The detailed determination of the adequacy of the environmental qualification of all safety-related systems is being accomplished separately f
by the Equipinent Qualification Branch of the NRC.
For the purposes of this
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review, NRC staff Criterion 5 is satisfied in the PCIS at Vermont Yankeg.
The overriding or resetting of any actuation signal will not cause any valve or damper to change position. This is accomplished by the use of
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permissive switches at the equipment level and also by the provision of re. set i
and override controls at the accident parameter level.
Therefore, it was l
concluded that NitC staff Criterion 6 has been satisfied in the PCIS at Vermont g
Yankee.
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1 2.4 OTHER ENGINEERED SAFETY FEATURE (ESP) SYSTD4 CIRCUITS s
'1b provide a complete evaluation of the ESF system circuits, a general audit of all ESF system circuits and an in-depth review of the circuit for the residual heat removal (R!m) cystem was conducted.
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2.4.1 Description of RHR System Design Initiation signals, Phase A and Phase B, are provided for all RHR eng'ineered safety feature, equipment on each of two separate electrical trains, A and B.
Each train consists of automatic and manual inputs processed through a relay logic circuitry to actuate a relay logic component actuation system.
The initiation signals for each electrical train are arranged to provide automatic initiation upon either of the following signals:
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- 1. high drywell pressure (1 of 2 taken twice) l
- 2. reactor low water level (1 of,2 taken twice) AND low reactor pressure,(1 of 2).
I The RHR logic circuit is activated when an isolation signal is received, thus energizing slave relay K73 (Figure 2). The K73 contacts in the auto initiatiori circuit are then closed, causing the seal-in (K77) relay and l
contacts to pick up.
Switch S9 in the RHR logic circuit is provided to allow testing of the high drywell pressure signal; it is a keylocked, tiwo-position (Auto-Te st) switch.
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Individual pump and valve control circisits have.both manual and automatic control schemes for start-stop or open-close as well as an indication for run r
status or position.
2.4.2 Evaluation of Other ESF Systems Design 2.4.2.1 RHR System No instandes were found in which the overriding of one type of safety i
j actuation signal caused the blockage of any other type of safety actuation signal for those valves that serve containment isolation function only.
l However, ten ESF actuated valves (MO-39 A & B, MO-31 A & B, MO-26 A & B, MO-34
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A&B, and MO-38 A & B), which have functions in addition to containment A
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TER-CS257-189 l
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isolation, are provided with control circuitry that allows the bypassing of I
l automatic ESF actuation (Figure 3).
However, operation of these valves may,be lI required to provide containment spray for pressure cbntrol of 'he containment 1
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at.mosphere in an accident enviro'nment.
I It should be noted that manual initiation of containment spray with an isolation signal present requires (1) activation of the containment spray valve' control circuit (positioning of S17, high.drywell pressure signal l
presen,t, and positioning of S18 [Rx low shroud level emergency override) or K10 [RHR auto initiation activated) and reactor water level above the shroud -
.h l
Figure 3), (2) manual operation of a control switch (access restricte'd) in each individual valve control circuit, and (3) a containment high pressure signal.
In addition, system-level annunciation of this condition is provided.
.I Consequently, Criteria 2 and 3 are satisfied for these ten valves.
l Other than for the ten valves identified above, Criteria 2 and 3 do n,ot'
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apply.
l Either high drywell pres'sure OR reactor low level AND low pr' essure will I
cause automatic initiation of the RHR system. Therefore, it was concluded that NRC staff Criterion 4 has been satisfied it the RHR system at Vermont Yankee.
l The Licensee has indicated that the instrumentation and control systems I
provided to initiate the RHR system were originally purchased and installed as I
safety grade equipment. A review of these circuits by FRC has revealed no credible single malfunction that would prevent proper protective action at the t
system level,when required. Therefore, for the purposes of this review, NRC g
staf f Criterion 5 is satisfied in the RHR system at Vermont Yankee.
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Cae overriding or resetting of any RHR actuation s,ignal will not cause
,any valve or damper to change position. Therefore, it was concluded that NRC staff Criterion 6 has been satisfied in' the RHR system at Vermont Yankee.
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2.4.2.2 Other EpF Systems l
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An audit conducted of other ESF valve control circuits indicate that f
equipment level bypasses are provided for several equipment items at f
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TER-C5257-189 i
r Vermont Yankee which, if actuated following one safety actuation signal, will.
block a second safety actuation signal (or will block an initiating signal 'if s
l actuated prior to receipt of the signal), thus preventing the equipment from l
performing its protective action. This equipment, however, serves functions i
other than containment isolation. These valves are identified as follows:
Sample System Oxygen Analyzer Valves
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Inboard Isolation FSO - 109-75-Al FSO - 109-75-B1 FSO - 109-75-Cl FSO - 109-75-D1 VG-26 Outboard Isolation FSO - 109-75-A2 FSO - 109-75-B2 FSO - 109-75-C2 FSO - 109-75-D2 VG-23 An acceptable alternative to Criteria 2 and 3 is provided, in that:
a.
