ML20058H222
| ML20058H222 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 12/08/1993 |
| From: | Mcgaha J GULF STATES UTILITIES CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20058H223 | List: |
| References | |
| RBEXEC-93-667, RBG-39552, NUDOCS 9312130023 | |
| Download: ML20058H222 (85) | |
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GULF STATES UTILITIES COMPANY l
RIVE & BENO STATION. 5486 U.S HIGHWAY 61 ST.
F F4 A N C. S V I L L E.
LOulSIANA ?0775 POST O F F IC E Box ??O AREA CODE 504 635-9094 346-8651 l,'
JOHN R. FfcGAHA. JR Vke hesident -
River Bend Nuclear Group t
(501) 3314374 rax (5G1) 3814872 December 8,1993 RBG-39552 s
File No. G9.5, G9.42 L
RBEXEC-93-667 i
U.S. Nuclear Regulatory Commission Document Centrol Desk Washington, D.C. 20555 River Bend Station - Unit 1 Docket No. 50 458 i
Gentlemen:
y Gulf States Utilities Company (GSU) hereby files an application to amend the River Bend Station - Unit 1 Technical Specifications. This application is filed to revise applicable Technical Specifications related to system testing and instrumentation calibration, response time testing, and Logic System Functional Tests to' allow a one-time extension of the surveillance intervals.
The pmposed extensions are mquested on a one-time only basis to support our current refueling outage schedule. The extensions will prevent a plant shutdown solely to perfonn surveillance tests which would cause an unnecessary transient on the plant and result in additional radiation exposure to personnel. In addition, the proposed extensions that extend into the refueling outage are necessary to maintain the ' defense in depth' and ~
critical path of the ot.tage as it is pmsently planned. GSU will make a good faith effon p
to complete the surveillance tests within the current frequency if an outage of sufficient duration occurs.
Attachment I and Enclosures 1 through 46 to this letter provide the justificat on for the pmposed revisions to the Technical Specifications as shown in Attachment 3. To aid in the review and at request of your Staff, thejustification for the individual line items of the l
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PDR ADOCK 05000458 p
pp 7.~l Page 2 of 2 Ixtter to U. S. Nuclear Regulatory Commission Document Control Desk December 8,1993 RBG-39552 RBEXEC-93-667 Technical Specifications which require revision has been provided as individual Enclosures to Attachment 1. provides the r.o significant hazards consideration discussion.
If you have any questions or comments, please contact Mr. Irif L. Dietrich of my staff at (504) 381-4866.
Sincerely,
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John R.
icGaha 1
Attachments xc:
U. S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011 NRC Resident Inspector P.O. Box 1051 St. Francisville, LA 70775 Mr. Edward T. Baker U.S. Nuclear Regulatory Commission M/S OWFN 13-H-15 Washington, D.C. 20555 Mr. Glenn Miller Department of Environmental Quality Radiation Protection Division P.O. Box 82135 Baton Rouge, LA 70884-2135 Attn: Administrator
l UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION 1
STATE OF LOUISIANA
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j PARISH OF WEST FELICIANA
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i Docket No. 50-458 l
In the Matter of
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GULF STATES UTILITIES COMPANY
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(River Bend Station
- Unit 1)
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f AFFIDAVIT i
John R. McGaha Jr., being duly sworn, states that he is a Vice President of Gulf States Utilities Company; that he is authorized
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on the part of said company to sign and file with the Nuclear l
Regulatory Commission the documents attached hereto; and that all f
such documents are true and correct to the best of his knowledge, information and belief.
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J hn R.
McGaha Jr.
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Subscribed and sworn to before me, a Notary Public in and for i
the State and Parish above
- named, this 8~L day of b ( CJ w } _f t a 19'M.
My Commission expireE w3 h Life.
l hfillkI(Lw_
YblA Claudia F.
Hurst "
l Notary Public in and for l
West Feliciana Parish, Louisiana I
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ATTACHMENT 1 GULF STATES UTILITIES COMPANY RIVER BEND STATION DOCKET 50-458/ LICENSE NO. NPF-47 l
i SYSTEM TESTING, INSTRUMENTATION CALIBRATION,
.l LOGIC SYSTEM ';UNCTIONAL TESTING AND RESPONSE TIME TESTING
~l (93-15) l DOCUMENT INVOLVED: Technical Specifications ITEMS:
Surveillance Requirement 4.3.1.1, Table 4.3.1.1-1, Item 2.b, footnote (i) l Surveillance Requirement 4.3.1.1, Table 4.3.1.1-1, Item 2.b, footnote (o) j Surveillance Requirement 4.3.1.1, Table 4.3.1.1-1, Item 3 i
Surveillance Requirement 4.3.1.1, Table 4.3.1.1-1, Item 9.a Surveillance Requirement 4.3.1.2, Table 4.3.1.1-1, Item 3 Surveillance Requirement 4.3.1.2, Table 4.3.1.1-1, Item 9.a Surveillance Requirement 4.3.1.3, Table 3.3.1-2, Item 2.b l
Surveillance Requirement 4.3.1.3, Table 3.3.1-2, Item 2.c
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Surveillance Requirement 4.3.1.3, Table 3.3.1-2, Item 3 Surveillance Requirement 4.3.2.1, Table 4.3.2.1-1, Item 6.e Surveillance Requirement 4.3.2.2, Table 4.3.2.1-1, Item 4.a l
Surveillance Requirement 4.3.2.2, Table 4.3.2.1-1, Item 4.b i
Surveillance Requirement 4.3.2.2, Table 4.3.2.1-1, nem 4.c Surveillance Requirement 4.3.2.2, Table 4.3.2.1-1, Item 4.d Surveillance Requirement 4.3.2.2, Table 4.3.2.1-1, Item 4.e Surveillance Requirement 4.3.2.2, Table 4.3.2.1-1, Item 4.f j
Surveillance Requirement 4.3.2.2, Table 4.3.2.1-1, Item 4.g Surveillance Requirement 4.3.2.2, Table 4.3.2.1-1, Item 4.h l
ourveillance Requirement 4.3.2.2, Table 4.3.2.1-1, Item 6.e Surveillance Requirement 4.3.2.3, Table 3.3.2-3, Item 1.a Surveillance Requirement 4.3.2.3, Table 3.3.2-3, Item 1.b Surveillance Requirement 4.3.2.3, Table 3.3.2-3, Item 2.a Surveillance Requirement 4.3.2.3, Table 3.3.2-3, Item 2.b l
Surveillance Requirement 4.3.2.3, Table 3.3.2-3, Item 2.c i
Surveillance Requirement 4.3.2.3, Table 3.3.2-3, Item 2.d Surveillance Requirement 4.3.2.3, Table 3.3.2-3, Item 3.a
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Sun eillance Requirement 4.3.2.3, Table 3.3.2-3, Item 3.b Surveillance Requirement 4.3.2.3, Table 3.3.2-3, Item 4.a Surveillance Requirement 4.3.2.3, Table 3.3.2-3, Item 4.e Surveillance Requirement 4.3.2.3, Table 3.3.2-3, Item 6.d Surveillance Requimnent 4.3.3.1, Table 4.3.3.1-1, Item C.l.f Surveillance Requirement 4.3.3.1, Table 4.3.3.1-1, Item D.I.a
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Suneillance Requirement 4.3.3.2, Table 4.3.3.1-1, Item C.I.f Surveillance Requirement 4.3.3.2, Table 4.3.3.1-1, Item D.l.a Surveillance Requirement 4.3.3.2, Table 4.3.3.1-1, Item D.I.b
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Surveillance Requirement 4.3.3.3, Table 3.3.3-3, Item 1 Surveillance Requirement 4.3.3.3, Table 3.3.3-3, Item 2a j
Surveillance Requirement 4.3.3.3, Table 3.3.3-3, Item 2b l
Surveillance Requirement 4.3.3.3, Table 3.3.3-3, Item 4 l
Surveillance Requirement 4.3.6, Table 4.3.6-1, Item 2.a, footnote (g) i Surveillance Requirement 4.3.6, Table 4.3.6-1, Item 5.a Surveillance Requirement 4.3.6, Table 4.3.6-1, Item 6.a, footnote (g) j Surveillance Requirement 4.3.7.4, Table 4.3.7.4-1, Item 1 l
Surveillance Requirement 4.3.7.4, Table 4.3.7.4-1, Item 2 Surveillance Requirement 4.3.7.5, Table 4.17.5-1, Item I r
Sun'eillance Requirement 4.3.7.5, Table 4.3.7.5-1, Item 9.b Surveillance Requirement 4.3.9.1, Table 4.3.9.1-1, Item 2.a Surveillance Requirement 4.3.9.2, Table 4.3.9.1-1, Item 2.a j
Sun eillance Requirement 4.6.4.2, Table 3.6.4-1, Item a.1 (RWCU Disch.
to Condenser, RWCU Return to FW, RWCU Pump Suction, RWCU i
Pump Disch., RWCU Backwash Disch.)
j Surveillance Requirement 4.8.2.1.c.1 Surveillance Requirement 4.8.2.1.c.2 Surveillance Requirement 4.8.2.1.c.3 Surveillance Requirement 4.8.2.1.c.4 l
Surveillance Requirement 4.8.2.1.d.1 Surveillance Requirement 4.8.2.1.e Surveillance Requirement 4.8.2.2 Surveillance Requirement 4.8.4.3.b REASON FOR REQUEW:
l The River Bend Station (RBS) Technical Specifications (TS) require, in several TS
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Surveillance Requirements (SRs), instmmentation calibration, response time testing and Logic System Functional Tests (LSFT). The TS to which these SRs apply include the-i Reactor Protection System Instrumentation, Isolation Actuation Instmmentation, Emergency Core Cooling System (ECCS) Actuation Instmmentation, Control Rod Block r
Instrumentation, Remote Shutdown Monitoring Instmmentation, Accident Monitoring j
Instmmentation, Plant Systems Actuation Instmmentation, and Reactor Protection System
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Electric Power Monitoring channels. The SRs specify that the required calibration, response time testing and/or LSFT be conducted nominally at refueling intervals but at j
least once every 18 months. Technical Specification 4.0.2 allows a 25% extension of the surveillance interval to 22.5 months, if required, to pmvide flexibility in cycle lengths.
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In addition, system testing involving valve movement or prformance testing, including Primary Containment and Drywell Isolation Valve actuation and de system battery and l
charger surveillances, are required to be perfonned nominally at refueling intervals but at l
least once every 18 months. Again, TS 4.0.2 allows a 25% extension of the surveillance j
interval to 22.5 months, M required, to provide flexibility in cycle lengths.
A one-time change is being requested to extend the surveillance intervals for the above i
cited TS SRs. G5U will make a good faith effort to conduct these surveillance tests on I
the required TS frequency if an outage of sufficient duration occurs. In order to perfonn the above surveillance tests, the plant must be placed in an undesirable configuration which may increase the probability of a plant trip, or the plant must be in a shutdown condition.
To require the plant to shut down solely to perfonn surveillance tests would cause an unnecessary thermal tmnsient on the plant and result in additional radiation exposure to l
personnel. GSU pmposes to amend the cited TS contained in Appendix A to the RBS i
Operating License, as discussed below and in the respective enclosures, to perfonn the subject surveillance tests during the fifth refueling outage (RF-5), presently scheduled to l
begin April 16,1994. The applicability of the extension should be until at least April 19, 1
1994, in order to allow time for cooldown to Mode 4 wherein a majority of the Limiting Conditions for Opention for the respective SRs are not applicable. The number of days mquired for extension are cited herein as nominal periods to allow flexibility in scheduling. Therefore, for these surveillance requirements the pmposed revisions to the i
i TS are worded such that the surveillance tests "may be performed during the fifth refueling outage scheduled to begin April 16, 1994" In addition, TSs wherein the applicability extends to Modes 4 and 5 are also being proposed for extension. The SRs associated with the TS are being extended into the refueling outage, scheduled to end June 8,1994, to i
support the current schedule and system windows already established to provide ' defense in depth' (NUMARC 91-06, " Guidelines for Industry Actions to Assess Shutdown i
Management", December 1991) during the shutdown period and to maintain the critical path of the outage. To require these SRs to be performed prior to the surveillance inten*al expiration within the outage would impact the ' defense in depth' concept, the critical path, and the duration of the outage. Therefore, in order that the refueling outage schedule may be maintained as it is presently planned and to ensure that an unnecessary extension, and thereby a reduction in the ' defense in depth' built into the outage schedule (due to extension of critical system outage times and increase in shutdown risk), does not result, 2
the pmposed sun'eillance interval extensions should be approved. In that the justification applies into the outage and the requested extensions are cited as nominal periods to accommodate necessary flexibility in the outage, the proposed revisions to the TS are worded such that the surveillance tests "may be extended to the completion of the fifth refueling outage scheduled to begin April 16, 1994."
It is also requested that the "N times 18 months" cumulative suneillance interval for various msponse time testing be baselined to this outage; i.e., the beginning of the "N times 18 months" interval be restarted at the respective response time testing dates to be performed during RF-5. This reestablishment of the baseline will ensure that future i
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response time testing intervals, with respect to the cumulative "N times 18 months" l
interval, will not become late due to the interval extensions that are required for RF-5.
The justification provided in this amendment request for the individual response time surveillance interval extensions applies to the "N times 18 months" cumulative surveillance l
interval extensions which would be granted by the reestablishment of the "N times 18 months" test baseline in that the cumulative surveillance interval would not be extended by mom than that being requested herein for individual response time tests.
Should the proposed changes not be granted by February 16,1994, GSU will be forced to implement an unplanned outage during this operating cycle.
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BACKGROUND:
l The River Bend Station - Unit I has been in Operating Cycle 5 since September 8,1992, j
after completing the fourth refueling outage which began March 12,1992. The extensive l
length of RF-4 and several forced outages during Operating Cycle 5 have impacted the 18-month surveillance intervals required by the TS including instrumentation calibration, response time testing, LSFTs, and system / component tests. This will result in several surveillance tests, perfonned during the last refueling outage, to exceed the surveillance interval plus the allowable extension to the interval specified in TS 4.0.2.
DISCUSSION:
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The following pmvides discussion of each of the proposed TS changes cited above.
As stated in the Updated Safety Analysis Report (USAR), Section 7.2.1.1, the Average Power Range Monitors (APRMs) receive input from the local power range monitor (LPRM) channels, and provide a continuous indication of average reactor power from a few percent to greater than rated thermal power. The APRM channels supply trip signals j
to the Reactor Protection System (RPS). The APRM upscale thermal power scram trip setpoints vary as a function of reactor recirculation loop flow. Each APRM channel mceives a flow signal representative of total recirculation flow. This signal is obtained I
by summing the flow signal from the two recirculation loops. These flow signals are sensed from four pairs of elbow taps, two in each recirculation loop. The APRM signal for the thermal power scram trip is passed through a 6-second time constant circuit to j
simulate thennal power. A faster response (approximately 0.09 seconds) APRM upscale trip has a fixed setpoint, not variable with recirculation flow. Any APRM upscale or inoperative alarm, per Section 7.7.1.1.3 of the USAR, initiates a Control Rod block to -
inhibit any movement of rods which would require RPS action if allowed to proceed. If the APRM upscale or inoperative trip occurs, the trip signal initiates a Neutron Monitoring System (NMS) trip in the RPS. Only the trip logic associated with that APRM is affected.
At least one APRM channel in each trip system of the RPS must trip to cause a scram.
Technical Specification SR 4.3.1.1, Table 4.3.1.1-1, Item 2.b, footnote (o), and TS SR 4.3.6, Table 4.3.6-1, Items 2.a, footnote (g) and 6.a. footnote (g) require that the
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Page 5 recirculation flow reference transmitters be calibrated at least once every 18 months (plus a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). The surveillance intervals for these line items will require an e.xtension for a nominal period of 10 days to reach the scheduled start of RF-5. The justification for these extensions is
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provided in Enclosure 1.
In addition, TS SR 4.3.1.3 requires response time testing of the one APRM channel per trip system at least once per 18 months, such that all channels are tested at least once l
every N times 18 months where N is the total number of redundant channels in each RPS -
trip system. The response time testing from the LPRM to the scram pilot solenoid valve i
for the APRM Flow Biased Simulated Thennal Power - High, Neutron Flux - High response time tests (Table 3.3.1-2, Items 2.b and 2.c) require extension of the surveillance intervals for a nominal period of 51 days to reach the scheduled start of RF-5. The justification for extension of these items is provided in Enclosures 2 and 3. Technical Specification SR 4.3.1.1, Table 4.3.1.1-1, Item 2.b, footnote (i) requires each of the eight Flow Biased Simulated Thermal Power - High Time Constant be calibated at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This calibmtion is essentially a response time test of the Timr anstant, and will require a surveillance interval extension for a nominal period of 58 days to reach the scheduled start of RF-5. The justification for this extension is provided in Enclosure 4.
As stated in USAR Section 7.2.1.1, a reactor vessel pressure increase during mactor j
4 operation compresses the steam voids and results in increased reactivity. This causes -
i increased core heat generation that could lead to fuel barrier failure and reactor overpressurization. A scram counteracts a pressure increase by quickly reducing core fission heat generation. The Reactor Vessel High Pressure scram works in conjunction with the pressure relief system to prevent reactor pressure from exceeding the maximum allowable pressure. The Reactor Vessel High Pressure scram setting also protects the core from exceeding thennal hydraulic limits that result from pressure increases during events that occur when the reactor is operating below rated power and Dow. Reactor pressure is monitored by four redundant pressure transmitters, each of.which provides a reactor high pressure signal to one of the four RPS trip logics. Also, USAR Section 5.4.7.1.2 states that the Residual Heat Removal (RHR) System suction valves at the low pressure /high pressure interface with the Reactor Coolant System receive input from the L
Reactor Vessel Pressure transmitters to provide an interlock function. This interlock function will prevent the operator from opening these valves when the reactor pressure is higher than 135 psig. Technical Specification SR 4.3.1.1, Table 4.3.1.1-1, Item 3 and TS SR 4.3.2.1, Table 4.3.2.1-1, Item 6.e require a channel calibration of the Reactor Vessel Steam Dome Pressure - High and Reactor Vessel (RHR Cut-In Permissive) Pressure functional units, respectively, at least once every 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). Technical Specification Table 4.3.1.1-1, Item 3 and TS Table 4.3.2.1-1, Item 6.e require an extension of the calibration i
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surveillance intervals for a nominal period of 5 days to reach the scheduled start of RF-5.
