ML20039D680
| ML20039D680 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 09/30/1980 |
| From: | Debby Hackett, Kountanis B, Radosevic J EG&G, INC. |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20039D677 | List: |
| References | |
| CON-FIN-A-0231, CON-FIN-A-231, TASK-2.E.4.2, TASK-TM 1183-4163, NUDOCS 8201050440 | |
| Download: ML20039D680 (15) | |
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E E TECHNICAL EVALUATION OF THE ELECTRICAL,1 INSTRUMENTATION, AND CONTROL DESIGN ASPECTS 8
QF THE OVERRIDE OF CONTAINMENT PURGE VALVE 15CLATION -
g AND OTHER ENGINEERED 5AFETY FEATURE SIGNALS FOR THE
- BROWNS FERRY NUCLEAR POWER STATION, UNITS 1, 2, AND 3 (DCCKET No 50-259.!C-2SC. AND SC-296) .
,1 by U .
- 0. B. Hackett/3. :<cuntinis Approved for Publication B
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, d. R. Racasevic l Depar1::: rent Manager Ihis document is UNCLASSIFIED a n'un; u enx M1Caoias f./ drocer1CK Depari::: tent Manager i1 f
Work Performed for t.awrence Livermore NationalLaboratory under U.S. Copartment of Energy -
Contract No. CE.ACC8-75 NVQ 1183.
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E ABSTRACT This report docments the technical evaluation of the electrical, instruentation, and control design aspects of the override of containment k
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purge valve isolation and other engineered safety feature signals for the Browns Ferry Nuclear Power Station, Units 1, 2, and 3. The review criteria are based on IEEE Std-2791971 requirements for the safety signals to all purge and ventilation isolation valves.
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FOREWORD I
This report is supplied as part of the Selected Electrical ,
Instrumentation, and Control Systems Issues (SEICSI) Program being con-ducted for the U.S. Nuclear Regulatory Comission, Office of Nuclear 4
Reactor Regulation, Division of Operating Reactors, by Lawrence Livermore National Laboratory, Field Test Systems Division of the Electronics Engineering Depart:nent.
The U.S. Nuclear Regulatory Comission funded the work under an authori::ation entitled "El ectrical , Instrumentation and Control System g
4 Support," B&R 20 19 04 031, FIN A-0231.
The work was perfonned by EG&G, Inc., Energy Measurements Group, San Ramon Operations, for Lawrence Livermore National Laboratory under U.S.
Department of Energy contract numoer DE-ACC8-76NV01183.
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TABLE OF CONTENTS
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- 1. INTRODUCTION . . . . . . . . . . . . . . . . . l
- 2. EVALUATION OF BROWNS FERRY NUCLEAR POWER STATION, UNITS 1, 2, A)0 3 . . . . . . . . . . . . . . . 3 :
2.1 Review Criteria. . . . . . . . . . . . . . 3 2.2 Containment Ventilation Isolation Circuits Design Description. 4 N 2.3 Containment Ventilation Isolation System Design Evaluation . . . . . . . . . . . . . . . 5 2.4 Other Engineered Safety Feature System Circuits . . . 8 g 3. CCNCm S1cNS . . . . . . . . . . . . . . . . . ,
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TECHNICAL EVALUATION OF THE ELECTRICAL, INSTRUMENTATION, AND CONTROL DESIGN ASPECTS
! 0F -
THE OVERRIDE OF CONTAINMENT PURGE VALVE ISCLATION AND OTHER ENGINEERED SAFETY FEATURE SIGNALS
- l FOR THE BROWNS FERRY NUCLEAR POWER STATION, UNITS 1, 2, AND 3 (Docket Nos. 50-259, 50-250, and 50-296) f.
O. B. Hackett/B. Kountants -
EGaG, Inc., Energy Measurements Group, San Ramon Operations 1 1. INTRODUCTION 1
Several instances have been reported where automatic closure of the containment ventilation / purge valves would not have occurred because the safety actuation signals were either manually overridden or blocked during nonsal plant operations. These events resulted from procedural inadequacies, design deficiencies, and lack of proper management controls.
7 These events also brought into question the mechanical operability of the containment isolation valves themselves. These events were determined by the U. S. Nuclear Regulatory Comission (NRC) to be an Abnonnal Occurrence
. (#78-5) and were, accordingly, reported to the U. S. Congress.
i i As a follow-up on this Abnonnal Occurrence, the NRC staff is i reviewing the electrical override aspects and the mechanical operability j
aspects of containment purging for all operating power reactors. On j- November 28, 1978, the NRC issued a letter entitled " Containment Purging j, Ouring Nonnal Plant Operation" [Ref.13 to all boiling water reactor (BWR)
!i and pressurized water reactor (PWR) licensees. In a letter [Ref. 23 dated December 21, 1978, aid a letter (Ref. 33 dated March 1,1979, the Tennessee j[ Valley Authority (TVA), the licensee for the Browns Ferry Nuclear Power
. Station, Units 1, 2, and 3, replied to the NRC generic letter. The licensee met [Ref. 43 with the MRC in Washington, D. C., on June 1,1979, lU
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to discuss this issue further. Additional inforniation [see References) was subsequently received and evaluated. .