Two or more sequential switch actions are required.
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b.
Bypass / override actions are administratively controlled by operating procedures, technical specifications, and/or plant directives
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(including access restriction).
I f
c.
Open/ closed valve position indication is provided for each valve.
t Criterion 4 does not apply to ESF valves other than PCIS valves.
The Licensee has indicated that the instrumentation and control systems were originally purchased and installed as safety grade equipment. A review
.I of these circuits by FRC has revealed no credible single malfunction that
. k.
would prevent proper protective action at the system level when required.
Therefore, Criterion 5 is satisfied.
The audit,of about five other ESF valve control circuits showed that at least two valves,(SB-1A and SB-1B, standby gas treatment system common inlet I
valves) will cha'nge position upon resetting of an ESF actuation signal.
I Therefore, Criteria 6 is not satisfied and further revieit is indicated.
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TER-C5257-189 3.
CONCLUSIONS The EI&C design aspects of ESF systems for Vermont Yankee were evaluated using staff design criteria.
i It is concluded that the PCIS circuit design satisfies the NRC staff criteria for containment ventilation and purging operation with the. exceptior) of Cr'iteria 3 and 4.
Satisfaction of Criterion 3 will' require the installation of an audible alarm associated with'the bypass switches for reactor building ventilation exhaust and refueling floor zone high radiation..
I Satisfaction of Criterion 4 will require tha't a radiation detector which monitors containment (i.e., drywell or torus) activity be provided and used to automatically initiate' primary containment isolation.
Other ESF System Circuits
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1.
RHR System i
The RHR circuit design at Vermont Yankee satisfies the NRC staff criteria j
with the exception of Criterion 1 for ten valves associated with the l
containment spray function of the RHR system. This matter is identified for NRC staff consideration.
I 2.
Sample System Oxygen Analyzer Valves l
l For these ten valves, bypasses exist which, if actuated prior to or following a safety actuation signal, will block the initiating signal or j
a second safety actuation signal, respectively. However', FRC has determined that the admini,strative controls and indication provided are i
an acceptable alternative to the NRC staff criteria.
FRC' recommends that a complete review be conducted by the Licensee to ensure that no violations c.f Criterion 1.for other ESF system valve control circuits exist and that, for the ten valves identified above, administrative controls should be established to prevent inadvertent blocking of ESF signals to these valves.
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Standby Gas Treatment System Common Inlet Valves g
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Satisf action of Criterion 6 will require the modific'ation of the two valve contr'ol circuits identified. In addition, it is recommended that j
vermont Yankee perform an in-depth' review of all other ESF valve control circuits to determine their compliance with Criterion 6, and modify those f
circuits which do not comply with the criterion.
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TER-CS257-189 4.
REFERENCES 1.
NRC, Letter to all BWR and PWR licensees,
Subject:
" Containment Purging I
During Normal Plant Operation," November 29, 1978.
I 2.
D. E. Vandenburgh (VYC)
Letter to T. Ippolito (NRC), Subj ect:
L Containment Purging During Normal Plant Operation
' Vermont Yankee Corporation, January 9, 197S.
I 3.,
W. P. Johnson (VYC)
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Letter to T. Ippolito (NRC),
Subject:
Purge and Vent valves Operability, December 27, 1978.
t 4.
T. Ippolito (NRC) l~
Letter to R. L. Smith (VYC), Subjects Request for Additional Information - Containment Purge System.
Vermont Yankee, March 12, 1980.
t l
S.
R. L. Smith (VYC)
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Letter to T. Ippolito (NRC), Subjects
- Additional Information Concerning
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the Electrical Design of Containment Purge Valves Vermont Yankee, May 28,.1980.
6.
R. L. Smith (VYC) i Letter to T.
Ippolito (NRC), Subjects Bypass and Reset of Engineered Safety Features for Vermont Yankee, January 15, 1981.
7.
IEEE Std 279-1971, Criteria for Protection Systems for Nuc, lear I
Power Generating Stations, Institute of Electrical and Electronics
(
' Engineers, Inc., New York, New York.
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'1 Il KSA.R C.D - HIGH DRYWELL PRESSURE OR REACTOR VESSEL LOW LEVEL 2)CLOSLO WHEN ALL VALUE SOLENOIDS ARE DE ENEEGlZED
- 3) TORUS BYPASS PERMtSSIVE
- 4) DRYWELL BYPASS PERMISSIVE E5$
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[ K23A.
K23
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.. K11 VALVE 22
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REFUELING FLOOR..