The justification for these extensions is provided in Enclosures 5 and 6, respectively.
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Technical Specification SR 4.3.1.2 and TS SR 4.3.2.2 require a LSFT and simulated automatic actuation of all channels of the RPS and the Isolation Actuation System, i
respectively, at least once per 18 months (with a maximum allowable wrveillance interval extension of 4.5 months per TS 4.0.2). The Reactor Vessel Steam Dome Pressure - High (TS Table 4.3.1.1-1, Item 3) and Reactor Vessel (RHR Cut-In Permissive) Pressure (TS Table 4.3.2.1-1. Item 6.e) will require surveillance interval extensions for these functional units' portions of the LSFT for a nominal period of 5 days to reach the scheduled start of RF-5.
The justification for these extensions is provided in Enclosures ~7 and 8, j
respectively.
I Technical Specification SR 4.3.1.3 requires the RPS response time testing of the functional
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units shown in TS Table 3.3.1.1-2 to be within the limits cited in that table at least once l
per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). The SR also provides that the response time test include at least onc channel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific trip system. The i
Reactor Vessel Steam Dome Pressure - High requires an extension of the interval for the response time testing of Functional Unit 3 of TS Table 3.3.1-2 for a maximum of 32 days.
The justification for this extension is provided in Enclosure 9.
As stated in USAR Section 7.2.1.1.8, water displaced by the Control Rod Dnve pistons during a scram goes to the Scram Discharge Volume (SDV). If the SDV fills with water j
so that insufficient capacity remains for the water displaced during a scram, Control Rod 4
movement would be hindered during a scram. To prevent this situation, the reactor is
.f scrammed by the RPS when the water level in the Discharge Instrument Volume is high.
enough to verify that the volume is filling up, yet low enough to ensure that the remaining capacity in the SDV can accommodate a scram.
In addition, USAR Section 7.7.1.1.3 states that the Control Rod block functions of the l
Rod Control and Information System (RC&IS) has circuitry to initiate-a rod block for Scram Discharge Volume (SDV) Instrument Volume High Water Ixvel. This assures that j
no control rod is withdrawn unless enough capacity is available in the Scram Discharge i
Volume to accommodate a scram. The setting is selected to initiate a rod block earlier i
than the RPS scram that is initiated on Scram Discharge Instrument Volume High Water l
- Izvel, j
i Technical Specification SR 4.3.1.1, Table 4.3.1.1-1, Item 9.a and TS SR 4.3.6, Table 4.3.6-1, Item 5.a require channel calibration of the SDV Water Level - High function at l
least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). The channel calibration of TS Table 4.3.6-1, Item 5.a, requires an e-tension of 24 days to reach the beginning of the refueling outage (April 16, 1994).
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4 The channel calibation of TS Table 4.3.1.1-1, Item 9.a does not require an extension to mach the scheduled beginning of the refueling outage; however, both TS Table 4.3.1.1-1, Item 9.a and TS Table 4.3.6-1, Item 5.a are required to be OPERABLE in Mode 5 during i
fuel movement and core alterations. In order that these may be scheduled during system windows during the outage, an extension for channel calibration surveillance intervals for the cited two requirements am needed to the end of the refueling outage, scheduled for l
June 8,1994. This will require that the surveillance interval for the channel calibration for TS Table 4.3.1.1-1, Item 9.a be extended within the outage for a nominal period of 29 days. The justiGcation for this extension is provided in Enclosure 10. As well, the surveillance interval for channel calibration of TS Table 4.3.6-1, Item 5.a requires a total extension to and into the outage for a nominal period of 77 days. Thejustification for this extension is pmvided in Enclosure 11.
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Technical Specification SR 4.3.1.2 requires a LSFT and simulated automatic actuation of all channels of the Reactor Protection System at least once per 18 months (with a maximum allowable surve.illance inten'al extension of 4.5 months per TS 4.0.2). The Scram Discharge Water Level - High Level Transmitter functional unit (TS Table 4.3.1.1-1, Item 9.a) will require a LSFT surveillance interval extension for a nominal period of 29 days to reach the scheduled end of RF-5. Thejustification for this extension is provided in Enclosure 12.
As stated in USAR Section 7.3.1.1.2, the Containment and Reactor Vessel Isolation Control System (CRVICS) includes the instmment channels, trip logics, and actuation circuits that automatically initiate valve closure providing isolation of the containment and/or reactor vessel, and initiation of systems provided to limit the release of radioactive materials. During normal plant operation, the isolation control system sensors and trip logic that are essential to safety are energized. When abnormal conditions are sensed,
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instrument channel relay contacts open and deenergize the trip logic and thereby initiate i
isolation. Once initiated, the CRVICS trip logics seal in and may be reset by the opemtor only when the initiating conditions return to normal. The following variables provide j
input to the CRVICS logics for initiation of reactor vessel and Containment isolation, as i
well as initiation of trip of other plant functions when predetennined limits are exceeded:
i a) Reactor Vessel Irw Water I2 vel - A low water level in the reactor vessel could indicate that reactor coolant is being lost through a breach of the Reactor Coolant Pressure Boundary (RCPB) and that the core is in danger of becoming overheated as the reactor coolant inventory diminishes. Thme reactor vessel low water level isolation trip settings are used to complete the isolation of Containment and the reactor vessel. The first and highest (trip Ixve13) reactor vessel low water level isolation trip setting initistes closure of RHR isolation valves. The second reactor vessel low water level (trip Level 2) initiates closure of all valves in major process pipelines except the Main Steam Lines and drains. The Main Steam Lines are left open to allow the removal of heat from the reactor core. The third and lowest (trip Level 1) reactor vessel low water level completes the isolation of Containment and
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reactor vessel by initiating closure of the Main Steam Isolation Valves (MSIVs) and Main Steam Line drain valves.
b) Drywe.ll High Pmssure - High pressure in the drywell could indicate a breach of the RCPB inside the Drywell and that the core is in danger of becoming.
overheated as reactor coolant diminishes. When a predetermined increase in Drywell pressure is detected, the CRVICS initiates RHR system isolation and closure of the Main Steam Line drain valves.
c) Main Steam Line - High Radiation - The Main Steam Line radiation monitoring senses the gross release of fission pmducts fmm the fuel and initiates appmpriate actions to limit fuel damage and contain the released fission products. When the l
Main Steam Line radiation level exceeds a predetermined value, CRVICS initiates closure of all MSIVs, Main Steam Line dmin valves, and Reactor Water Sample valves.
The Off Gas System mechanical vacuum pump is tripped and the l
mechanical pump lines are isolated.
d) Main Steam - High Flow - Main Steam high flow could indicate a breach in the Main Steam Line. Automatic closure of isolation valves prevents excessive loss of reactor coolant and release of significant amounts of radioactive material from j
the RCPB. When excessive steam flow is detected, trip signals initiate closure of the MSIVs and Main Steam Line drain valves.
i c) Main Steam Turbine Inlet - Low Steam Pressure - Low steam pressure while the reactor is operating could indicate a malfunction of the nuclear system pressure l
regulator in which the Turbine Control Valves or Turbine Bypass Valves become fully open, and cause rapid depressurization of the reactor vessel.- When a decrease in Main Steam pressure below a preselected value is detected, the CRVICS initiates closum of the MSIVs Main Steam Line drain valves.
f) Reactor Water Cleanup (RWCU) System. - High Differential Flow - High i
differential flow in the RWCU system could indicate a breach of the RCPB in the l
cleanup system. The flow at the inlet to the system (suction from the Recirculation lines) is compared with the flow at the outlets of the system (flow return to Feedwater or flow to the Main Condenter and/or Radwaste). When an increase in RWCU system differential flow is detected, the CRVICS initiates closure of all RWCU system isolation valves. An automatic timing circuit is provided to bypass the RWCU system high differential flow trip during normal RWCU system surges.
This time delay bypass prevents inadvertent system isolations during system opemtional changes.
g) RWCU System - Area High Ambient Temperature and Differential Temperature
- High temperature in the equipment mom areas of the RWCU system could indicate a b cach of the RCPB in the cleanup system. A breach may also be
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detected by high diffwential temperature between inlet and outlet ventilation air for these areas.
I h) RWCU System - SLCS Initiation (USAR Section 9.3.5.2)- Actuation of either SLCS keylocked switch closes either the RWCU outboard or inboard isolation valve to prevent loss or dilution of the boron.
Technical Specification SR 4.3.2.3 requires the Isolation System response time testing of l
the trip functions units shown in TS Table 3.3.2-3 to be within the limits cited in that table at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). The SR also provides that the response time test include at i
least one channel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific trip system. The functional units of TS Table 3.3.2-3 requiring an extension are as follows:
i 1.a -
Primary Containment Isolation and Manual Initiation Switches, Reactor
- Vessel Water Level - Low Low Level 2. Requires a surveillance inten'al extension for a nominal period of 58 days to reach the scheduled stan of RF-5.
i 1.b - Primary Contaii: ment Isolation and Manual Initiation Switches, Drywell i
Pressure - High. Requires a surveillance interval extension for a nominal l
period of 31 days to reach the scheduled start of RF-5.
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2.a - Main Steam Line Isolation and Manual Initiation Switches, Reactor Vessel Water Level - Ixw Low Low level 1. Requires a sun'elllance interval l
cxtension for a nominal period of 58 days to reach the scheduled start of l
RF-5.
j 2.b - Manual Initiation Switches, Main Steam Line Radiation - High. Requires f
a sun'eillance interval extension for a nominal period of 7 days to reach the scheduled stan of RF-5.
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2.c - Main Steam Line Isolan a and Manual Initiation Switches, Main Steam i
Line Pressure - Low. Requires a surveillance interval extension for a t
nominal period of 58 days to reach the scheduled stan of RF-5.
2.d - Main Steam Line Isolation and Manual Initiation Switches, Main Steam i
Line Flow - High. Requires a surveillance interval extension for a nominal i
period of 59 days to reach the scheduled stan of RF-5.
3.a - Secondary Containment Isolation and Manual Initiation Switches, Reactor j
Vessel Water Izvel - Low Iow Ixvel 2. Requires a surveillance inten'al l
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extension for a nominal period of 58 days to reach the scheduled start of' RF-5.
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3.b - Secondary Containment Isolation and ManualInitiation Switches, Drywell Pressure - High. Requires a surveillance intenral extension for a cominal period of 24 days to reach the scheduled stan of RF-5.
i 4.a - Reactor Water Cleanup System Isolation, Differential Flow - High.
Requires a sun'elllance interval extension for a nominal period of 37 days to reach the scheduled start of RF-5.
4.e - Reactor Water Cleanup System Isc' tion, Reactor Vessel Water Level -
Low law Level 2. Requires a surveillance interval extension for a nominal period of 58 days to reach the scheduled start of RF-5.
6.d - RHR System Isolation, Reactor Vessel Water Izvel - Iow Iow Low Level l
- 1. Requires a surveillance interval extension for a nominal period of 47 days to reach the scheduled stan of RF-5.
l The justification for the response time test surveillance interval extension of each of the above items is provided in Enclosures 13 through 23.
]
Technical Specification SR 4.3.2.2 requires an LSFT and simulated automatic actuation l
of all channels of the Isolation System to be perfonned at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). Table 4.3.2.1-1, Items 4.a (RWCU Isolation - Differential Flow - High), 4.b (RWCU Isolation -
Differential Flow Timer), 4.c (RWCU Isolation - Equipment Area Temperature - High),
l 4.d (RWCU Isolation - Equipment Area Diffemntial Temperature - High), 4.e (RWCU Isolation - Reactor Vessel Water Izvel - Iow law level 2), 4.f (RWCU Isolation - Main 4
I Steam Line Tunnel Ambient Temperature - High),4.g (RWCU Isolation - Main Steam Line Tunnel Differential Temperature - High), and 4.h (RWCU Isolation - SLCS Initiation) require a LSFT surveillance interval extension for a nominal period of 13 days i
to teach the scheduled start of RF-5. The justification for this extension is provided m j 4.
i In addition to the above, TS SR 4.6.4.2 requires the automatic isolation valves shown on TS Table 3.6.4-1 be demonstrated OPERABLE at least once per 18 months (with a ~
j maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2) by verifying that, on an isolation test signal, the automatic isolation valves actuate to their isolation position. As pan of the RWCU extension cited above, several RWCU valves of TS Table 3.6.4-1 require extension to meet the requirement of TS SR 4.6.4.'2. The valve designations are:
1G33 *MOVF001, IG33 *MOVF004, IG33 *MOVF028, i
1 G33 *MOVF034, 1G33 *MOVF039, 1G33 *MOVF040, 1 G33 *MOVF053, l
IG33*MOVF054. These valves require a surveillance interval extension for a nominal e
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period of 12 days to reach the beginning of RF-5. The justification for this extension is provided in Enclosure 25.
l As stated in USAR Section 7.3.1.1, the purpose of the Emergency Core Cooling Systems (ECCS) instrumentation and controls is to initiate appropriate msponses from the_ system to ensure that the fuel is adequately cooled in the event of a Design Basis Accident (DBA).
The cooling provided by the system restricts the release of radioactive materials fmm the j
fuel by preventing or limiting the extent of fuel damage following the situations in which coolant is lost from the RCPB. Automatic initiation of the ECCS is provided when reactor vessel water level or drywell high pressure reach predetermined limits. These parameters i
have been discussed above for the CRVICS. The ECCS network has built-in redundancy so that adequate cooling can be provided, even in the event of specified single failures.
As a minimum, the following equipment makes up the ECCS-I a) High Pressure Core Spray (HPCS) b) Low Pressure Core Spray (LPCS) c) Low Pressure Coolant Injection (LPCI) loops d) Automatic Depressurization System (ADS)
As part of the ECCS function, the HPCS provides high pmssure reactor vessel core spray for small breaks which do not depressurize the reactor vessel. As cited in USAR Section 6.3.2.2.1, a low flow bypass line with a motor-operated gate valve connects to the HPCS discharge line upstream of the check valve on the pump discharge line. The line bypasses water to the suppression pool to prevent pump damage due to overheating when other I
discharge line valves are closed. The valve automatically closes when flow in the main discharge line is sufficient to provide required pump cooling. Technical Specificatic-n SR 4.3.3.1, Table 4.3.3.1-1, Item C.1.f and TS SR 4.3.3.2 mquire channel calibration and i
LSFT, respectively, of the HPCS Pump Discharge Pressure - High at least once per 18 months (with an allowable surveillance interval extension of 4.5 months per TS 4.0.2).
p The calibration surveillance requirement for TS Table 4.3.3.1-1 and TS SR 4.3.3.2 for trip function Item C.I.f require extension in the surveillance intervals for a nominal period of 27 days to reach beginning of the refueling outage (April 16, 1994); however, since the system is being relied upon for a ' defense-in-depth' during refueling operations, a total extension of the existing surveillance intervals for a nominal period of 57 days is required to allow reaching the scheduled system outage window during RF-5. Thejustification for the channel calibration extension is provided in Enclosure 26. The justification for l
extension of the LSFT surveillance interval is discussed in Enclosure 27.
Technical Specification SR 4.3.3.3 requires the ECCS response time testing of the i
functional units shown in TS Table 3.3.3-3 to be within the limits cited in that table at j
least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months). The SR also provides that the response time test include at least one channel per l
trip system such that all channels are tested at least once every N times 18 months where i
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N is the total number of redundant channels in a specific trip system. The ftmetional units
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of TS Table 3.3.3-3 requiring an extension are as follows:
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1-Low Pressure Core Spray System.
Requires a surveillance interval
.l extension of 47 days to reach the scheduled start of RF-5. However, in that l
LPCS is required to be OPERABLE in all Modes the nominal period of l
extension for this item is 100 days to allow the testing to be completed by
-l the scheduled end of RF-5.
2.a - Ixw Pmssure Coolant Injection Mode of RHR System, Pumps A and B.
l Requires a surveillance interval extension for a nominal period of 47 days to reach the scheduled stan of RF-5. However, in that LPCI is required to be OPERABLE in all Modes the nominal period of extension for this item is 100 days to allow the testing to be completed by the scheduled end of
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RF-5.
2.b - Iow Pressure Coolant Injection Mode of RHR System, Pump C. Requires a surveillance interval extension of 47 days to reach the scheduled stan of l
RF-5. However, in that LPCI is required to be OPERABLE in all Modes j
the nominal period of extension for this item is 100 days to allow the i
1 testing to be completed by the scheduled end of RF-5.
v 4-High Pressure Core Spray System.
Requires a surveillance interval i
extension of 31 days to reach the scheduled start of RF-5. However, in that HPCS is being used to provide ' defense in depth' during refueling l
operations, a nominal period of extension of 61 days is required to allow the system to reach its outage window during RF-5.
The justification for surveillance interval extension for each of the above items is provided in Enclosures 28 through 31.
The ECCS systems are powered from the Class IE 4.16-kv buses. Each bus is divided into divisions for separation.
Associated with each of these buses are three diesel l
generators. As cited in USAR Section 8.3.1.1.3.6, diesel generators I A (Division I) and
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IB (Division U) are devoted to safety-related equipment and diese.1 generator IC (Division
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III) energizes the HPCS system. As stated in USAR Section 8.3.1.1.3.9, each of the j
divisions are provided with two (2) completely separate scP.aes of undervoltage protection j
at the 4.16-kv level. The first undervoltage scheme detects a loss of power at the Class lE buses. This undervoltage setpoint is set below any anticipated transient undervoltage condition. The second level of undervoltage protection is set at approximately 90 percent 7
and utilizes a time delay which establishes a sustained degraded voltage condition. Upon i
detection and exceedance of the time delays, the Class IE systems are automatically separated fmm the offsite power system, the load shed logic and load sequence timers i
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t start, and the diesel generator (s) stans and permits auto-close of the diesel generator breakers.
Technical Specification SR 4.3.3.1, Table 4.3.3.1-1, and TS SR 4.3.3.2 for functional unit Items D.I.a and D.l.b require a channel calibration and LSFT, respectively, of the 4.16-j kv Standby Bus Undervoltage (Sustained Undervoltage) and 4.16-kv Standby Bus l
Undervoltage (Degraded Voltage) at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). The surveillance intervals for o
the Division II undervoltage and degraded voltage relays will require extensions to reach i
the scheduled stan of the refueling outage (April 16,1994) for a nominal period of I day.
l However, in accordance with TS 3/4.8.1.2, Division I or II power sources am required j
to be OPERABLE in all Modes of operation; therefore, a nominal period of extension for
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the existing surveillance intervals of 31 days is required for Items D.l.a and D.I.b to allow reaching the scheduled system outage window during RF-5. The justification for the channel calibration interval extensions for TS Table 4.3.3.1-1, Items D.I.a and D.I.b is provided in Enclosures 32 and 33, respectively. The justification for extension of the LSFT intervals is provided in Enclosures 34 and 35.