I' i'_ This document addresses only the electrical, instrumentation, and
>f control (EI&C) design aspects of the containment ventilation isolation ._
.- (CY!) and other engineered safety features (ESFs). _
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) 2. EVALUATION OF BROWNS FERRY NUCLEAR POWER STATION, UNITS 1, 2, AND 3
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2.1 REVIEW CRITERIA l
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The primary intent of this evaluation is to determine that the -
following requirements are met for the safety signals to all ESF equipment.
l (1) Criterion no.1-In keeping with the requirements of GDC 55 and 56 (Ref. 53, the overriding
- of one
- type of safety actuation signal (e.g., radiation) j, should not cause the blocking of any other type of safety actuation signal (e.g., pressure) for those . .
I' valves that have no function besides containment i solation.
(2) Criterion no. 2--Sufficient physical features (e.g., keylock switches) are to be provided to facilitate adequate administrative controls.
. (3) Criterion no. 3--The system-level annunciation of the overridden status should be provided for every I safety system impacted wnen any override is active (see R.G.1.47) .
il Incidental to this review, the following additional NRC staff
. t' design criteria were used in the evaluation:
l (1) Criterion no. 4-Oiverse signals should be pro-l l
vided to initiate isolation of the containment ventilation system. Specifically, containment ll8 high radiation, safety injection actuation, and containment high pressure (where containment high l]
l i pressure is not a portion of safety injection
' actuation) should autcmatically initiate CVI.
- 1 "ine following definition is given for clarity of use in this evaluation:
Override: The signal is still present, and it is blocked in l
order to perfonn a function contrary to the signal.
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Criterion no. 5--The instrumentation and ' control systems provided to initiate the ESF should be '
designed and qualified as safety-grade equipment.
(3) Criterion no. 6--The overriding or resetting
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- the ESF actuation signal should not cause any valve or damper to change position.
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Criterion 5 in this review applies primarily to related ESF systems because implementation of this criterion for containment isolation systems have been reviewed by the Lessons Learned Task Force, based on the recomendations in NUREG 0578, Section 2.1.4 [Ref. 6]. Automatic valve repositioning upon reset may be acceptable when containment isolation is .
not involved; consideration will be given on a case-by-case basis. Accept-ability would be dependent upon system function, design intent, and suit-able operating procedures. l 2.2 CCNTAINMENT VENTILATION ISOLATION CIRCUITS DESIGN DESCRIPTION Browns Ferry Nuclear Power Station, Units 1, 2, and 3, each have two ESF trains which can cause isolation of the containment ventilation i system, which is labeled Primary Containment Isolation System (PCIS) on i,
these units. One train controls the inboard containment ventilation / purge isolation valves, and the other train controls the outboard isolation f valves. The initiating contacts for each train are described below:
(1) Automatic Contacts (all one-out-of-two, taken twice logic)
(a) Reactor zone vent exhaust high radiation.
(b) Refueling zone vent exhaust high radiation.
(c) High drywell pressure. 3
' (d) Low reactor water level. J "The following definition is given for clarity of use in this evaluation:
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- Reset: l The signal has come and gone, and the circuit is being cleared to return it to the normal condition.
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Il (2) Manual Contacts
$ ( a) None at the system level. (Isolation may be accomplished by the individual PCIS valve manual switches.
The relays for each of the monitored plant conditions have con- -
tacts in each of the two trains that control the PCIS isolation valves. -
Each train is powered by a different electrical bus. The PCIS circuits contain a reset switch, as defined in this report.
i Several 2-inch-diameter isolation valves, which bypass the large ventilation valves, are utilized by the Containment Atmospheric Control -
System (CAC) and the Containment Atmospheric Dilution Sys*an (CAD) to control the containment during nomal plant operations and post-accident operations, respectively. These valves, which go to the flammable gas recombiner and standby gas treat:r.ent systems, have a designed electrical override. The override circuit contains an interlock to prevent its operation if the reactor is in the "run" mode. The override (bypass) switch is a keylock switch with the key controlled by the shift supervisor.
T.~ a override condition is annunciated.