.. BYPASS Q
HIGH RADIATION ' --
-- SWITCH A fr #
8O RX BLDG SAMPLING 8 VENTEXH
-- BYPASS liii SYSTEM HIGH RADIATION --
SWITCH A OUTBOARD 8K42 K43*
ISOLATlON VLVS REFUELING FLOOR _
_1. BYPASS
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.o HIGH RADIATION
- - SWITCH B 41 44 LOCALSW' RX BLM
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CONTACTS HIGH ATION' SWITCH B 7-1-
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BYPASS A
M SW 10AS50 VENT EXH.1 1 RXDLDG RXBLDG LOW RAD ~~
REFUELING REFUELING FLOOR LOW
.: FLOOR LOW RAD RAD K58 -
- K5A LOCAL.- LOCAL
. LOCAL _. LOCAL..
K5C
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K50 r r
PILOT PILOT PILOT PILOT s
K23A K23 SOLENOID SOLENOID SOLENOBO 8OLENOID
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Figure 1.
PCIS Control $cheme l
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Hi DRY'VELL PRESSURE.
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RX LOW SHROUD LEVEL ~ ".
- Sta K10 ~. !
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-Figure 3.
Containment Spray Valve control Circuit
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J ENCLOSURE 6 CONTAINMENT SYSTEMS LI!1ITING CONDITION FOR OPERATION C
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3.6.1.7 The containment purge supply and exhaust iso'lation valves may be open for safety-related reasons [or shall be locked. closed).
The containment vent line isolation valves may be open for safety-related reasons [or shall be locke,d clossd].
APPLICABILITY:
MODES 1, 2, 3, and 4.
ACTION':
(For plants with valves closed by technical specification)
With one containment purge supply and/or one exhaust isolation valve open, close the open valve (s) within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
(For plants with valves that may be opened by technical specifications)
With one containment purge supply and/or o.ne exhaust isolaticf6 or vent 1.
valve. inoperable, close the associated OPERABLE valve and either restore the inoperable valve to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or lock the
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OPERABLE valve closed.
Operation may then continue until performance of.the next r.equired 2.
valve test provided that the OPERABLE valve is verified to be locked closed at least once per 31 days.
3.
Otherwise, be in at least HOT STANDBY within the next six hours and in COLD SHUTDOWN within the following 30' hours.
4.
The provisions of Specification 3.0.4 are not applicable.
SURVEll' LANCE REQUIREMENTS 4.6.1.7.1 The
-inch containment purge supply and exhaust isolation valves and the _ -incF vent line isolation valves shall be determined locked closed at least once per 31 days.
4.6.1.7.2 The valve seals of the purge supply and exhaust isolation valves and the vent line isolation valves shall be replaced at least one per _ years.
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t CONTAINMENT"SYSTEHS 3/4 4.6.3 CONTAINMENT ISOLATION VALVES LIljITINGCONDITIONFOROPERATION' 5
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3.6.3 The. containment isolation valves specified in Table 3.6-1 shall be OPERABLE With isolation times as shown in Table 3.6-1.
APPLICdBILITY:
MODES 1, 2, 3 and 4.
ACTION:
s.,,
With one or more of the isolation valves (s) specified in Table 3.6-1 inoperable,
. maintain at least one isolation valve OPERABLE in each affected penetration that is open and either:
Restore the inoperable valve (s) to OPERABLE' status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> a.
or Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at leest b.
one deactivated automatic valve secured in the isolation position,
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or Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least c.
one closed manual valve or blind flange; or Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD d.
SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE RE0VIREMENTS The isolation valves specified in Table 3.6-1 shall be demonstated 4.6.3.1 OPERABLE prior to returning the valve to service af ter mainte. nance, repair or l.
replacement work is performed on the valve or its associated actu'ator, control or power circuit by performance'of a cycling test, and verification of isola-tion time.
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CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.6.3.2 Each isolation valve specified ln Table 3.6-1 shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 months by:
Verifying that on a Phase A containment isolation test signal, each a.
Phase A isolation valve actuates to its i' solation position.
b.
Verifying that on a Phase B containment isolation test signal, each Phase B isolation valve actuates to its isolation position.
~4.6.3.3 The isolation time of e'ach power operated or automatic valve of Table 3.6-1 shall be determined to be within its limit when tested pursuant to Specification 4.0.5.
4.6.3.4 The etntainment purge and vent isolation valves shall be demonstated OPERABLE at intervals not to exceed months.
Yalve OP.ERABILITY shall.be determined by ve-ifying that when the measuued leakage rat.e is added to the leakage rates determined pursuant to Specification 4.6.1.2.d for all other Type B and C penetrati~on, tt e combined leakage ratt is less than or equal to 0.60La.
However, the leaktee rate for, the containment purge and vent isolation valves shall be compared to the previously measured leakage rate to detect excessive valve degradation.
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