As stated in USAR Section 7.4.1.4, the Remote Shutdown System (RSS) is designed to l'
achieve and maintain hot reactor shutdown and subsequently to achieve cold shutdown from outside the Main Cont ol Room following these postulated conditions:
a.
The plant is at normal operating conditions, all plant personnel have been 5
evacuated from the Main Control Room, and it is inaccessible for control-i of the plant.
I b.
The initial event that causes the Main Control Room to become inaccessible is assumed to be such that the Reactor Operator can manually scram the reactor before leaving the Main Contml Room.
l The RSS is required only during times of Main Control Room inaccessibility when normal i
plant operating conditions, i.e., no transients or accidents, are occurring. For this reason,
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only the equipment which interfaces with safety-related equipment (RHR, RCIC [ Reactor Core Isolation Cooling], etc.) is required to be of safety-related quality. As pan of the indication at the Division I and II Remote Shutdown Panels, reactor leve1 and pressure indicators am provided. Technical Specification SR 4.3.7.4.1, Table 4.3.7.4-1, Items 1
}
and 2, require that these instmmentation channels units be calibrated at least once per 18 l
months (with a maximum allowable surveillance interval extension of 4.5 months per TS
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4.0.2). The Reactor Vessel Pressure instmmentation channel (Item 1) will require a j
nominal surveillance interval extension of 11 days to reach the scheduled refueling outage stan date (April 16, 1994). The justification for this surveillance interval extension is provided in Enclosure 36. The Reactor Vessel Water Level instrumentation channel (Item
- 2) will require a nominal surveillance interval extension of I day to reach the scheduled beginning of RF-5. The justification for this extension is provided in Enclosure 37.
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The RBS Post-Accident Monitoring instmmentation is shown on in USAR Table 7.5-2.
l This safety-related display instrumentation provides adequate infonnation to allow the j
operator to perform the necessary manual safety functions during accident conditions.
l Technical Specification SR 4.3.7.5, Table 4.3.7.5-1, Items I and 9.b require a channel calibration of the instmmentation at least once per 18 months (with an allowable surveillance interval extension of 4.5 months per TS 4.0.2). Technical Specification Table 4.3.7.5-1, Item 1, Reactor Vessel level, will require a surveillance interval extension for a nominal period of 10 days to reach the scheduled start of RF-5. Thejustification for this surveillance interval extension is pmvided in Enclosure 38. Technical Specification Table l
4.3.7.5-1, Item 9.b, Drywell Area Radiation, will require a surveillance interval extension j
for a nominal period of 2 days to reach the start of RF-5. The justification for this extension is provided in Enclosure 39.
i As cited in USAR Section 7.7.1.3, the Feedwater Contml System controls the flow of feedwater into the reactor vessel to maintain the vessel water level within predetennined i
limits during all normal plant operating modes. The range of water level is based upon the requirements of the steam sepantors (this includes limiting carryover, which affects turbine performance, and carryunder, which affects recirculation pump opemtion). The
[
Feedwater Control System utilizes reactor vessel water level, steam flow, and feedwater flow as a three-element contml. Single element contml is also available based on water i
level only. Normally, the signal from the feedwater flow is equal to the steam flow signal; thus, if a change in the steam flow occurs, the feedwater flow follows. The steam flow signal pmvides anticipaticn of the load changes in water level that will result from a change in load. The level signal pmvides a correction for any mismatch between the steam and feedwater flow which causes the level of the water in the reactor vessel to rise or fall accordingly. A Main Turbine and Feedwater Pump trip will occur if a two-out-of-three logic is satisfied on reactor vessel high level. Technical Specification SR 4.3.9.1, l
Table 4.3.9.1-1 Item 2.a, requires the Reactor Vessel Water Level - High Level 8 be calibrated at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). In addition, TS SR 4.3.9.2 requires a LSFT and simulated automatic operation of all channels of the trip functions of TS Table 4.3.9.1-1 be performed at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2).
9,e surveillance intervals for the channel i
calibration and LSFT for trip function 2.a of TS Table 4.3.9.1-1 require an extension for l
a nominal period of 10 days to reach the scheduled beginning of RF-5. The justification for these extensions is provided in Enclosures 40 and 41, respectively.
1 As stated in USAR Sections 8.3.2.1 and 8.3.2.2, station service de power is available at 125-V. There are three ungrounded Class IE safety-related 125-V de systems. Each system includes a 480-V ac to 125-V de or 120-V ac to 40-V de static battery charger with a control panel. Safety-related chargers are powered from the standby system of their own division.
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The objective of the Division I and II safety-related 125-V de systems is to provide a highly reliable source of de power and control power for necessary de loads, such as
- 1 pumps, valves, relays, control devices, cimuit breaker operating mechanisms, inverters of unintermptible power supply systems, and similar equipment requiring de power. The l
devices provided de power by the Division I and II safety-related 125-V de systems am i
required for safe operation of the station and for safe reactor shutdown under DBA l
conditions. In the same vein, the objective of the Division HI 125-V de power system to l
provide a reliable, continuous, and independent 125-V de power source of control and motive power for the HPCS system logic, HPCS diesel generator set control and protection, and all Division III related control.
j Technical Specifications 4.8.2.1.c.1,4.8.2.1.c.2, and 4.8.2.1.c.3 require at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per j
TS 4.0.2) that the batteries be subjected to an inspection to detect deterioration, damage, corrosion and cell-to-cell resistance. The Division II batteries do not mquire a surveillance l
interval extension to reach the scheduled start of RF-5. However, in that a division of the i
batterits are required to be OPERABLE during Modes 4 and 5, the Division H batteries do require an extension of the interval for a nominal period of 29 days to reach its i
scheduled outage window to perform the inspections. Thejustification for this surveillance i
a interval extension is provided in Enclosure 42.
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Technical Specification SR 4.8.2.1.c.4 requires at least once per 18 months (with a maximum allowable extension of the interval of 4.5 months per TS 4.0.2) the battery l
chargers be verified to be capable, during tests of at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, of supplying 300
.i amperes for chargers I A and IB (Division I and II, respectively) and 50 amperes for charger 1C (Division III). The Division II battery charger load test surveillance interval requires an extension of 11 days to reach the scheduled beginning of RF-5. The Division I battery charger load test surveillance interval will not expire before the scheduled start of RF-5, but will expim during the outage. In that at least one division of the de sources 1
are requimd to be OPERABLE during Modes 4 and 5, a surveillance interval extension i
for both divisions of battery chargers is needed through the end of the outage so that the testing for the respective division may reach its scheduled window within the outage. The 1 A (Division I) battery charger requires a surveillance interval extension within the outage for a nominal period of 20 days. The IB (Division H) battery charger requires a i
surveillance interval extension to the scheduled end of the outage for a nominal period of 64 days. Thejustification for the surveillance interval extension is provided in Enclosure
]
43.
Technical Specification SR 4.8.2.1.d.1 requires at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2 ) the batteries be subjected to a battery service test. The Division H battery will require a surveillance interval extension for a nominal period of 9 days to reach the scheduled stait of RF-5. However, to allow the battery to remain in service during Modes 4 and 5 until 3
it reaches its scheduled outage window, a surveillance interval extension for a nominal
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period of 40 days is required. The justification for this extension is provided in Enclosure 44.
l Technical SpeciHcation SR 4.8.2.1.e requires at Icast once per 60 months (with a maximum allowable surveillance interval extension of 15 months per TS 4.0.2) the battery j
capacity be verified to be at least 80% of the manufacturer's rating when subjected to a perfomiance discharge test. The Division III batteries will require a surveillance interval l
extension of 13 days to mach the scheduled start of RF-5. However, since the HPCS is being relied upon to provide ' defense in depth' during refueling operations, a total surveillance interval extension for a nominal period of 44 days is required for the Division j
III batteries to reach the scheduled HPCS outage window. The justi6 cation for this i
extension is pmvided in Enclosure 45.
l Technical Specification 3/4.8.2.2, D.C. Sourtes - Shutdown, provides the reqmremer.ts far de power sources during Modes 4 and 5 and when irradiated fuel is being handled in the Primary Containment or Fuel Building. Surveillance Requirement 4.8.2.2 references SR 4.8.2.1 for the surveillances required to demonstate operability of the required de power sources. A note, provided for consistency and clarification, has been added to the SR 4.8.2.2 to identify that SR 4.8.2.1 sun'eillance items have been extended. In that the
-l addition of this note is editorial only, no furtherjustification is required.
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As stated in USAR Section 8.3.1.1.3.8, protection is provided for RPS buses A and B i
from voltage and frequency anomalies which could damage RPS components and thus preclude proper operation of the RPS. The protection is afforded by use of electrical protection assemblies (EPAs) which are Class IE. The EPAs provide redundant pmtection j
to the buses by acting to disconnect the buses from the power sources not within design specifications.
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The EPA consists of a circuit breaker within a trip coil driven by logic circuitry which senses line voltage and frequency and trips the circuit breaker open on conditions'of overvoltage, undervoltage, and underfrequency. Technical Specification SR 4.8.4.3.b requires the overvoltage, undervoltage and underfrequency protective instmmentation be demonstrated OPERABLE at least once per 18 months (with a maximum allowable I
surveillance interval extension of 4.5 months per TS 4.0.2). The operability demonstration is to consist of a channel calibration, including simulated automatic actuation of the protective relays, tripping logic and output circuit breakers and verifying the setpoints.
The EPAs are required in all Modes of operation and require a surveillance interval
(
extension to the scheduled end of the outage to support various windows within the outage.
The existing surveillance interval expires after the scheduled start of RF-5, and requires an extension for a nominal period of 42 days to reach the scheduled end of RF-5. The justincation for this extension is included in Enclosure 46.
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'Il REVISED TECHNICAL SPECIFICATIONS B
i The requested revisions to the Technical Specifications identified above are shown on l. The revisions are one-time only extensions of the surveillance intervals to a
allow the surveillance testing to be performed during the fifth mfueling outage scheduled to begin April 16, 1994.
SCHEDULE FOR ATTAINING COMPLIANCE 1
As indicated above, RBS is currently in compliance with the applicable Technical Specifications. The Technical Specification revisions are required prior to Febmary 16, t
1994, in order to avoid a unit outage or placing the plant in an undesimble condition solely
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to conduct the required surveil:ance tests as discussed herein, e
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NOTIFICATION OF STATE PERSONNEL g
A copy of this amendment request has been pmvided to the State ofIouisiana, Department of Environmental Quality - Radiation Protection Division.
ENVIRONMENTAL IMPACT APPRAISAL GSU has reviewed the proposed license amendment request against the criteria of f
10 CFR 51.22 for categorical exclusion from environmental review. The proposed changes to the Technical Specifications do not involve a significant hazards consideration, i
do not significantly change the types or significantly increase the amounts of effluents which may be released offsite, and do not significantly increase individual or cumulative occupational exposure. Based on the foregoing, GSU concludes that the proposed change i
meets the criteria given in 10 CFR 51.22(c)(9) for categorical exclusion fmm the requirement for environmental review.
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i ENCLOSURE 1 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION j
TECHNICAL SPECIFICATION SR 4.3.1.1, TABLE 4.3.1.1-1, 1
ITEM 2.b, FOOTNOTE (o) l TECHNICAL 51'ECIFICATION SR 4.3.6, TABLE 4.3.6-1, ITEMS 2.a.
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FOOTNOTE (g) AND 6.a, FOOTNOTE (g) l RPS/ CONTROL ROD BLOCK RECIRCULATION FLOW TRANSMITTER CALIBRATION i
i Technical Specification SR 4.3.1.1, Table 4.3.1.1-1, Item 2.b, footnote (o) and TS SR 4.3.6, Table 4.3.6-1, Items 2.a, footnote (g) and 6.a, footnote (g) require the Recirculation
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Flow Reference Transmitters which provide input to the Average Power Range Monitors Flow Biased Thermal Power - High Instrumentation be demonstrated OPERABLE by performance of a channel calibration at least once per 18 months (with a maximum
{
allowable extension of 4.5 months per TS 4.0.2). The flow transmitters, Rosemount Model 1152 tnmsmitters, will requim an extension of the SR intervals cited in footnote (b) to TS Table 4.3.1.1-1 and footnote (g) te TS Table 4.3.6-1 for a nominal pedod of 10 j
days to reach the scheduled stan of RF-5.
l In Febmary 1990, Rosemount published a repon, "30 Month Stability Specification For Rosemount Model 1152,1153,1154 Pressum Transmitters" (Rosemount Repon D8900126, Revision A) [ accepted by NRC Safety Evaluation Repon dated August 2,1993 l
on Peach Bottom Atomic Power Station docket). This repon supponed the extension of
)
the calibration interval for the transmitters from 18 months to 30 months basM on a j
reduction in the drift allowance from 0.29% URL [ upper range limit] (2 sigma) for 18
)
months to 0.20% URL (2 sigma) for 30 months. In addition, General Electric (GE) j setpoint calculations assumed 18 month calibration of the trip interval for trip units.
i However, the trip units are calibrated either monthly or quanerly, depending upon the TS requirement for channel functional testing. The GE setpoint calculations utilized a drift j
value of 0.23% SP [setpoint] (2 sigma) which is bounded by the required drift value of i
0.13% SP (2 sigma).
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The existing GE setpoint calculations for Rosemount transmitters and trip unit channels am boundng. There is adequate allowance in the calculations for 30 month drift and no potential impact on plant safety analyses (i.e., no analytic limit changes are equired).
Therefore, the requested extension is justified.
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ENCLOSURE 2 l
JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION l
RESPONSE TIh1E TESTING FOR TFCHNICAL SPECIFICATION SR 4.3.1.3, TABLE 3.3.1-2, ITEh! 2.b APRhi - FLOW BIASED SIMULATED THERA 1AL POWER - HIGH REACTOR PROTECTION SYSTEh! INSTRUh1ENTATION i
i Technical Specification SR 4.3.1.3 requires the Response Time of the Reactor Pmtection System (RPS) Instrumentation shown on Table 3.3.1-2 be demonstrated to be within the limits at least once per 18 months. Each of the tests are to include at least one channel-j per trip system such that all channels are tested at least once every N times 18 months j
i where N is the total number of redundant channels in a specific reactor trip system. Trip function Item 2.b, Average Power Range hionitor Flow Biased Simulated Thermat Power -
l High, of Table 3.3.1-2 will become overdue prior to the beginning of RF-5 scheduled to i
begin April 16, 1994, This RPS response time test requires an extension for a nominal l
period of 51 days to reach the scheduled start of RF-5.
l a
The' extension would have no substantial measumble effect on plant safety because:
- a. There are several redundant APRhis that can initiate the scram operation.
- b. One-out-of-two redundancy exists in every individual instrument channel within l
cach trip ftmetion.
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- c. The instmmentation failure pmbability is a very small fraction of the total
-l control rod insenion (scram failure probability).
l
- d. There are several redundant and diverse instrument channels which can detect and generate a scram signal (e.g., flux, pressure, etc).
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- e. The failure of instrumentation in the sluggish responding mode is a small fraction of its overall failure.
Extension of the Peach Bottom Atomic Power Station Units 2 and 3 surveillance intervals e
for RPS response time testing from 18 to 24 months was accepted by the NRC in the l
Safety Evaluation Repon dated August 2,1993. The justification cited above is very 1
similar to that provided in the NRC Safety Evaluation Report which states:
"The RPS system consists of two independent trip systems with at least two subchannels of a parameter per trip system. The logic of the RPS' system is such j
that either subchannel can trip a trip system and that both trip systems must trip to i
cause a reactor trip. The logic is such that a single failure will neither cause nor prevent a required reactor scram. The licensee states that, based on the inherent i
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redundancy in the RPS system, the impact of extending the response time surveillance interval on system availability is small."
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4 Based on the above, a one-time extension of the RPS Instrumentation, TS Table 3.3.1-2, Item ~2.b, response time testing surveillance requirement interval is justified.
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i ENCLOSURE 3 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION RESPONSE TIME TESTING FOR TECHNICAL SPECIFICATIOM SR 4.3.1.3, TABLE 3.3.1-2, ITEM 2.c APRM - NEUTRON FLUX - HIGH REACTOR PROTECTION SYSTEM INSTRUMENTATION Technical Specification SR 4.3.1.3 mquires the Response Time of the Reactor Protection
[
System (RPS)1nstmmentation shown on Table 3.3.1-2 be demonstrated to be within the
(
limits at least once per 18 months. Each of the tests am to include at least one channel i
per trip system such that all channels am tested at least once every N times 18 months i
where N is the total number of mdundant channels in a specific reactor trip system. Trip j
function Item 2.c, Average Power Range Monitor Neutron Flux - High, of Table 3.3.1-2 l
will become overdue prior to the beginning of RF-5 scheduled to begin April 16, 1994.
[
This RPS response time test requims an extension for a nominal period of 51 days to reach j
the scheduled stan of RF-5.
The extension would have no substantial measumble effect on plant safety because:
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- b. One-out-of-two redundancy exists in every individual instmment channel within i
each trip function.
i 9
9
- c. The instmmentation failure probability is a very small fraction of the total f
i control rod insenion (scram failum probability).
e
- d. There are several redundant and diverse mstmment channels which can detect l
and generate a scram signal (e.g., flux, pmssure, etc).
i i
- e. The failure ofinstrumentation in the sluggish responding mode is a small 3
fraction of :ts overall failure.
'i i
Extension of the Peach Bottom Atomic Power Station Units 2 and 3 surveillance intervals for RPS response time testing fmm 18 to 24 months was accepted by the NRC in the i
Safety Evaluation Repon dated August 2,1993. The justification cited above is very l
similar to that provided in the NRC Safety Evaluation Repon which states-I "The RPS system consists of two independent trip systems with at least two subchannels of a parameter per trip system. The logic of the RPS system is such -
l that either subchannel can trip a trip system and that both trip systems must trip to cause a reactor trip. The logic is such that a single failure will neither cause nor prevent a mquired reactor scram. The licensee states t;iat, based on the inherent j
4 Attached to: RBG-39552 Page 22 redundancy in the RPS system, the impact of extending the response time survelitance interval on system availability is small."