When a monitored plant condition calls for isolation, electric power is removed from the slave relays (e.g.,16A-K23). The slave relay contacts open to remove electric power from the solenoid valves, causing
} the isolation valves to close.
qi The PCIS-valve solenoid valves must remain energized in order for the isolation valve to be kept open. The slave relay circuit contains a seal-in contact to maintain electric power to it as long as a PCIS signal l* is not present. With a PCIS signal present, the valves will not remain fq open and cannot be opened by their manual switch (with the exception of the 45 valves in the bypass circuit described above).
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I The PCIS signal cannot be cleared until the initiating condi-tions(s) is (are) cleared. When all initiating conditions are cleared, pressing the PCIS reset button will restore power to the solenoid valve circuits. Pressing the PCIS reset button would also clear a bypass con-dition of the 2-inch valves described above, if this condition existed.
The individual manual isolation valve switches are pistol-grip handle, maintained-contact type (GE type SBM SW). Hence, once electric power is ,
restored to the solenoid-valve circuits, any isolation valve with i cs
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switch in the "open" position will automatically re-open.
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Incidental to this review, we have noted that the PCIS circuits do not contain a system-level, manual isolation swi tch. To manually isolate the PCIS valves, each of their manual switches must be turned to "close."
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2.3 CONTAINMENT VENTILATION ISOLATION SYSTEM DESIGN EVALUATION In response to this issue, on an interim basis, the 18-inch con-tainment ventilation / purge isolation valves at Browns Ferry Nuclear Power Station, Units 1, 2, and 3, are currently limited to a maximum open position so that the maximum potential closing loads on the valve will not be exceeded.
The PCIS actuation system has a reset switch, but it does not have an override capability. However, the CAC/ CAD systems have a post-
, accident bypass capability for several 2-inch valves as discussed in Section 2.2. Since, the CAC/ CAD systems are designed to control and/or mitigate the containment atmosphere following an accident, their evaluation l
has not been included because it is beyond the scope of criterion 1.
Therefore, we conclude that NRC staff criterion no.1 is satisfied for the PCIS system except that no determination has been made for the CAC,/ CAD I system.
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The CAC/ CAD system bypass has a keylock switch with proper admin-istrative control, as well as annunciation of the bypass condition. . We conclude that NRC staff criteria nos. 2 and 3 are satisfied.
The containment isolation automatic actuation signal is fonned by the four conditions described in Section 2.2. We conclude that NRC staff criterion no. 4 is satisfied. _
The information provided by the licensee states that the instru-mentation and control systems provided to ' initiate ESF at the Browns Ferry Nuclear Power Station, Units 1, 2, and 3, meet the standards for engineered safeguard features. Based on this general statement, we conclude that NRC staff criterion no. 5 is satisfied *. _
f as When all initiating isolation conditions have been cleared, tne PCIS actuation signal can be reset. Upon resetting, any of the isolation valves with a manual switch in the "open" position will automatically re-open. This condition does not satisfy criterion 6. In addition, if the post-accident bypass capability is exercised, any of the subject 2-inch CAC/CAO valves with a manual switch in the "open" position will autcmatic-l ally re-open as soon as the circuit is switched to " bypass." However, evaluation of the CAC/ CAD system is beyond the scope of criterion 6 and no l _
evaluation has been made.
The PCIS circuits do not contain a system-level, manual isolation switch as specified in the 1971 revision of IEEE Std-279. The switch was not included in the design because the date of these units precedes the 1971 revision. However, the addition of a manual isolation switch would j provide a quick and efficient means for the operator to isolate the con-tainment in an emergency. The work involved would be to install the switch in the control panel and connect the associated wiring to the PCIS logic I circuits.
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a "Ine cetails of the qualification of ESF equipment must be confirned during d the review of the response to I.E.Bulletin 79-01.
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2.4 OTHER ENGINEERED SAFETY FEATURE SYSTEM CIRCUITS The licensee discussed its evaluation of ether ESF systems (Reactor Protection System, Containment Isolation, Containment Spray, HPCI, RCIC, CSS, LPCI, and ADS) in response to IE bulletin 79-08 [Refs. 7 througn
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133 Based on these submittals, we conclude that the NRC criteria are satisfied.
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{ 3. CONCLUSIONS
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The EI&C design aspects of containment purge valve isolation and f
i other ESF signals for the Browns Ferry Nuclear Power Station, Units 1, 2, and 3, were evaluated using those design criteria stated in Section 2.1 of -
this report.
h We conclude that, with one exception, the PCIS circuit design
{_ meets the NRC staff criteria. This exception is that under certain con-g-
4 ditions the PCIS valves could automatically re-coen thus not meeting E criterion 6. The licensee has comitted to modify the system to satisfy the criterion for the isolation valves. The evaluation of the CAC/ CAD system is beyond the scope of this review.