Based on the above, a one-time extension of the RPS Instntmentation, TS Table 3.3.1-2, Item 2.c, response time testing surveillance requirement interval is justified.
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l Attached to: RBG-39552 Page 23 ENCLOSURE 4 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION I
RESPONSE TIME TESTING FOR TECHNICAL SPECIFICATION SR 4.3.1.3, TABLE 3.3.1-2, ITEh! 2.b, FOOTNOTE (i) l APRM FLOW BIASED SIMULATED THERMAL i
POWER - HIGH TIME CONSTANT REACTOR PROTECTION SYSTEM INSTRUMENTATION
-i Technical Sp-cification Table 4.3.1.1-1, Item 2.b, footnote (i) requires the Average Power i
Range Monitor Flow Biased Simulated Thermal Power - High Time Constant be calibrated at least once per 13 months (with a maximum allowable extension of the surveillance interval of 4.5 momhs per TS 4.0.2). The calibration of the Time Constant is essentially i
a msponse time test. This time test requires an extension for a nominal period of 58 days to reach the scheduled start of RF-5.
The extension would have no substantial measumble effect on plant safety because:
I
- b. One-out-of-two redundancy exists in every individual instrument channel within each trip function.
l
- c. The instmmentation failure probability is a very small fraction of the total j
control ind insertion (scram failure probability).
- d. There are several redundant and diverse instrument channels which can detect and generate a scram signal (e.g., flux, pressure, etc).
l
- e. The failure of instrumentation in the sluggish responding mode is a small fraction of its overall failure.
j Extension of the Peach Bottom Atomic Power Station Units 2 and 3 surveillance intervals for RPS response time testing from 18 to 24 months was accepted by the NRC in the Safety Evaluation Report dated August 2,1993. The justification cited above is very similar to that provided in the NRC Safety Evaluation Report which states:
"The RPS system consists of two independent trip systems with at least two subchannels of a parameter per trip system. The logic of the RPS system is such that either subchannel can trip a trip system and that both trip systems must trip to cause a reactor trip. The logic is such that a single failure will neither cause nor prevent a required reactor scram. The licensee states that, based on the inherent redundancy in the RPS system, the impact of extending the response time-l l
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f Attached to: RBG-39552 Page 24 surveillance interval on system availability is small."
Based on the above, a one-time extension of the RPS Instmmentation, TS Table 3.3.1-2, Item 2.b, footnote (i), calibration (response time testing) surveillance requirement interval is justified.
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r Attached to: RBG-39552 Page 25
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ENCLOSURE 5
^
JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICA' HON SR 4.3.1.1, TABLE 4.3.1.1-1, ITEM 3 t
REACTOR VESSEL STEAM DOME PRESSURE - HIGH l
TRANSMITTER CALIBRATION s
REACTOR PROTECTION SYSTEM INSTRUMENTATION Technical Specification SR 4.3.1.1, Table 4.3.1.1-1, Item 3 requires the Reactor Protection System (RPS) Reactor Vessel Steam Dome Pressure - High Instmmentation be demonstrated OPERABLE by performance of a channel calibration at least once per 18 months (with a maximum allowable extension of 4.5 months per TS 4.0.2). The pressure transmitters, Rosemount Model 1152 transmitters, will require an extension of the SR interval cited in TS Table 4.3.1.1-1, Item 3 for a nominal period of 5 days to reach the scheduled start of RF-5.
In February 1990, Rosemount published a report, "30 Month Stability Specification For l
Rosemount Model 1152, 1153, 1154 Pressure Transmitters" (Rosemount Report D8900126, Revision A) [ accepted by NRC Safety Evaluation Repon dated August 2,1993 l
on Peach Bottom Atomic Power Station, Units 2 and 3 docket]. This report supponed the i
extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance from 0.29% URL (2 sigma) for 18 months to
[
0.20% URL (2 sigma) for 30 months. In addition, General Electric (GE) setpoint calculations assumed 18 month calibration of the trip interval for trip units. However, the trip units are calibrated either monthly or quanerly, depending upon the TS requirement for channel functional testing. The GE setpoint calculations utilized a drift value of 0.23 %
SP (2 sigma) which is bounded by the required drift value of 0.13% SP (2 sigma).
The existing GE setpoint calculations for Rosemount transmitters and trip unit channels are bounding. There is adequate allowance in the calculations for 30 month drift and no i
potential impact on plant safety analyses (i.e., no analytic limit changes am required).
Therefore, the requested extension is justified.
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l' Attached to: RBG-39552 Page 26 i
ENCLOSURE 6 t
JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.2.I. TABLE 4.3.2.1-1, ITEM 6.e REACTOR VESSEL (RHR CUT-IN PERMISSIVE) PRESSURE - HIGH CALIBRATION ISOLATION ACTUATION INSTRUMENTATION Technical Specificatica SR 4.3.2.1 Table 4.3.2.1-1, Item 6.e requims the Isolation Actuation System Reactor Vessel (RHR Cut-In Permissive) Pressure - High Instrumentation be demonstrated OPERABLE by performance of a channel calibration at least once per 18 months (with a maximum allowable extension of 4.5 months per TS i
4.0.2). The pressure transmitters, Rosemount Model 1152 transmitters, will require an extension of the SR interval cited in TS Table 4.3.2.1-1, Item 6.e for a nominal period of 5 days to reach the scheduled stan of RF-5.
In February 1990, Rosemount published a repon, "30 Month Stability Specification For Rosemount Model 1152, 1153, 1154 Pmssure Transmitters" (Rosemount Report D8900126, Revision A) [ accepted by NRC Safety Evaluation Repon dated August 2,199; on Peach Bottom Atomic Power Station, Units 2 and 3 docket]. This repon supported the extension of the calibration interval for the transmitters from 18 months ta 30 months based on a reduction in the drift allowance from 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. In addition, General Electric (GE) setpoint calculations assumed 18 month calibration of the trip interval for trip units. However, the i
trip units are calibrated either monthly or quarterly, depending upon the TS requirement for channel functional testing. The GE setpoint calculations utilized a drift value of 0.23 %
SP (2 sigma) which is bounded by the required drift value of 0.13% SP (2 sigma).
The existing GE seipoint calculations for Rosemount transmitters and trip unit channels are bounding. There is adequate allowance in the calculations for 30 month drift and no 4
potential impact on plant safety analyses (i.e., no analytic limit changes are required).
Therefore, the requested extension is justified.
i i
i
Attached to: RBG-39552 Page 27 ENCLOSURE 7 i
JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION.
LOGIC SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.1.2 REACTOR VESSEL STEAM DOME PRESSURE - HIGH REACTOR PROTECTION SYSTEM INSTRUMENTATION Technical Specification SR 4.3.1.2 requires a LSFT and simulated automatic actuation of-all channels of the Reactor Protection System (RPS) at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). -The Reactor Vessel Steam Dome Phsure - High functional unit (TS Table 4.3.1.1-1, Item 3) mquims a surveillance interval extension for this functional unit's portion of the LSFT for s
a nominal period of 5 days to reach the scheduled start of RF-5.
As stated in the NRC Safety Evaluation Report (dated August 2,1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling water reactors (BWRs), prepared by the BWR Owners Gmup (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the nchnical components, (e.g., pumps and valves), which are consequently i
tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component faibre, increasing the logic system functional test interval represents no sigmficant change in the ovemil safety system unavailability."
The evaluation above is applicable to RBS and the surveillance inte-a c.'. 'sion of 5 days is bounded by the interval accepted on the Peach Bottom ocka therefore, the surveillance interval extension is justiGed.
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Attached to: RBG-39552 Page 28 l
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ENCLOSURE 8 l
4 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION LOGIC SYSTEM FUNCTIONAL TESTING FOR i
TECHNICAL SPECIFICATIN SR 4.3.2.2 i
REACTOR VESSEL (RHR CUT-IN PEGilSSIVE) PRESSURE - HIGH ISOLATION ACTUATICN INSTRUMENTATION Technical Specification SR 4.3.2.2 requhes a LSFT and simulated automatic actuation of all channels of the Isolation Actuadon System at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). The Reactor Vessel (RHR Cut-In Pennissive) Pressure trip function _(TS Table 4.3.2.1-1, Item 6.e) requires a suneillance interval extension for this trip function's portion of the LSFT for nominal period of 5 days to reach the scheduled start of RF-5.
As stated in the NRC Safety Evaluation Report (dated August 2,1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling water reactors (BWRs), prepared by the BWR Owners Group (NEDC-30936P) show that the overall safety systems' l
reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability."
l The evaluation above is applicable to RBS and the surveillance interval extension of 5 days is bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified.
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Attached to: RBG-39552 f
Page 29 l
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ENCLOSURE 9 i
JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION RESPONSE TIME TESTING FOR TECHNICAL SPECIFICATION SR 4.3.1.3, TABLE 3.3.1-2, ITEM 3 REACTOR VESSEL STEAM DOME PRESSURE - HIGH l
REACTOR PROTECTION SYSTEM INSTRUMEKfATION Technical Specification SR 4.3.1.3 requires the Response Time of the Reactor Protection System (RPS) Instmmentation shown on Table 3.3.1-2 be demonstrated to be within the limits at least once per 18 months. Each of the tests am to include at least one channel i
per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip system. Trip function Item 3, Reactor Pressure Steam Dome Pressure - High, of Table 3.3.1-2 will become overdue prior to the beginning of RF-5 scheduled to begin April 16,1994. This RPS response time test requires an extension for a nominal period of 32 days to reach the l
s+2.M start of RF-5.
~
The extension would have no substantial measumble effect on plant safety because:
3 i
- a. There are redundant sensors that can initiate the scram operation.
t
- b. One-out-of-two mdundancy exists in every individual instmment channel within i
cach trip function.
I
- c. The instrumentation failure probability is a very small fraction of the total l
contml rod insertion (scram failure pmbability).
f
- d. There are several redundant and diverse instmment channels which can detect and generate a scram signal (e.g., flux, pressure, etc).
i
- e. The failure of instrumentation in the sluggish responding mode is a sman l
fraction of its overall failure.
Extension of the Peach Bottom Atomic Power Station Units 2 and 3 suneillance intervals for RPS response time testing from 18 to 24 months was accepted by the NRC in the Safety Evaluation Repon dated August 2,1993. The justification cited above is very similar to that provided in the NRC Safety Evaluation Repon which states:
j "The RPS system consists of two independent trip systems with at least two subchannels of a parameter per trip system. The logic of the RPS system is such that either subchannel can trip a trip system and that both trip systems must trip to cause a reactor trip. The logic is such that a single failure will neither cause nor prevent a required reactor scram. The licensee states that, based on the inherent i
r r
Attached to: RBG-39552 Page 30 -
t i
redundancy in the RPS system, the impact of extending the response time I
surveillance interval on system availability is small."
Based on the above, a one-time extension of the RPS Instrumentation, TS Table 3.3.1-2, i
Item 3, response time testing surveillance requirement interval is justified.
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Attached to: RBG-39552 Page 31 l
i ENCLOSURE 10 i
4' JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION -
TECHNICAL SPECIFICATION SR 4.3.1.1, TABLE 4.3.1.1-1, ITEM 9.a RPS/ SCAM DISCHARGE VOLUME WATER LEVEL - HIGH I
LEVEL TRANSMITI'ER CALIBRATION REACTOR PROTECTION SYSTEM INSTRUMENTATION Technical Specification SR 4.3.1.1, Table 4.3.1.1-1, Item 9.a requires the Reactor l
Protection System (RPS) Scram Discharge Volume Water Level - High I.evel Transmitter be demonstrated OPERABLE by performance of a channel calibration at least once per 18 months (with a maximum allowable extension of 4.5 months per TS 4.0.2). The level transmitters, Gould Model PD3218 Remote Diaphragm Differential Pressure Transmitters, will require an extension of the SR interval cited in TS Table 4.3.1.1-1, Item 9.a for a nominal period of 29 days to reach the scheduled start of RF-5.
However, this instrumentation is required in Mode 5 when Contml Rods are withdrawn; therefom, j
extension of the surveillance interval for a nominal period of 82 days is required to the l
planned end of the refueling outage.
j The Gould transmitters have very large temperature effects associated with its calibration.
These tansmitters monitor the Scram Discharge Volume, a very warm area.
The temperature effect used in the setpoint calculations (12 % URU200"F (maximum span 40-250"F),18% URU200"F (minimum span,45-250"F)) overwhelms any drift analysis (0.25 %URU6 months). In addition, the drift values for 18 months (0.43 % URL) and 24 months (0.50% URL) are smaller than that which was used in the original setpoint calculations (0.75% URL). Therefore, the extension of the surveillance interval is i
justified.
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. Attached to: RBG-39552 i
Page 32 ENCLOSURE 11 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.6, TABLE 4.3.6-1, ITEM 5.a i
CONTROL ROD BLOCK / SCAM DISCHARGE VOLUME WATER LEVEL - HIGH LEVEL TRANSMITTER CALIBRATION CONTROL ROD BLOCK INSTRUMENTATION Technical Specification SR 4.3.6, Table 4.3.6-1, Item 5.a requires the Control Rod Block f
Instmmentation Scram Discharge Volume Water I_evel - High be demonstrated OPERABLE by performance of a channel calibration at least once per 18 months (with a t
maximum allowable extension of 4.5 months per TS 4.0.2). The level transmitters, Gould Model PD3018 Remote Diaphragm Diffemntial Pmssure Transmitters, will' require an l
extension of the SR interval cited in TS Table 4.3.6-1, Item 5.a for a nominal period of 24 days to reach the scheduled start of RF-5. However, this instrumentation is required in Mode 5 when Control Rods are withdrawn; therefore, extension of the sun >cillance interval for a nominal period of 77 days is required to the planned end of the refueling
- outage, j
l The Gould transmitters have very large temperature effects associated with their calibration, These transmitters monitor the Scram Discharge Volume, a very wann area.
The temperature effect used in the setpoint calculations ( 2% URL/200"F (maximum span 40-250"F),18% URU200 F (minimum span, 45-2507)) overwhelms any drift analysis (0.25 %URIJ6 months). In addition, the drift values for 18 months (0.43 % URL) and 24 months (0.50% URL) are smaller than that which was used in the original setpoint calculations (0.75% URL). Therefom, the extension of the surveillance inten'al is justified.
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Attached to: RBG-39552 Page 33 I
t ENCLOSURE 12 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION LOGIC SYSTEh! FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.1.2 SCRAM DISCHARGE VOLUME WATER LEVEL - HIGH REACTOR PROTECTION SYSTEh! INSTRUMENTATION i
J Technical Specification SR 4.3.1.2 requires a LSFT and simulated automatic actuation of i
all channels of the Reactor Protection System at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). The Semm Discharge Water Level - High Level Tmnsmitter functional unit (TS Table 4.3.1.1-1, Item 9.a) will require a surveillance interval extension for a nominal period of 29 days to reach the scheduled end of RF-5.
As stated in the NRC Safety Evaluation Report (dated August 2,1993) related to extension i
of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, sunreil..nce intervals from 18 to 24 months:
" Industry reliability studies for boiling water reactors (BWRs), prepared by the i
BWR Owners Gmup (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the mliabilities of the logic systcm, but by that of the mechanical components, (e.g., pumps and valves), which are consequently j
tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no signi0 cant change in the overall i
5 safety system unavailability."
The evaluation above is applicable to RBS and the surveillance interval extension is bounded by the interval accepted on the Peach Bottom docket; therefore, the sun'eillance interval extension is justified.
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e Attached to: RBG-39552 Page 34 t
ENCLOSURE 13 l
JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION
[
RESPONSE TIME TESTING FOR TECHNICAL SPECIFICATION SR 4.3.2.3, TABLE 3.3.2-3, ITEM 1.a REACTOR VESSEL WATER LEVEL - LOW LOW LEVEL 2 PRIMARY CONTAINMENT ISOLATION AND MANUAL INITIATION SWITCH 83S ISOLATION ACTUATION INSTRUMENTATION Technical Specification SR 4.3.2.3 requires the Response Time of the Isolation Actuation Instmmentation shown on Table 3.3.2-3 be demonstrated to be within the limits at least once per 18 months. Each of the tests are to include at least one channel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific isolation trip system. Technical SpeciGeation Table 3.3.2-3, Item 1.a, Primary Containment Isolation, Reactor Vessel Water I.cVel -
l Low I.ow Level 2 and the Manual Initiation Switch testing of this function will become overdue prior to the beginning of RF-5 scheduled to begin April 16,1994. The response time test require an extension of the suneillance interval for a nominal period of 58 days to reach the scheduled start of RF-5.
Regulatory Guide 1.118 (Revision 2) states:
" Response time testing of all safety related equipment, per se, is not required if, in lieu of response time testing, the response time of the safety equipment is i
verined by functional testing, calibration checks or other tests, or both. This is acceptable if it can be demonstrated that changes in response time beyond acceptable limits are accompanied by changes in performance characteristics which are detectable during routine tests."
On March 27, 1992, the BWR Owner's Group submitted a Licensing Topical Report i
(LTR) prepared by the General Electric Company, NEDO-32013P, " System Analyses For Elimination Of Selected Response Time Testing Requirements", March 1992, for NRC l
review. GSU believes the information contained therein justines a one-time extension of j
the surveillance requirement intervals of TS Table 3.3.2-3, Item 1.a.
The LTR provided justification for the elimination of selected BWR Response Time Testing or Tests (R'IT), as denned in the Instmment Society of America (ISA) Standard S67.06, from the plant TS SRs. The analyses included the affected instrumentation loops which could potentially impact the instrument. loop response time. In addition, plant operating experiences were reviewed to identify response time failures and how they were i
detected. The failure modes identified were then evaluated to determine if the effect on j
response time would be detected by other testing requirements contained in the TS. Based j
on the analyses presented in the LTR, it was concluded that there were no failure modes I
t
Attached to: RBG-39552 Page 35 which will affect the response time of the instrumentation loop which would not be detected by other surveillances such as channel calibration (at least once per 31 days for trip setpoint calibration and at least once per 18 months for sensor calibration), channel functional tests (at least once per 31 days), channel checks (at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />),
or other techniques. [It should be noted herein that the "other techniques" phrase applies to Rosemount transmitter slow oil loss detennination.]
e In addition, individual instrument channel response time delays for specific trip functions (on the order of a fraction of a second) have very little safety significance. Redundancy and diversity exist in most instmmentation trip functions (e.g., flux, water level, pressure).
For the Isolation Actuation Instmmentation, the instmmentation response times are a small fraction of the overall response timen of the actuating devices.
Based on the above, a one-time ; tension of the Isolation Actuation Instrumentation TS SR 4.3.2.3, Table 3.3.2-3,Iten' i.a response time testing surveillance intervalisjustified.
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Attached to: RBG-39552 Page 36 i
f ENCLOSURE 14 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION l
RESPONSE TIME TESTING FOR i
TECHNICAL SPECIFICATION SR 4.3.2.3, TABLE 3.3.2-3, ITEM 1.b DRYWELL PRESSURE - HIGH PRIMARY CONTAINMENT ISOLATION AND I
MANUAL INITIATION SWITCHES I
ISOLATION ACTUATION INSTRUMENTATION I
Technical Specification SR 4.3.2.3 requires the Response Time of the Isolation Actuation Instmmentation shown on Table 3.3.2-3 be demonstrated to be within the limits at least once per 18 months. Each of the tests are to include at least one channel pe.r trip system such that all channels am tested at least once every N times 18 months where N is the total j
number of redundant channels in a specific isolation trip system. Technical Specification 1
Table 3.3.2-3, Item 1.b, Primary Containment Isolation, Drywell Pressure - High, and the Manual Initiation Switch testing of this trip function will become overdue prior to the beginning of RF-5 scheduled to begin April 16,1994. This response time testing requires j
an extension of the surveillance intenal for a nominal period of 31 days to reach the l
scheduled stan of RF-5.
l Regulatory Guide 1.118 (Revision 2) states:
" Response time testing of all safety related equipment, per se, is not required if, in lieu of response time testing, the response time of the safety equipment is l
verified by functional testing, calibration checks or other tests, or both. This is acceptable if it can be demonstrated that changes in response time beyond acceptable limbs are accompanied by changes in perfonnance characteristics which are detectable during routine tests."
{
On March 27, 1992, the BWR Owner's Group submitted a Licensing Topical Report (LTR) prepared by the General Electric Company, NEDO-32013P, " System Analyses For Elimination Of Selected Response Time Testing Requirements", March 1992, for NRC review. GSU believes the information contained therein justifies a one-time extension of the surveillance requirement interval of TS Table 3.3.2-3, Items 1.b.
The LTR provided justification for the elimination of selected BWR Response Time l
Testing or Tests (RTT), as defined in the Instmment Society of America (ISA) Standard
-l S67.06, fmm the plant TS SRs. The analyses included the affected instrumentation loops l
which could potentially impact the instrument loop response time. In addition, plant l
operating experiences were reviewed to identify response time failures and how they were detected. The failure modes identified were then evaluated to determine if the effect on response time would be detected by other testing requirements contained in the TS. Based on the analyses presented in the LTR, it was concluded that there were no failure modes I
1
Attached to: RBG-39552 Page 37 i
i which will affect the response time of the'instmmentation loop which would not be i
detected by other surveillances such as channel calibration (at least once per 31 days for trip setpoint calibration and at least once per 18 months for sensor calibration), channel.
functional tests (at least once per 31 days), channel checks (at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />),
or other techniques. [It should be noted herein that the "other techniques" phrase applies to Rosemount transmitter slow oil loss determination.]
i In addition, individual instmment channel response time delays for specific trip functions (on the order of a fraction of a second) have very little safety significance. Redundancy and diversity exist in most instmmentation trip functions (e.g., flux, water level, pressure).
- For the Isolation Actuation Instrumentation, the instmmentation response times are a small fraction of the overall response times of the actuating devices.
Based on the above, a oi..:-time extension of the Isolation Actuation Instrumentation TS SR 4.3.2.3, Table 3.3.2-3, Item 1.b response time testing surveillance interval is justified.
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Attached to: RBG-39552 Page 38
)
i ENCLOSURE 15 I
JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION RESPONSE TIME TESTING FOR TECHNICAL SPECIFICATION SR 4.3.2.3, TABLE 3.3.2-3, ITEM 2.a REACTOR VESSEL WATER LEVEL - LOW LOW LOW LEVEL 1 l
MAIN STEAM ISOLATION AND MANUAL INITIATION SWITCHES i
ISOLATION ACTUATION INSTRUMENTATION
[
Technical Specification SR 4.3.2.3 requires the Response Time of the Isolation Actuation j
InstmmemaGon shown on Table 3.3.2-3 be demonstrated to be within the limits at least once per 18 months. Each of the tests are to include at least one channel per trip system j
such that all channels are tested at least once every N times 18 months where N is the total l
number of redundant channels in a specific isolation trip system. Technical Specification l
Table 3.3.2-3, Item 2.a, Main Steam Line Isolation, Reactor Vessel Water Level - Low low Irw Level 1 and the Manual Initiation Switch testing of this trip function will i
become overdue prior to the beginning of RF-5 scheduled to begin April 16,1994. This msponse time testing requires an extension of the surveillance interval for a nominal period of 58 days to mach the scheduled start of RF-5.
The extension would have no substantial measurable effect on plant safety because:
l t
- a. Redundancy and diversity exist in individual instmment channels within a trip
(
function (e.g, one-out-of-two twice for level, pmssure, flow, temperature) i
- b. The instrumentation response time is a small fraction of the overall response time of the actuating device.
- c. The instrumentation failure pmbability is a very small fraction of the total main steam line isolation failure probability.
- d. The failure of instmmentation in the sluggish responding mode is a small
(
fraction of its overall failure.
Based on the above, a one-time extension of the Isolation Actuation Instmmentation TS f
SR 4.3.2.3, Table 3.3.2-3, Item 2.a response time testing sun'eillance intervals is justified.
l i
I
Attached to: RBG-39552 Page 39 ENCLOSURE 16 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION RESPONSE TIME TESTING FOR TECHNICAL SPECIFICATION SR 4.3.2.3 TABLE 3.3.2-3, ITEM 2.b MAIN STEAM LINE RADIATION - HIGH, MANUAL INITIATION SWITCHES ISOLATION ACTUATION INSTRUMENTATION 1
Technical Specification SR 4.3.2.3 requires the Response Time of the Isolation Actuation Instrumentation shown on Table 3.3.2-3 be demonstrated to be within the limits at least once per 18 months. Each of the tests are to include at least one channel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific isolation trip system. Item 2.b of TS Table 3.3.2-3, Main Steam Line Isolation, Main Steam.Line Radiation - High, Manual Initiation Switches will become overdue prior to the beginning of RF-5 scheduled to begin April 16, i
1994. This msponse time testing requires an extension of the surveillance inten>al for a nominal period of 7 days to reach the scheduled start of RF-5.
On July 7,1987, the BWR Owner's Group submitted the General Electric Licensing Topical Repor1 (LTR) " Safety Evaluation For Eliminating The Boiling Water Reactor Main Steam Line Isolation Valve Cicsure And Scram Function Of Main Steam Line Radiation Monitor", NEDO-31400, to the NRC for review and appmval. The purpose of this submittal was to justify the removal of the MSIV closure and reactor scram function from the Main Steam Line Radiation Monitors (MRLRMs) to reduce the potential for unnecessary plant shutdowns caused by spurious actuation of the MSLRM trips and to provide for improved availability of the main condenser for removal of decay heat.. The NRC provided its acceptance and issued its Safety Evaluation Report concerning NEDO-31400 on May 15,1991.
l The LTR, NEDO-31400, is applicable to RBS and, therefore, the NRC acceptance can be applied to this extension.
Based on the above, a one-time extension of the Isolation Actuation Instrumentation TS SR 4.3.2.3, Table 3.3.2-3, Item 2.b msponse time testing surveillance requirement interval j
is justified.
j 4
. Attached to: RBG-39552 Page 40 i
ENCLOSURE 17 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION RESPONSE TIME TESTING FOR TECHNICAL SPECIFICATION SR 4.3.2.3, TABLE 3.3.2-3, ITEM 2.c j
MAIN STEAM LINE PRESSURE - LOW MAIN STEAM ISOLATION AND MANUAL INITIATION SWITCHES f
ISOLATION ACTUATION INSTRUMENTATION Technical Specification SR 4.3.2.3 requires the Response Time of the Isolation Actuation Instrumentation shown on Table 3.3.2-3 be demonstrated to be within the limits at least once per 18 months. Each of the tests am to include at least one channel per trip system i
such that all channels are tested at least once every N times 18 months whem N is the total number of redundant channels in a specific isolation trip system. Item 2.c of TS Table 3.3.2-3, Main Steam Line Isolation, Main Steam Line Pmssure - Low, and the Manual Initiation Switch testing of this trip function will become overdue prior to the beginning
(
of RF-5 scheduled to begin April 16, 1994. This msponse time testing requires an extension of the surveillance interval for a nominal period of 58 days to mach the 1
scheduled start of RF-5.
The extension would have no substantial measurable effect on plant safety because:
t
- a. Redundancy and diversity exist in individual instrument channels within a trip function (e.g, one-out-of-two twice for level, pressure, flow, temperature) l
- b. The instrumentation response time is a small fraction of the overall msponse i
time of the actuating device.
[
t
- c. The instmmentation failum probability is a very small fraction of the total main steam line isolation failum probability.
- d. The failure of instrumentation in the sluggish responding mode is a small fraction of its overall failure.
Based on the above, a one-time extension of the Isolation Actuation Instrumentation TS SR 4.3.2.3, Table 3.3.2-3, Item 2.c response time testing surveillance requimment interval is justified.
4 h
f e
Attached to: RBG-39552 Page 41 ENCLOSURE 18 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION RESPONSE TIME TESTING '<OR TECHNICAL SPECIFICATION SR 4.3.2.3, TABLE 3.3.2-3, ITEM 2.d MAIN STEAM LINE FLOW - HIGH MAIN STEAM ISOLATION AND MANUAL INITIATION SWITCHES ISOLATION ACTUATION INSTRUMENTATION Technical Specification SR 4.3.2.3 requires the Response Time of the Isota. ion Actuation Instrumentation shown on Table 3.3.2-3 be demonstrated to be within the limits at least once per 18 months. Each of the tests are to include at least one channel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific isolation trip system. Item 2.d of TS Table 3.3.2-3, Main Steam Line Isolation, Main Steam Line Flow - High, and the Manual Initiation Switch testing of this trip function will become overdue prior to the beginning of RF-5 scheduled to begin April 16, 1994. This msponse time testing requims an extension of the sunreillance interval for a nominal period of 59 days to reach the scheduled stan of RF-5.
The extension would have no substantial measurable effect on plant safety because:
- a. Redundancy and diversity exist in individual instmment channels within a trip function (e.g, one-out-of-two twice for level, pressure, flow, temperature)
- b. The instmmentation response time is a small fmetion of the overall response time of the actuating device.
- c. The instmmentation failure probability is a very small fmetion of the tntal main 2
steam line isolation failure probability.
- d. The failure of instrumentation in the sluggish esponding mode is a small fraction of its overall failure.
Based on the above, a one-time extension of the Isolation Actuation Instmmentation TS SR 4.3.2.3, Table 3.3.2-3, Item 2.d response time testing surveillance requirement interval i
is justified, t
l i
i i
i Attached to: RBG-39552 Page 42 j
i I
ENCLOSURE 19
)
JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION RESPONSE TIME TESTING FOR i
TECHNICAL SPECIFICATION SR 4.3.2.3, TABLE 3.3.2-3, ITEM 3.a REACTOR VESSEL WATER LEVEL - LOW LOW LEVEL 2 SECONDARY CONTAINMENT ISOLATION AND l
MANUAL INITIATION SWITCHES ISOLATION ACTUATION INSTRUMENTATION i
Technical Specification SR 4.3.2.3 requires the Response Time of the Isolation Actuation i
Instrumentation shown on Table 3.3.2-3 be demonstmted to be within the limits at least once per 18 months. Each of the tests are to include at least one channel per trip system l
such that all channels are tested at least once every N times 18 months whem N is the total number of redundant channels in a specific isolation trip system. Technical Specification i
Table 3.3.2-3, Item 3.a, Secondary Containment Isolation, Reactor Water Level - Low I.ow 12 vel '2, and the Manual Initiation Switch testing of this trip function will become overdue prior to the beginning of RF 5 scheduled to begin April 16,1994. These response time tests require an extension of the surveillance intervals for a nominal period of 58 days to reach the scheduled start of RF-5.
Regulatory Guide 1.118 (Revision 2) states:
l
" Response time testing of all safety related equipment, per se, is not required if, in lieu of response time testing, the response time of the safety equipment is verified by functional testing, calibration checks or other tests, or both. This is acceptable if it can be demonstrated that changes in msponse time beyond acceptable limits am accompanied by changes in perfonnance characteristics which are detectable during routine tests."
l On March 27, 1992, the BWR Owner's Group submitted a Licensing Topical Report (LTR) pmpared by the General Electric Company, NEDO-32013P, " System Analyses For Elimination Of Selected Response Time Testing Requirements", March 1992, for NRC review. GSU believes the information contained themin justifies a one-time extension of the surveillance mquirement interval of TS Table 3.3.2-3, Item 3.a.
l The LTR provided justification for the elimination of selected BWR Response Time i
Testing or Tests (RTT), as deGned in the Instrument Society of America (ISA) Standard S67.06. from the plant TS SRs. The analyses included the affected instrumentation loops j
which could potentially impact the instmment loop response time. In addition, plant.
operating experiences were reviewed to identify response time failures and how they were detected. The failure modes identified were then evaluated to determine if the effect on response time would be detected by other testing requirements contained in the TS. Based on the analyses presented in the LTR, it was concluded that there were no failure modes q
i i
~.
, ~.
Attached to: RBG-39552 Page 43 which will affect the response time of the instmmentation loop which would not be detected by other surveillances such as channel calibration (at least once per 31 days for j
trip setpoint calibration and at least once per 18 months for sensor calibation), channel i
functional tests (at least once per 31 days), channel checks (at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />),
t or other techniques. [It should be noted herein that the "other techniques" phrase applies to Rosemount tmnsmitter slow oil loss determination.]
In additi +a, individual instmment channel response time delays for specific trip functions (on the ; cder of a fraction of a second) have very little safety significance. Redundancy and divenity exist in most instrumentation trip functions (e.g., flux, water level, pressure).
i For the Isolation Actuation Instmmer.tation, the instrumentation response times are a small i
fmetion of the overall response times of the actuating devices.
Based on the above, a one-time extension of the Isolation Actuation Instmmentation TS SR 4.3.2.3, Table 3.3.2-3, Item 3.a response time testing surveillance interval is justified.
I e
(
i I
k i
l Attached to: RBG-39552 i
Page 44 l
ENCLOSURE 20 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION RESPONSE TIME TESTING FOR TECHNICAL SPECIFICATION SR 4.3.2.3, TABLE 3.3.2-3, ITEM 3.b i
DRYWELL PRESSURE - HIGH SECONDARY CONTAINMENT ISOLATION AND MANUAL INITIATION SWITCHES ISOLATION ACTUATION INSTRUMENTATION Technical Specification SR 4.3.2.3 requires the Response Time of the Isolation Actuation Instrumentation shown on Table 3.3.2-3 be demonstrated to be within the limits at least once per 18 months. Each of the tests am to include at least one channel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific isolation trip system. Technical Specification i
Table 3.3.2-3, Item 3.b, Secondary Containment Isolation, Drywell Pressure - High, and the Manual Initiation Switch testing of this trip function will become overdue prior to the beginning of RF-5 scheduled to begin April 16,1994. This response time testing requires nn extension of the suneillance inten>al for a nominal period of 24 days to reach the s~heduled stan of RF-5.
Regulatory Guide 1.118 (Revision 2) states:
" Response time testing of all safety related equipment, per se, is not required if, in lieu of response time testing, the response time of the safety equipment is verified by functional testing, calibration checks or other tests, or both. This is acceptable if it can be demonstrated that changes in response time beyond acceptable limits are accompanied by changes in performance characteristics which are detectable during routine tests."
On March 27, 1992, the BWR Owner's Gmup submitted a Licensing Topical Repon (LTR) prepared by the General llectric Company, NEDO-32013P, " System Analyses For Elimination Of Selected Response Time Testing Requirements", March 1992, for NRC review. GSU believes the infom1ation contained therein justifies a one-time extension of the surveillance interval of TS Table 3.3.2-3, Item 3.b.
The LTR provided justification for the elimination of selected BWR Response Time j
Testing or Tests (RTT), as defined in the Instrument Society of America (ISA) Standard S67.06, from the plant TS SRs. The analyses included the affected instrumentation loops l
which could potentially impact the instrument loop response time. In addition, plant operating experiences were reviewed to identify response time failures and how they were detected. The failure modes identified were then evaluated to determine if the effect on response time would be detected by other testing requirements contained in the TS. Based on the analyses presented in the LTR, it was concluded that there were no failure modes i
Attached to: RBG-39552 l
Page 45 I
i which will affect the response time of the instrumentation loop which would not be detected by other surveillances such as channel calibration (at least once per 31 days for trip setpoint calibration and at least once per 18 months for sensor calibration), channel-l functional tests (at least once per 31 days), channel checks (at least once per 12 houc),-
l or other techniques. [It should be noted herein that the "other techniques" phrase applies to Rosemount transmitter slow oil loss determination.]
v In addition, individual instrument channel response time delays for specific trip functions (on the order of a fraction of a second) have very little safety significance. Redundancy
[
and diversity exist in most instrumentation trip functions (e.g., flux, water level, pressure).
l For the Isolation Actuation Instmmentation, the instrumentation response times am a small i
fraction of the overall response times of the actuating devices.