Based on our audit, we conclude that the other ESF systems meet the NRC staff criteria.
The PCIS circuits do not contain a system-level manual isolation switch as specified by IEEE Std-279-1971 [Ref.14]. We recommend that a manual system-level isolation switch be incorporated.
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il0 REFERENCES - -
. 1. NRC/00R letter (A. Schwencer) to TVA, " Containment Purging During Nonnal Plant Operation," dated November 28, 1978.
- 2. TVA letter (J. E. Gilleland) to NRC (Thomas A. Ippoli to), Occkets -
50-259, 50-250, and 50-296, (no title), dated December 21, 1978.
- 3. TVA letter (J. E. Gilleland) to NRC (Thomas A. Ippoli to) , Occkets 50-259, 50-250, and 50-296 (no title), dated March 1,1979.
- 4. TVA meeting with the NRC in '4ashington, D. C., on " Containment Purging During Nonnal Plant Operation," dated June 1,1979.
- 5. U.S. Nuclear Regulatory Comission, Standard Review Plan, " Containment -
Isolation System," NUREG 75/087, Rev. L, Section 6.2.4, (n.d.) .
- 5. U.S. Nuclear Regulatory Comission, "TMI-2 Lessons Learned Task Force Status Report and Short-tenn Recomendations," NUREG-0578.
- 7. TVA letter (J.E. Gilleland) to NRC (James P. O'Reilly), Occkets 50-259, 50-250, and 50-296, (no title), dated April 24, 1979. -
- 8. TVA letter (J.E. Gilleland) to NRC (Thomas A. Ippolito) , Occkets 50-259, 50-250, and 50-296, (no title), dated June 12, 1979.
- 9. TVA letter (L.M. Mills) to NRC (Themas A. Ippolito), Occkets 50-259, 50-250, and 50-296, (no title), dated August 5,1979.
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- 10. TVA letter (L.M. Mill s) to NRC (D. G. Eisenhut), Occkets 50-259, f 50-250, and 50-296, (no title), dated October 17, 1979.
- 11. TVA letter (L.M. Mill s) to NRC (O. G. Eisenhut), Occkets 50-259, 50-250, and 50-295, (no title), ' dated November 13, 1979.
- 12. TVA letter (L.M. Mills) to NRC (Thomas A. Ippolito), Occkets 50-259, 50-250, and 50-296, (no title), dated March 17, 1980.
- 13. TVA letter (L.M. Mills) to NRC (Thomas A. Ippolito), Occkets 50-259, 50-250, and 50-296, (no title), dated April 7,1980.
. 14. IEEE Std-279-1971: Criteria for Protection Systems for Nuclear Power Generating Stations (n.c.) .
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ENCLOSURE 4 ,
CONTAINMENT SYSTEMS LI!!ITING CONDITION FOR OPERATION ___
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! 3.6.1.7 The containment purge supply and exhaust isolatien 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 locked closed). .
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 ~
STAND 3Y 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 clants with valves that may be opened by technical specifications)
- 1. With one containment purge supply and/or o.ne exhaust isolation or vent 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 OPERABLE valve closed.
- 2. Operation cay tnen continue until performance of the next required 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 STANOSY within the next six hours and l 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 />, 4 The provisions of Specification 3.0.4 are not applicable.
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- SURVEILLANCE REOUIREMENTS lf 4.6.1.7.1 The -inch containment purge supply and exhaust isolation valves and the ~ -inen vent line isolation valves shall be determined locked closed at least once per 31 days.
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' 4.6.1.7.2 The valve seals of the purge supply and exhaust isolation valves l
and the vent line isolation valves shall be replaced at least one per ___ years.
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CONTAINMENT SYSTEMS 3/4 4.6.3 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 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.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
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: .
- a. 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 /> or
- b. 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 one deactivated automatic valve secured in the isolation position, or
- c. 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 *
- one closed manual valve or blind flange; or
- d. 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 folicwing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.6.3.1 The isolation valves specified in Table 3.6-1 shall be demonstated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is perfomed on the valve er its associated actuator, control or power circuit by perfomance 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 in Table 3.6-1 shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 -
months by:
- a. Verifying that on a Phase A containment isolation test signal, each Phase A isolation valve actuates to its isolation 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 each 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 TPe containment purge and vent isolation valves shall be demonstated OPERABLE at intervals not to exceed months. Valve OPERABILITY shall be determined by verifying that when the measuued leakage rate is added to the leakage rates determined pursuant to Specification 4.6.1.2.d for all other Type B and C penetration, the combined leakage rate is less than or equal to 0.60La.
However, the leakage 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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