3 2., Tab e 3 3 2 3 Ire i 3.b res onse t i e te g unei ce nt na ju ified i
i l
i I
i i
c j
l
)
m t
i Attached to: RBG-39552 Page 46 i
l i
ENCLOSURE 21 l
i JUSTIFICATION FOR SURVEILLANCE LNTERVAL EXTENSION l
RESPONSE TIME TESTING FOR
}
TECHNICAL SPECIFICATION SR 4.3.2.3, TABLE 3.3.2-3, ITEM 4.a l
DIFFERENTIAL FLOW - HIGH l
REACTOR WATER CLEANUP SYSTEM ISOLATION AND l
MANUAL INITIATION SWITCIIES i
ISOLATION ACTUATION INSTRUMENTATION Technical Specification SR 4.3.2.3 requires the Response Time of the Isolation Actuation j
Instmmentation shown on Table 3.3.2-3 be demonstrated to be within the limits at least once per 18 months. Each of the tests are to include at least one channel per trip system such that all channels are tested at least once every N times 18 months where N is the total j
number of redundant channels in a specific isolation trip system. Item 4.a of TS Table j
3.3.2-3, Reactor Water Cleanup System Isolation, Differential Flow - High, and the i
Manual Initiation Switch testing of this trip function will become overdue prior to the i
beginning of RF-5 scheduled to begin April 16,1994. This response time testing requires
-l an extension of the surveillance interval for a nominal period of 37 days to reach the j
scheduled start of RF-5.
l Regulatory Guide 1.118 (Revision 2) states:
4
' Response time testing of all safety related equipment, per se, is not required if, in lieu of response time testing, the response time of the safety equipment is verified by functional testing, calibration checks or other tests, or both. This is acceptable if it can be demonstrated that changes in response time beyond e
acceptable limits are accompanied by changes in performance characteristics which are detectable during routine tests."
l l
On March 27. 1992, the BWR Owner's Group submitted a Licensing Topical Report (LTR) prepared by the General Electric Company, NEDO-32013P, " System Analyses For l
Elimination Of Selected Response Time Testing Requirements", March 1992, for NRC review. GSU believes the information contained therein justifies a one-time extension of l
the surveillance interval of TS Table 3.3.2-3, Item 4.a.
(
The LTR p ovided justification for the climination of selected BWR Response Time Testing or Tests (RTT), as defined in the Instmment Society of America (ISA) Stimdard S67.06, from the plant TS SRs. The analyses included the affected instmmentation loops i
which could potentially impact the instmment loop response time. In addition, plant l
operating experiences were reviewed to identify response time failures and how they were detected. The failure modes identified were then evaluated to determine if the effect on response time would be detected by other testing requirements contained in the TS. Based f
on the analyses presented in the LTR, it was concluded that there were no failure modes
)
t i
l
i Attached to: RBG-39552 Page 47 I
i which will affect the response time of the instmmentation loop which would not be detected by other suneillances such as channel calibration (at least once per 31 days for trip setpoint calibration and at least once per 18 months for sensor calibration), channel l
functional tests (at least once per 31 days), channel checks (at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />),
{
or other techniques. [It should be noted herein that the "other techniques" phrase applies i
to Rosemount transmitter slow oil loss determination.]
l l
t In addition, individual instmment channel msponse time delays for specific trip functions j
(on the order of a fraction of a second) have very little safety significance. Redundancy and diversity exist in most instmmentation trip functions (e.g., flux, water level, pressure).
l For the Isolation Actuation Instmmentation, the instmmentation response times are a small fraction of the overall response times of the actuating devices.
1 Based on the above, a one-time extension of the Isolation Actuation Instrumentation TS f
SR 4.3.2.3, Table 3.3.2-3, Item 4.a response time testing surveillance interval isjustified.
k j
i h
1 t
h
+
i I
I t
4
=m e
e
-s-r r-
i Attached to: RBG-39552 Page 48 ENCLOSURE 22 i
JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION
+
RESPONSE TIME TESTING FOR l
TECHNICAL SPECIFICATION SR 4.3.2.3, TABLE 3.3.2-3, ITEM 4.e i
REACTOR VESSEL WATER LEVEL - LOW LOW LEVEL 2 REACTOR WATER CLEANUP SYSTEM ISOLATION AND MANUAL INITIATION SWITCHES ISOLATION ACTUATION INSTRUMENTATION l
Technical Specification SR 4.3.2.3 requires the Response Time of the Isolation Actuation i
Instmmentation shown on Table 3.3.2-3 be demonstrated to be within the limits at least once per 18 months. Each of the tests are to include at least one channel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific isolation trip system. Technical Specification Table 3.3.2-3, Item 4.e, Reactor Water Cleanup System Isolation, Reactor Water Ixvel -
?
Low Ixw Level 2, and the Manual Initiation Switch testing of this trip function will
- i become overdue prior to the beginning of RF-5 scheduled to begin April 16,1994. These response time tests require an extension of the surveillance intervals for a nominal period of 58 days to reach the scheduled start of RF-5.
Regulatory Guide 1.118 (Revision 2) states:
" Response time testing of all safety related equipment, per se, is not required if, in lieu of response time testing, the response time of the safety equipment is verified by functional testing, calibration checks or other tests, or both. This is acceptable if it can be demonstrated that changes in response time beyond j
acceptable limits are accompanied by changes in performance characteristics which am detectable during routine tests."
On March 27, 1992, the BWR Owner's Group submitted a Licensing Topical Report i
(LTR) prepared by the General Electric Company, NEDO-32013P, " System Analyses For Elimination Of Selected Response Time Testing Requirements", March 1992, for NRC review. GSU believes the information contained therein justifies a one-time extension of the surveillance requirement interval of TS Table 3.3.2-3, Items 4.e.
1 The LTR provided justification for the elimination of selected BWR Response Time Testing or Tests (RTT), as defined in the Instrument Society of America (ISA) Standard S67.06, from the plant TS SRs. The analyses included the affected instrumentation loops which could potentially impact the instmment loop response time. In addition, plant operating experiences were reviewed to identify response time failures and how they we:e detected. The failure modes identified were then evaluated to determine if the effect on response time would be detected by other testing requirements contained in the TS. Based on the analyses presented in the LTR, it was concluded that there were no failure modes
Attached to: RBG-39552 i
Page 49 i
i which will affect the response time of the instrumentation loop which would not be detected by other surveillances such as channel calibration (at least once per 31 days for trip setpoint calibration and at least once per 18 months for sensor calibration), channel functional tests (at least once per 31 days), channel checks (at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />),
or other techniques. [It should be noted hemin that the "other techniques" phrase applies to Rosemount transmitter slow oil loss determination.]
i In addition, individual instrument channel response time delays for specific trip functions (on the order of a fraction of a second) have very little safety significance. Redundancy and diversity exist in most instmmentation trip functions (e.g., flux, water level, pmssure).
For the Isolation Actuation Instmmentation, the instnnact.tation response times are a small fraction of the overall response times of the actuating devices.
l l
3 Based on the above, a one-time extension of the Isolation Actuation Instmmentation TS SR 4.3.2.3, Table 3.3.2-3, Item 4.e response time testing surveillance interval is justified.
I i
I i
t t
i l
d i
t d
i l
[
9 i
i f
l i
=. - - _ _. -
f Attached to: RBG-39552 Page 50 i
ENCLOSURE 23
{
JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION j
RESPONSE TIME TESTING FOR TECHNICAL SPECIFICATION SR 4.3.2.3, TABLE 3.3.2-3, ITEh! 6.d a
REACTOR VESSEL WATER LEVEL - LOW LOW LOW LEVEL 1 RHR SYSTEh! ISOLATION l
ISOLATION ACTUATION INSTRUMENTATION-i Technical Specification SR 4.3.2.3 requires the Response Time of the Isolation Actuation f
Instntmentation shown on Table 3.3.2-3 be demonstrated to be within the limits at least 1
once per 18 months. Each of the tests are to include at least one channel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific isolation trip system. Technical Specification Table 3.3.2-3, Item 6.d, RHR System Isolation, Reactor Vessel Water Level - Low Low Low Level I will become overdue prior to the beginning of RF-5 scheduled to begin April 16, 1994. This response time testing requires an extension for a nominal period of 47 days to reach the scheduled stan of RF-5.
Regulatory Guide 1.118 (Revision 2) states:
" Response time testing of all safety related equipment, per se, is not required if,
{
in lieu of response time testing, the response time of the safety equipment is verilled by functional testing, calibration checks or other tests, or both. This is i
acceptable if it can be demonstrated that changes in response time beyond l
acceptable limits are accompanied by changes in perfonnance characteristics which 2
are detectable during mutine tests."
On March 27, 1992, the BWR Owner's Group submitted a Licensing Topical Repon (LTR) prepared by the General Electric Company, NEDO-32013P, " System Analyses For i
Elimination Of Selected Response Time Testing Requirements", March 1992, for NRC l
review. GSU believes the information contained therein justifies a one-time extension of i
the surveillance requirement intervals of TS Table 3.3.2-3, Item 6.d.
l The LTR provided justification for the elimination of selected BWR Response Time i
Testing or Tests (R7T), as defined in the Instnament Society of America (ISA) Standard l
S67.06, frem the plant TS SRs. The analyses included the affected instnnnentation k> ops which could potentially impact the instrument loop response time. In addition, plant operating experiences were reviewed to identify response time failures and how they were detected. The failure modes identified were then evaluated to determine if the effect on response time would be detected by other testing requirements contained in the TS. Based
- l on the analyses presented in the LTR., it was concluded that there were no failure modes j
which will affect the response time of the instnamentation loop which would not be detected by other surveillances such as channel calibration (at least once per 31 days for t
-i
Attached to: RBG-39552 j
Page 51 I
trip setpoint calibration and at least once per 18 months for sensor calibration), channel j
functional tests (at least once per 31 days), channel checks (at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />),
or other techniques. [It should be noted herein that the "other techniques" phrase applies to Rosemount transmitter slow oil loss determination.]
In addition, individual instmment channel response time delays for specific trip functions j
(on the order of a fraction of a second) have very little safety significance. Redundancy and diversity exist in most instmmentation trip functions (e.g., flux, water level, pressure).
!^
For the Isolation Actuation Instmmentation, the instrumentation response times are a small fraction of the overall response times of the actuating devices.
s 1
Based on the above, a one-time extension of the Isolation Actuation Instmmentation TS i
SR 4.3.2.3 Table 3.3.2-3, Item 6.d response time testing sun'elllance interval isjustified.
j l
I l
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a
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i 4
4 h
I 1
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i
1 Attached to: RBG-39552 Page 52 ENCLOSURE 24 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION LOGIC SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.2.2 RWCU MANUAL ISOLATION DIFFERENTIAL FLOW - HIGH, DIFFERENTIAL FLOW TIMER, EQUIPMENT AREA TEMPERATURE - HIGH, EQUIPMENT AREA DIFFERENTIAL TEMPERATURE - HIGH,
[
REACTOR VESSEL WATER LEVEL - LOW LOW LEVEL 2, MAIN STEAM LINE TUNNEL AMBIENT TEMPERATURE - HIGH, MAIN STEAM LINE TUNNEL DIFFERENTIAL TEMPERATURE - HIGH, AND SLCS INITIATION ISOLATION ACTUATION INSTRUMENTATION Technical Specification SR 4.3.2.2 requims a LSFT and simulated automatic actuation of all channels of the Isolation System to be performed at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). Table 4.3.2.1-1, Items 4.a (RWCU Isolation - Differential Flow - High), 4.b (RWCU Isolation -
Differential Flow Timer), 4.c (RWCU Isolation - Equipment Area Temperature - High),
4.d (RWCU Isolation - Equipment Area Differential Temperature - High), 4.e (RWCU Isolation - Reactor Vessel Water Level - Low low Level 2), 4.f (RWCU Isolation - Main Steam Iine Tunnel Ambient Temperature - High), 4.g (RWCU Isolation - Main Steam Line Tunnel Differential Tempemture - High), and 4.h (RWCU Isolation - SLCS Initiation) require a surveillance interval extension for a nominal period of 13 days to reach the scheduled stan of RF-5.
I As stated in the NRC Safety Evaluation Report (dated August 2,1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling water reactors (BWRs), prepared by the i
BWR Owners Group (NEDC-30936P) show that the overall safety systems' j
reliabilities are not dominated by the reliabilities of the logic system, but by that i
of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability."
The evalur. tion above is applicable to RBS and the surveillance interval extension of 13 days is bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified.
i
l Attached to: RBG-39552 I
Page 53 l
i ENCLOSURE 25
{
JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION CONTAINMENT ISOLATION VALVE ACTUATION i
TECHNICAL SPECIFICATION SR 4.6.4.2, TABLE 3.6.4-1, ITEM a.1
(
REACTOR WATER CLEANUP SYSTEM VALVES IG33*MOVF001, IG33*MOVF004,1G33*MOVF028, IG33*MOVF034, IG33*MOVF039, IG33*MOVF040, IG33*MOVF053, IG33*MOVF054 i
Technical Specification SR 4.6.4.2 requires the automatic isolation valves shown on TS l
Table 3.6.4-1 be demonstrated OPERABLE at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2) by verifying that, on an isolation test signal, the automatic isolation valves actuate to their isolation position.-
Several RWCU valves of TS Table 3.6.4-1 require extension to meet the requimment of l
TS SR 4.6.4.2.
The valve designations are: IG33*MOVF001, _1G33*MOVF004, IG33 *MOVF028, I G33 *MOVF034, IG33 *MOVF039, IG33 *MOVF040, IG33*MOVF053, IG33*MOVF054.
These valves requim a surveillance intenral extension for a nominal period of 13 days to reach the scheduled start of RF-5.
The penetrations providing isolation of containment have redundancy so that an active l
failure of any single valve or component does not prevent containment isolation. In i
addition, periodic testing of the containment isolation system is performed during power operation, including Inservice Testing of valves. Based on the redundancy provided, l
testing during power operation and the short time period for which the surveillance interval extension is requested, the extension is justified.
l i
I e
I I
Attached to: RBG-39552 Page 54 ENCLOSURE 26 l
JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.3.1, TABLE 4.3.3.1-1, ITEM C.l.f HPCS PUMP DISCHARGE PRESSURE - HIGH TRANSMI' ITER CAI!BRATION EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION Technical Specification SR 4.3.3.1, Table 4.3.3.1-1, Item C.l.f requires the Emergency Core Cooling System (ECCS) High Pressure Core Spray (HPCS) Pump Discharge Pressure - High Instrumentation be demonstrated OPERABLE by performance of a channel calibration at least once per 18 months (with a maximum allowable extension of 4.5 months per TS 4.0.2). The pressure transmitters, Rosemount Model 1152 transmitters, will require an extension of the SR interval cited in TS Table 4.3.3.1-1, Item C.I.f for a nominal period of 57 days to allow reaching the scheduled system outage window during RF-5.
In February In J, Rosemount published a report, "30 Month Stability Specification For Rosemount Model 1152, 1153, 1154 Pressure Transmitters" (Rosemount Repon D8900126, Revision A) [ accepted by NRC Safety Evaluation Repon dated August 2,1993 on Peach Bottom Atomic Power Station, Units 2 and 3 docket). This repon supponed the extension of the 2libration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance from 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. In addition, General Electric (GE) setpoint calculations assumed 18 month calibration of the trip interval for trip units. However, the trip units are calibrated either monthly or quanerly, depending upon the TS requirement for channel functional testing. The GE setpoint calculations utilized a drift value of 0.23 %
SP (2 sigma) which is bounded by the required drift value of 0.13% SP (2 sigma).
The existing GE setpoint calculations for Rosemount transmitters end trip unit channels are bounding. There is adequate allowance in the calculations for 30 month drift and no potential impact on plant safety analyses (i.e., no analytic limit changes are required).
Therefore, the requested extension is justified.
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Attached to: RBG-39552 Page 55 l
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ENCLOSURE 27 i
JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION LOGIC SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.3.2 HPCS PUMP DISCHARGE PRESSURE - HIGH EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION Technical Specification SR 4.3.3.2 requims a LSFT and simulated automatic operation of all channels of the Emergency Core Cooling System Actuation Instrumentation at least l
once per 18 months (with an allowable sun'eillance interval extension of 4.5 months per TS 4.0.2). The HPCS Pump Discharge Pmssure - High trip unit (TS Table 4.3.3.1-1, Item C.I.f) mquires a surveillance intenal extension for a nominal period of 27 days to reach the scheduled refueling outage date (April 16,1994); however, since the system is i
being relied upon for a ' defense-in-depth' during refueling operations, a extension of the existing surveillance intervals for a nominal period of 57 days is required to allow reaching the scheduled system outage window during RF-5.
As stated in the NRC Safety Evaluation Report (dated August 2,1993) related to extension i
of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals l
from 18 to 24 months:
" Industry reliability studies for boiling water reactors (BWRs), prepared by the BWR Owners Group (NEDC-30936P) show that the overall safety systems' i
reliabilities are not dominated by the mliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failum, increasing the logic system functional test interval represents no significan' change in the overall safety system unavailability."
The evaluation above is applicable to RBS and the surveillance interval extension of 57 days is bounded by the interval accepted on the Peach Bottom docl.:et; therefore, the surveillance interval extension is justified.
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Attached to: RBG-39552 Page 56 l
ENCLOSURE 28 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION RESPONSE TihiE TESTING FOR i
TECHNICAL SPECIFICATION SR 4.3.3.3, TABLE 3.3.3-3, ITEM 1 LOW PRESSURE CORE SPRAY SYSTEh!
Eh1ERGENCY CORE COOLING SYSTEh! ACTUATION INSTRUMENTATION Technical Speci5 cation SR 4.3.3.3 requires the Response Time of the Emergency Core Cooling System (ECCS) trip functions shown on Table 3.3.3-3 be demonstrated to be within the limits at least once per 18 months. Each of the tests am to include at least one channel per trip system such that all channels are tested at least once every N times 18 l
months where N is the total number of redundant channels in a specific isolation trip system. Technical Specification Table 3.3.3-3, Item 1, Low Pressure Core Spray System (LPCS), will become overdue prior to the beginning of RF-5 scheduled to begin April 16, 1994. The msponse time test requires an extension for a nominal period of 47 days to mach the scheduled stan RF-5. However, in that LPCS is mquimd to be OPERABLE in j
all Modes the extension for this item for a nominal period of 100 days is mquimd to allow the testing to be completed by the scheduled end of RF-5.
Regulatory Guide 1.118 (Revision 2) states:
" Response time testing of all safety related equipment, per se, is not required if, in lieu of response time testing, the response time of the safety equipment is verified by functional testing, calibration checks or other tests, or both. This is acceptable if it can be demonstmted that changes in response time beyond acceptable limits are accompanied by changes in performance characteristics which are detectable during mutine tests."
l On March 27, 1992, the BWR Owner's Group submitteo a Licensing Topical Report (LTR) prepared by the General Electric Company, NEDO-32013P, " System Analyses For Elimination Of Selected Response Time Testing Requirements", March 1992, foi NRC j
review. GSU believes the information contained themin justifies a one-time extension of l
the surveillance requirement intervals of TS Table 3.3.3-3, Item 1.
l The LTR pmvided justification for the elimination of selected BWR Response Time
[
Testing or Tests (RTT), as defined in the Instmment Society of America (ISA) Standard S67.06, from the plant TS SRs. The analyses included the affected instrumentation loops which could potentially impact the instrument loop response time. In addition, plant operating experiences were reviewed to identify response time failures and how they were detected. The failure modes identified were then evaluated to determine if the effect on response time would be detected by other testing requirements contained in the TS. Based l
on the analyses presented in the LTR, it was concluded that there were no failum modes which will affect the response time of the instrumentation loop which would not be i
i Attached to: RBG-39552 Page 57 l
t I
detected by other surveillances such as channel calibration (at least once per 31 days for trip setpoint calibration and at least once per 18 months for sensor calibration), channel functional tests (at least once per 31 days), channel checks (at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />),
j or other techniques. [It should be noted herein that the "other techniques" phrase applies j
to Rosemount tmnsmitter slow oil loss determination.] -
In addition, individual instmment channel response time delays for specific trip functions (on the order of a fraction of a second) have very little safety significance. Redundancy and diversity exist in most instrumentation trip functions (e.g., flux, water level, pressure).
j For the ECCS Actuation Instmmentation, the instmmentation response times are a small fraction of the overall response times of the actuating devices.
Based on the above, a one-time extension of the ECCS Actuation Instmmentation TS SR 4.3.3.3, Table 3.3.3-3, Item 1 msponse time testing suneillance intenral is justified.
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Attached to: RBG-39552 Page 58 ENCLOSURE 29 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION RESPONSE TIh1E TESTING FOR TECHNICAL SPECIFICATION SR 4.3.3.3, TABLE 3.3.3-3, ITEh! 2.a LOW PRESSURE CORE INJECTION h10DE OF RHR SYSTEhi, PUhfPS A AND B Eh1ERGENCY CORE COOLING SYSTEh! ACTUATION INSTRUMENTATION Technical Specification SR 4.3.3.3 requires the Response Time of the Emergency Com Cooling System (ECCS) trip functions shown on Table 3.3.3-3 be demonstrated to be within the limits at least once per 18 months. Each of the tests am to include at least one channel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific isolation trip system. Technical Specification Table 3.3.3-3, Item 2.a, I.aw Pressure Core Injection (LPCI) Mode Of RHR System - Pumps A and B, will become overdue prior to the beginning of RF-5 scheduled to begin April 16,1994. The response time test requires an extension for a nominal period of 47 days to reach the scheduled start of RF-5. However, in that LPCI is required to be OPERABLE in all Modes an extension for this item for a nominal period of 100 days is required to allow the testing to be ' completed by the scheduled end of RF-5.
Regulatory Guide 1.118 (Revision 2) states:
" Response time tes*ing of all safety related equipment, per se, is not required if, in lieu of response time testing, the response time of the safety equipment is verified by functional testing, calibration checks or other tests, or both. This is acceptable if it can be demonstrated that changes in response time beyond acceptable limits are accompanied by changes in performance characteristics which am detectable during routine tests."
On March 27, 1992, the BWR Owner's Group submitted a Licensing Topical Repon (LTR) prepamd by the General Electric Company, NEDO-32013P, " System Analyses For Elimination Of Selected Response Time Testing Requirements", March 1992, for NRC myiew. GSU believes the information contained therein justines a one-time extension of the suneillance requirement intervals of TS Table 3.3.3-3, Item 2.a.
The LTR provided justification for the elimination of selected BWR Response Time Testing or Tests (RTT), as denned in the Instmment Society of America (ISA) Standard S67.06, from the plant TS SRs. The analyses included the affected instmmentation loops which could potentially impact the instmment loop response time. In addition, plant operating experiences were reviewed to identify response time failures and how they were detected. The failum modes identiGed were then evaluated to determine if the effect on response time would be detected by other testing requirements contained in the TS. Based on the analyses presented in the LTR, it was concluded that there were no failure modes
Attached to: RBG-39552
.i Page 59 which will affect the response time of the instmmentation loop which would not be detected by other surveillances such as channel calibration (at least once per 31 days for l
trip setpoint calibration and at least once per 18 months for sensor calibration), channel l
functional tests (at least once per 31 days), channel checks (at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />),
{
or other techniques. [It should be noted hemin that the "other techniques" phrase applies j
to Rosemount transmitter slow oil loss determination.]
In addition, individual instrument channel resnonse time delays for specific trip functions (on the order of a fraction of a second) have very little safety significance. Redundancy and diversity exist in most instrumentation trip functions (e.g., flux, water level, pressure).
For the ECCS Actuation Instmmentation, the instnimentation response times are a small i
fmetion of the overall response times of the actuating devices.
Based on the above, a one-time extension of the ECCS Actuation Instrumentation TS SR 4.3.3.3, Table 3.3.3-3, Item 2.a response time testing surveillance interval is justified.
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Attached to: RBG-39552 Page 60 ENCLOSURE 30 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION I
RESPONSE TIME TESTING FOR TECHNICAL SPECIFICATION SR 4.3.3.3, TABLE 3.3.3-3, ITEM 2.b LOW PRESSURE CORE INJECTION MODE OF RHR SYSTEM, PUMP C EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION f
Technical Specification SR 4.3.3.3 requires the Response Time of the Emergency Com Cooling System (ECCS) trip functions shown on Table 3.3.3-3 be demonstrated to be within the limits at least once per 18 months. Each of the tests are to include at least one l
channel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific isolation trip system. Technical Specification Table 3.3.3-3, Item 2.b, Low Pressure Core Injection (LPCI) Mode Of RHR System - Pump C, will become overdue prior to the beginning of RF-5 scheduled to begin April 16,1994. The response time test requires an extension for a nominal period of 47 days to reach the scheduled start of RF-5. However, in that LPCI is mquired to be OPERABLE in all Modes an extension for this item for a nominal period of 100 days is required to allow the testing to be completed by the scheduled end of RF-5.
.l Regulatory Guide 1.118 (Revision 2) states:
" Response time testing of all safety related equipment, per se, is not required if, in lieu of response time testing, the response time of the safety equipment is verified by functional testing, calibmtion checks or other tests, or both. This is acceptable if it can be demonstrated that changes in response time beyond acceptable limits are accompanied by changes in performance chameteristics which l
are detectable during routine tests."
On March 27, 1992, the BWR Owner's Group submitted a Licensing Topical Report (LTR) prepared by the General Electric Company, NEDO-32013P, " System Alalyses For Elimination Of Selected Response Time Testing Requirements", March 1992, for NRC l
myiew. GSU believes the information contained therein justifies a one-time extension of 1
the surveillance requirement intervals of TS Table 3.3.3-3, Item 2.b.
i The LTR pmvided justification for the elimination of selected BWR Response Time.
Testing or Tests (RTT), as defined in the Instmment Society of America (ISA) Standard S67.06, from the plant TS SRs. The analyses included the affected instrumentation k> ops l
which could potentially impact the instrument loop response time. In addition, plant operating experiences were reviewed to identify response time failures and how they were detected. The failure modes identified were then evaluated to determine if the effect on response time would be detected by other testing requirements contained in the TS. Based on the analyses presented in the LTR, it was concluded that there were no failure modes which will affect the response time of the instmmentation loop which would not be l
Attached to: RBG-39552 i
Page 61 l
detected by other surveillances such as channel calibration (at least once per 31 days for trip setpoint calibration and at least once per 18 months for sensor calibration), channel functional tests (at least once per 31 days), channel checks (at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />),
or other techniques. [It should be noted herein that the "other techniques" phrase applies to Rosemount transmitter slow oil loss determination.]
In addition, individual instmment channel response time delays for specific trip functions (on the order of a fraction of a second) have very little safety significance. Redundancy and diversity exist in most instmmentation trip functions (e.g., flux, water level, pressure).
l For the ECCS Actuation Instmmentation, the instmmentation response times are a small l
fraction of the overall response times of the actuating devices.
Based on the above, a one-time extension of the ECCS Actuation Instrumentation TS SR 4.3.3.3, Table 3.3.3-3, Item 2.b response time testing surveillance interval is justified.
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Page 62 l
l ENCLOSURE 31 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION RESPONSE TIME TESTING FOR i
TECHNICAL SPECIFICATION SR 4.3.3.3, TABLE 3.3.3-3, ITEM 4 HIGH PRESSURE CORE SPRAY SYSTEM EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION l
Technical Specification SR 4.3.3.3 requires the Response Time of the Emergency Core Cooling System (ECCS) trip functions shown on Table 3.3.3-3 he demonstrated to bc within the limits at least once per 18 months. Each of the tests are to include at least one channel per trip system such that all channels are tested at least once every N times 18 months whem N is the total number of redundant channels in a specific isolation trip -
i system. Technical Specification Table 3.3.3-3, Item 4, High Pressum Core Spray System f
(HPCS), will become overdue prior to the beginning of RF-5 scheduled to begin April 16, i
1994. The response time test requires an extension for a nominal period of 31 days to i
reach the scheduled start of RF-5. However, in that HPCS is being used to pmvide
' defense in depth' during refueling operations, an extension for a nominal period of 61 l
days is required to allow the system to reach its outage window during RF-5.
j Regulatory Guide 1.118 (Revision 2) states:
" Response time testing of all safety related equipment, per se, is not required if, in lieu of response time testing, the response time of the safety equipment is verified by functional testing, calibration checks or other tests, or both. This is acceptable if it can be demonstrated that changes in response time beyond l
acceptable limits are accompanied by changes in performance characteristics which are detectable during routine tests."
l On March 27, 1992, the BWR Owner's Group submitted a Licensing Topical Report (LTR) prepared by the General Electric Company, NEDO-32013P, " System Analyses For i
Elimination Of Selected Response Time Testing Requirements", March 1992, for NRC eview. GSU believes the information contained therein justifies a one-time extension of the surveillance requirement intervals of TS Table 3.3.3-3, Item 4.
l The LTR provided justification for the elimination of selected BWR Response Time Testing or Tests (RTT), as defined in the Instmment Society of America (ISA) Standard l
S67.06, from the plant TS SRs. The analyses included the affected instrumentation loops which could potentially impact the instrument loop response time. In additica, plant operating experiences were reviewed to identify response time failures and how they were i
detected. The failure modes identified wem then evaluated to determine if the effect on response time would be detected by other testing requimments contained in the TS. Based on the analyses pmsented in the LTR, it was concluded that there were no failure modes which will affect the msponse time of the instrumentation loop which would not be
.. ~
i Attached to: RBG-39552 l
Page 63 I
detected by other surveillances such as channel calibration (at least once per 31 days for l
trip setpoint calibration and at least once per 18 months for sensor calibration), channel functional tests (at least once per 31 days), channel checks (at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />),
or other techniques. [It should be noted herein that the "other techniques" phrase apphes to Rosemount transmitter slow oil loss determination.]
l In addition, individual instmment channel response time delays for specific trip functions (on the order of a fractin of a second) have very little safety significance. Redundancy and diversity exist in most instrumentation trip functions (e.g., flux, water level, pressure).
For the ECCS Actuation Instrumentation, the instnimentation response times are a small l
fraction of the overall response times of the actuating devices.
l r
Based on the above, a one-time extension of the ECCS Actuation Instnimentation TS SR I
4.3.3.3, Table 3.3.3-3, Item 4 response time testing surveillance interval is justified.
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i Attached to: RBG-39552_
j Page 64 ENCLOSURE 32 i
JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION i
TECHNICAL SPECIFICATION SR 4.3.3.1, TABLE 4.3.3.1-1, ITEh! D.I.a l
LOSS OF POWER - DIVISION II 4.16-kv STANDBY BUS UNDERVOLTAGE i
(SUSTAINED UNDERVOLTAGE) l CALIBRATION i
EhiERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION
[
Technical Specification SR 4.3.3.1, Table 4.3.3.1-1, Item D.I.a requires the Division II l
4.16-kv Standby Bus Undervoltage (Sustained Undervoltage) relays be demonstrated i
OPERABLE by performance of a channel calibration at least once per 18 months (with a
{
maximum allowable extension of 4.5 months per TS 4.0.2). The undervoltage relay, Brown Boveri Type ITE 27H Relay, will require an extension of the SR interval cited in TS Table 4.3.3.1-1, Item D.I.a for a nominal period of 31 days to allow maching the l
scheduled system outage window during RF-5.
{
The Brown Boveri Instmetion Manual (IB 18.4.7-2, Issue E) for the Type ITE 27H miay I
specifies a mpeatability of 0.2 volt for a 10 volt variation. The text is worded to imply
{
that repeatability is equivalent to drift. The Type ITE 27H undervoltage relay is of the l
same design as the Type ITE 59H that was analyzed in the BWR Owner's Group l
Licensing Topical Report (LTR) concerning Calibration Interval Extension (NEDC-32160P). In this report it is sho vn that the Observed Inservice Difference (OISD)is very random between 18 and 24 mc.dhs. (OISD is the difference between the as-left and as-found values for a given calibntion interval). There are no trends that show an increase in OISD due to large observed drift (NEDC-32160P, Page E-55). It is important to note that all data for the LTR study of the Type ITE 59H relay was obtained from River Bend records; therefore, it is expected that the Type 27H relay would have the same drift j
characteristics.
l i
Based on the negligible changes in OISDs, the extension of the surveillance interval for 4
TS SR 4.3.3.1, Table 4.3.3.1-1, Item D.I.a calibration is justified.
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Attached to: RBG-39552 Page 65 ENCLOSURE 33 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION a
TECHNICAL SPECIFICATION SR 4.3.3.1, TABLE 4.3.3.1-1, ITEhi D.l.b LOSS OF POWER - DIVISION II 4.16-kv STANDBY BUS UNDERVOLTAGE (DEGRADED VOLTAGE)
CALIBRATION EMERGENCY CORE COOLING SYSTEh! ACTUATION INSTRUMENTATION l
Technical Speci6 cation SR 4.3.3.1, Table 4.3.3.1-1, Item D.I.b requires the Division II 4.16-kv Standby Bus Undervoltage (Degraded Voltage) relays be demonstrated OPERABLE by performance of a channel calibration at least once per 18 months (with a maximum allowable extension of 4.5 months per TS 4.0.2). The undenroltage relay, j
Brown Boveri Type ITE 27H Relay, will require an extension of the SR interval cited in TS Table 4.3.3.1-1, Item D.I.b for a nominal period of 31 days to allow reaching the' scheduled system outage window during RF-5.
The Brown Boveri Instruction Manual (IB 18.4.7-2, Issue E) for the Type ITE 27H relay.
l speci6es a repeatability of 0.2 volt for a 10 volt variation. The text is worded to imply that repeatability is equivalent to drift. The Type ITE 27H undervoltage relay is of the same design as the Type ITE 59H that was analyzed in the BWR Owner's Group Licensing Topical Report (LTR) concerning Calibration Interval Extension (NEDC-j 32160P). In this report it is shown that the Observed Inservice Difference (OISD)is very random between 18 and 24 months. (OISD.is the difference between the as-left and as-found values for a given calibration interval). There are no trends that show an increase s
in OISD due to large observed drift (NEDC-32160P, Page E-55). It is important to note i
that all data for the LTR study of the Type ITE 59H relay was obtained from River Bend j
records; therefore, it is expected that the Type 27H relay would have the same drift characteristics.
l l
Based on the negligible changes in OISDs, the extension of the surveillance interval for TS SR 4.3.3.1, Table 4.3.3.1-1, Item D.I.b calibration is justified.
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Attached to: RBG-39552 Page 66 ENCLOSURE 34 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION t
LOGIC SYSTEM FUNCTIONAL TESTING FOR j
TECHNICAL SPECIFICATION SR 4.3.3.1, TABLE 4.3.3.1-1, ITEM D.I.a LOSS OF POWER - DIVISION II 4.16-kv STANDBY BUS UNDERVOLTAGE (SUSTAINED UNDERVOLTAGE) l EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION f
Technical Specification SR 4.3.3.2 requires a LSFT and simulated automatic operation of-all channels of the Emergency Core Cooling System Actuation Instmmentation at least
.l once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). The Division II 4.16-lw Standby Bus Undervoltage (Sustained Unden'oltage) trip unit (TS Table 4.3.3.1-1, Item D.I.a) requires a surveillance interval extension for this trip unit's portion of the LSFT to reach the scheduled start of the
[
refueling outage (April 16,1994) for a nominal period of I day; however, since Division i
II power is required to be OPERABLE in all Modes of operation, an extension of the existing surveillance interval for a nominal period of 31 days is required for Item D.I.a i
to allow reaching the scheduled system outage window during RF-5.
t As stated in the NRC Safety Evaluation Report (dated August 2,1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance inten als from 18 to 24 months:
i
" Industry reliability studies for boiling water mactors (BWRs), prepared by the j
BWR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a mlay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall l
safety system unavailability."
l I
The evaluation above is applicable to RBS and the surveillance interval extension of 31 days is bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified.
+
Attached to: RBG-39552 Page 67 ENCLOSURE 35 IUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION-LOGIC SYSTEh! FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.3.1, TABLE 4.3.3.1-1, ITEh! D.I.b LOSS OF POWER - DIVISION II 4.16-kv STANDBY BUS UNDERVOLTAGE I
(DEGRADED VOLTAGE)
EhiERGENCY CORE COOLING SYSTEh! ACTUATION INSTRUMENTATION Technical Specification SR 4.3.3.2 requires a LSFT and simulated automatic operation of.
all channels of the Emergency Core Cooling System Actuation Instrumentation at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). The Division II 4.16-kv Standby Bus Undervoltage (Degraded 1
Vol' age) trip unit (TS Table 4.3.3.1-1, Item D.l.b) requires a surveillance interval i
ext.asion for this trip unit's ponion of the IEFT to reach the scheduled stan of the refueling outage (April 16,1994) for a nominal period of I day; however, since Division II power is required to be OPERABLE in all Modes of operation, an extension of the l
existing surveillance interval for a nominal period of 31 days is required for Item D.I.b to allow reaching the scheduled system outage window during RF-5.
As stated in the NRC Safety Evaluation Repon (dated August 2,1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling water reactors (BWRs), prepared by the BWR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability."
The evaluation above is applicable to RBS and the surveillance inten'al extension of 31 days is bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified.
Attached to: RBG-3955?
Page 68
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ENCLOSURE 36 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.7.4.1, TABLE 4.3.7.4-1, ITEM 1 REACTOR VESSEL PRESSURE TRANSMITTER CALIBRATION REMOTE SHUTDOWN MONITORING INSTRUMENTATION Technical Specification SR 4.3.7.4.1, Table 4.3.7.4-1, Item I requires the Remote Shutdown Monitoring System Reactor Vessel Pmssure Instmmentation be demonstrated OPERABLE by performance of a channel calibration at least once per 18 months (with a maximum allowable extension of 4.5 months per TS 4.0.2). The pressure transmitters, Rosemount Model 1152 transmitters, will require an extension of the SR interval cited in TS Table 4.3.7.4-1, Item 1 for a nominal period of 11 days to reach the scheduled stan of RF-5.
In February 1990, Rosemount published a repon, "30 Month Stability Specification For Rosemount Model 1152, 1153, 1154 Pressure Transmitters" (Rosemount Repon D8900126, Revision A) [ accepted by NRC Safety Evaluation Report dated August 2,1993.
on Peach Bottom Atomic Power Station, Units 2 and 3 docket]. This report supported the extension of the calibration interval for the transmitters from 18 mouths to 30 months i
based on a reduction in the drift allowance from 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. In addition, General Elec'ric (GE) setpoint calculations assumed 18 month calibration of the trip interval for trip units. However, the trip units are calibrated either monthly or quanerly, depending upon the TS requirement for channel functional testing. The GE setpoint calculations utilized a drift value of 0.23 %
SP (2 sigma) which is bounded by the required drift value of 0.13% SP (2 sigma).
The existing GE setpoint calculations for Rosemount transmitters and trip unit channels are bounding. There is adequate allowance in the calculations for 30 month drift and no potential impact on plant safety analyses (i.e., no analytic limit changes are mquired).
Therefore, the requested extension is justified.
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ENCLOSURE 37 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION l
TECHNICAL SPECIFICATION SR 4.3.7.4.1, TABLE 4.3.7.4-1, ITEM 2 i
REACTOR VESSEL WATER LEVEL TRANSMITTER CALIBRATION
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REMOTE SHUTDOWN MONITORING INSTRUMENTATION Technical Specification SR 4.3.7.4.1, Table 4.3.7.4-1, Item 2 requires the Remote Shutdown Monitoring System Reactor Vessel Water IzvelInstrumentation be demonstrated OPERABLE by perfonnance of a channel calibration at least once per 18 months (with a maximum allowable extension of 4.5 months per TS 4.0.2). The pressure transmitters, Rosemount Model 1152 tmnsmitters, will require an extension of the SR inteival cited in TS Table 4.3.7.4-1, Item 2 for a nominal period of I day to reach the scheduled start of RF-5.
In Febmary 1990, Rosemoent published a mport, "30 Month Stability Specification For Rosemount Model 1152, i153, 1154 Pressure Transmitters" (Rosemount Report D8900126, Revision A) [ accepted by NRC Safety Evaluation Repon dated August 2,1993 on Peach Bottom Atomic Power Station, Units 2 and 3 docket]. This eport supponed the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance fmm 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. In addition, General Electric (GE) setpoint i
calculations assumed 18 month calibration of the trip interval for trip units. However, the trip units are calibrated either monthly or quarterly, depending upon the TS requirement for channel ftmetional testing. The GE setpoint calculations utilized a drift value of 0.23 %
SP (2 sigma) which is bounded by the required drift value of 0.13% SP (2 sigma).
The existing GE setpoint calculations for Rosemount tmnsmitters and trip unit channels are bounding. There is adequate allowance in the calculations for 30 month drift and no potential impact on plant safety analyses (i.e., no analytic limit changes am required).
Therefore, the requested extension is justified.
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ENCLOSURE 38 l
JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION i
TECHNICAL SPECIFICATION SR 4.3.7.5, TABLE 4.3.7.5-1, ITEM 1 REACTOR VESSEL PRESSURE TRANSMITTER CALIBRATION i
ACCIDENT MONITORING INSTRUMENTATION Technical Specification SR 4.3.7.5, Table 4.3.7.5-1, Item I requires the Accident Monitoring System Reactor Vessel Pressure Instrumentation be demonstrated OPERABLE i
by performance of a channel calibration at least once per 18 months (with a maximum i
allowable extension of 4.5 months per TS 4.0.2). The pressure transmitters, Rosemount Model 1152 transmitters, will require an extension of the SR interval cited in TS Table 4.3.7.5-1 Item 1 for a nominal period of 10 days to reach the scheduled start of RF-5.
t In February 1990, Rosemount published a repon, "30 Month Stability Specification For Rosemount Model 1152, 1153, 1154 Pressure Transmitters" (Rosemount Repon l
D8900126, Revision A) [ accepted by NRC Safety Evaluation Repon dated August 2,1993 on Peach Bottom Atomic Power Station, Units 2 and 3 docket]. This repon supponed the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance from 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. In addition, General Electric (GE) setpoint calculations assumed 18 month calibration of the trip interval for trip units. However, the i
trip units are calibrated either monthly or quarterly, depending upon the TS requi ement l
for channel functional testing. The GE setpoint calculations utilized a drift value of 0.23 %
SP (2 sigma) which is bounded by the mquired drift value of 0.13% SP (2 sigma).
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The existing GE setpoint calculations for Rosemount transmitters and trip unit channels are l
bounding. There is adequate allowance in the calculations for 30 month drift and no potential impact on plant safety analyses (i.e., no analytic limit changes are required).
Theref ore, the requested extension is justified.
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i ENCLOSURE 39 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.7.5, TABLE 4.3.7.5-1, ITEM 9.b DRYWELL AREA RADIATION MONITOR CALIBRATION j
ACCIDENT MONITORING INSTRUMENTATION Technical SpeciGcation SR 4.3.7.5, Table 4.3.7.5-1, Item 9.b requires the Accident i
Monitoring System Drywell Area Radiation Monitor be demonstrated OPERABLE by perfonnance of a channel calibration at least once per 18 months (with a maximum -
allowable extension of 4.5 months per TS 4.0.2). The radiation monitors will require an extension of the SR :nterval cited in TS Table 4.3.7.5-1, Item 9.b for a nominal period of 2 days to reach the scheduled start of RF-5.
The radiation monitoring system at RBS consists ofion detectors and a microprocessor that have drift characteristics which are dominated by the power supplies. In that there is essentially no drift in the radiation monitoring system, an extension of the surveillance interval is justined.
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Page 72 ENCLOSURE 40 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION -
I TECHNICAL SPECIFICATION SR 4.3.9.1, TABLE 4.3.9.1-1, ITEM 2.a i
REACTOR VESSEL WATER LEVEL - HIGH LEVEL 8 TRANSMITTER CALIBRATION i
PLANT SYSTEMS ACTUATION INSTRUMENTATION Technical Specification SR 4.3.9.1, Table 4.3.9.1-1, Item 2.a mquires the Plant Systems
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Actuation Instmmentation, Reactor Vessel Water Level - High Level 8 for the Feedwater System / Main Turbine Trip System be demonstrated OPERABLE by performance of a 1
channel calibration at least once per 18 months (with a maximum allowable extension of 4.5 months per TS 4.0.2). The level transmitters, Rosemount Model 1152 tmnsmitters, l
will require an extension of the SR interval cited in TS Table 4.3.9.1-1, Item 2.a for a nominal period of 10 days to reach the scheduled stan of RF-5.
In February 1990, Rosemount published a repon, "30 Month Stability Specification For Rosemount Model 1152,1153,1154 Pressum Transmitters" (Rosemount Report D8900126, Revision A) [ accepted by NRC Safety Evaluation Report dated August 2,1993 -
on Peach Bottom Atomic Power Station, Units 2 and 3 docket]. This repon supported the exteasion of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance from 0.29% URL (2 sigma) for 18 months to l
0.20% URL (2 sigma) for 30 months. In addition, General Electric (GE) setpoint 5
calculations assumed 18 month calibmtion of the trip interval for trip units. However, the trip units are calibrated either monthly or quarterly, depending upon the TS requirement for channel functional testing. The GE setpoint calculations utilized a drift value of 0.23 %
i SP (2 sigma) which is bounded by the required drift value of 0.13% SP (2 sigma).
r The existing GE setpoint calculations for Rosemount transmitters and trip unit channels are f'
bounding. There is adequate allowance in the calculations for 30 month drift and no potential impact on plant safety analyses (i.e., no analytic limit changes am required).
I Therefore, the requested extension is justified.
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Page 73 ENCLOSURE 41 i
JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION LOGIC SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.9.1, TABLE 4.3.9.1-1, ITEM 2.a REACTOR VESSEL WATER LEVEL - HIGH LEVEL 8 PLANT SYSTEMS ACTUATION INSTRUMENTATION
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Technical Specification SR 4.3.9.2 requims a LSFT and simulated automatic opemtion of j
all channels of the Plant Systems Actuation Instrumentation be performed at least once per l
18 months (with a maximum allowable surveillance interval extension of 4.5 months per l
TS 4.0.2). The Reactor Vessel Water Level - High Ixvel 8 (TS Table 4.3.9.1-1, item l
2.a) requires a surveillance interval extension for this trip unit's ponion of the LSFT for
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a nominal period of 10 days to reach the scheduled beginning of RF-5.
As stated in the NRC Safety Evaluation Report (dated August 2,1993) mlated to extension f
of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling water reactors (BWRs), prepared by the l
BWR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that j
of the mechanical components, (e.g., pumps and valves), which are consequently i
tested on a more frey mt basis...Since the probability of a relay or contact failure i
is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall i
safety system unavailability."
The evaluation above is applicable to RBS and the suneillance interval extension of 31 days is bounded by the interval accepted on the Peach Bottom cocket; themfore, the surveillance interval extension is justified.
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Attached to: RBG-39552 Page 74 ENCLOSURE 42 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATIONS 4.8.2.l.c.1,4.8.2.1.c.2, AND 4.8.2.1.c.3 BATTERY INSPECTION DC SOURCES - OPERATING Technical Specifications 4.8.2.1.c.1,4.8.2.1.c.2, and 4.8.2.1.c.3 require that at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5. months per TS 4.0.2) the batteries be subjected to an inspection to detect deterioration, damage, corrosion and cell-to-cell resistance. The Division II batteries do not require a surveillance interval extension to reach the scheduled start of RF-5. However, in that a division of the batteries are required to be OPERABLE during Mode 4 and 5, the Division II batteries do require an extension of the interval for a nominal period of 29 days to reach its scheduled outage window to perform the inspections.
During the past performances of the Division I and II battery inspections, there were no abnonnalities found with regard to visual indication of battery conditions per TS SR 4.8.2.1.c.1 and TS SR 4.8.2.1.c.2. Resistance measurements meeting the requirements of TS SR 4.8.2.1.c.3 have been consistently met for the cell-to-cell and terminal connections. In addition, as a further indication of battery condition, past sen' ice tests of the batteries have shown that the batteries are capable of supplying and maintaining the necessary loading. Based on the past indication of battery condition, extension of the sun'cillance interval is justified.
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Page 75 I
i ENCLOSURE 43 l
JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION i
TECHNICAL SPECIFICATION SR 4.8.2.1.c.4 BATTERY CHARGER LOAD TEST -
j DC SOURCES - OPERATING i
Technical Specification SR 4.8.2.1.c.4 requires at least once per 18 months (with a maximum allowable extension of the interval of 4.5 months per TS 4.0.2) the battery chargers be verified to be capable, during tests of at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, of supplying 300
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amperes for chargers I A and IB (Division I and II, respectively). The Division II battery l
charger load test surveillance interval requires an extension for a nominal period of 11 days to reach the scheduled beginning of RF-5. The Division I battery charger load test r
surveillance interval will not expire before the scheduled start of RF-5, but will expire during the outage. In that at least one division of the de sources are required to be OPERABLE during Modes 4 and 5, a surveillance interval extension for both divisions of l
battery chargers is needed through the end of the outage so that the testing for the l
4 respective division may reach its scheduled window s.Jiin the outage. The 1 A (Division i
I) battery charger requires a surveillance interval extension within the outage for a nominal j
period of 20 days. The IB (Division II) battery charger requires a surveillance interval extension to the scheduled end of the outage for a nominal period of 64 days.
GSU has had good experience with the battery chargers at RBS. No charger has ever failed during a test which required the test to be renm. The charger load tests have always had satisfactory results with the voltage never falling below the test acceptance criteria.
l Based on the reliability indicated by the battery chargers at RBS and the short period for which the extension is being requested, a sun'eillance interval extension to the scheduled l
completion of RF-5 is justified.
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Page 76 l
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ENCLOSURE 44 1
f JUSTIFICATION FOR SURVEILLANCE INTERVAL ErfENSION TECHNICAL SPECIFICATION SR 4.8.2.1.d.1 BATTERY SERVICE TEST DC SOURCES - OPERATING
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Technical Specification SR 4.8.2.1.d.1 requires at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2 ) the l
batteries be subjected to a battery service test. The Division II battery will require a suneillance intenal extension for a nominal period of 9 days to reach the scheduled stan i
of RF-5. However, to allow the battery to remain in senrice during Mode 4 and 5 until it reaches its scheduled outage window, a suneillance interval extension for a nominal period of 38 days is mquired.
Past Division II battery senrice tests have always pmvided satisfactory results on their first performance at the 18-month test interval. Pilot cells are monitored weekly to provide a frequent measure of battery condition. Pilot cell non.bal since 1990 has not indicated a l
degmded battery condition. In addition, quarterly cell voltage, tempenture, and specific gravity have been taken and do not indicate any battery degradation. Based on the past service test results, nominal of weekly battery pilot cell surveillance results, and the short i
period for which the extension is being mquested, a sunreillance interval extension is justified.
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Page 77 j
i ENCLOSURE 45 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.8.2.1.e i
BATTERY PERFORMANCE TEST DC SOURCES - OPERATING i
Technical Specification SR 4.8.2.1.e mquires at least once per 60 months (with a maximum allowable surveillance interval extension of 15 months per TS 4.0.2) the battery
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capacity be verified to be at least 80% of the manufacturer's rating when subjected to a i
perfonnance discharge test. The Division III batteries will require a surveillance interval i
extension for a nominal period of 13 days to reach the scheduled stan of RF-5. However, i
since the HPCS is being relied upon to provide ' defense in depth' during refueling
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operations, a surveillance interval extension for a nominal period of 44 days is required l
for the Division III batteries to reach the scheduled HPCS out2ge window.
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The Division III (HPCS) battery was replaced in December of 1987. At that time the j
battery capacity was 93.6% cf the manufacturer's rating (rated at 100 Amp-hours for 8 l
hours). The Senice Discharge Test performed on 18-month intervals for this battery have i
always yielded acceptable results during its first perfonnance at the test inten>al. Pilot
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cells are monitored weekly to pmvide a frequent measure of battery condition. Pilot cell nominal since 1990 has not indicated a degraded battery condition. In addition, quarterly cell voltage, temperature, and specific gravity have been taken and do not indicate any battery degradation. Based on the past service test results, nominal of weekly battery pilot t
cell sun'eillance results, and the short period for which the extension is being requested, j
the surveillance interval extension is justified.
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ENCLOSURE 46 l
JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION l
TECHNICAL SPECIFICATION SR 4.8.4.3.b RPS ELECTRIC POWER MONITORING CHANNELS l
(ELECTRICAL PROTECTION ASSEMBLIES (EPAs))
-l CALIBRATION AND VERIFICATION OF SETPOINTS l
f Technical Spc fication SR 4.8.4.3.b requires the Reactor Protection System (RPS) l Eectric Power Monitoring Channels (Electrical Protection Assemblies (EPAs)) be demonstrated OPERABLE by performance of a channel calibration including simulated j
automatic actuation and verifying setpoints at least once per 18 months (with a maximum i
allowable extension of 4.5 months per TS 4.0.2). The EPAs am required in all Modes of operation and require extension of the surveillance interval to the end of the outage to i
support various windows within the outage. The existing surveillance interval expires after the scheduled start of RF-5, and requires an extension for a nominal period of 42 days to I
reach the scheduled end of RF-5.
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l The RPS EPAs monitor overvoltage, undervoltage, and underfrequency and have an
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overall error tolerance of the EPA function (which is effectively a trip pobt repeatability l
error) that does not have specific vendor accuracy and drift values. The overall ermr
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terms are +0%,-2.5 % (of trip setting) for overvoltage, +2.5 %,-0% for undervoltage, and i
+2%,-0% for underfrequency.
These overall errors are entered in the setpoint calculations as instrument accuracy. with the channel drift assigned a value of zero.
l Additional basis which supports this assumption is the fact that additionallabomtory testing
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of the EPAs, utilizing an updated logic card smilar to that used at RBS, has exhibited little i
or no drift.
Based on the above, the existing setpoints will not be affected by the extension of the l
surveillance interval and is therefore acceptable.
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i A'ITACHMENT 2 NO SIGNIFICANT HAZARDS CONSIDERATION In accordance with the requirements of 10 CFR 50.92, the following discussion is l
provided in support-of the determination that no significant hazards are cmated or i
increased by the change requested in the submittal.
t 1.
The proposed change would not significantly increase the probability or consequences of an accident previously evaluated because:
The proposed TS change requests a one-time only extension of the surveillance-f inten'als for TSs related to: a) RPS Instmmentation calibration, LSFTs, and response time testing; b) Isolation Actuation System Instmmentation calibration, LSFrs, and response time testing; c) ECCS Actuation Instmmentation calibration, LSFTs, and response time testing; d) Control Rod Block Instrumentation calibration; e) Remote Shutdown Monitoring Instrumentation calibation; f)
Accident Monitoring Instrumentation calibration; g) Feedwater System / Main i
Turbine Trip System Instrumentation calibration and LSFT; h) Primary Containment RWCU automatic isolation valve actuation; i) de Sources (Batteries i
and Chargers) inspection, load tests, service tests, and performance tests; and j)
RPS Electric Power Monitoring channels (Electrical Protection Assemblies) calibration.
t Also pmposed is a reestablishment of the baseline for the "N times 18 months" cumulative surveillance interval for response time testing.
Based on the discussion in the License Amendment Request which shows:
a) Rosemount transmitter calibration period extension is acceptable based on Rosemount Report D8900126, Revision A which supponed extension of the
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calibration interval from 18 months to 30 months based on a reduction in the drift i
allowance; b) Gould transmitter calibration period extension is acceptable based on calibration and drift data which supports extension past 18 months; c) The Brown Boveri ITE relay calibration period extension is acceptable based on t
calibration and drift data which suppons extension past 18 months; d) LSFT interval extension is acceptable based on the NRC Safety Evaluation Repon (Peach Bottom Atomic Power Plant, Units 2 and 3, dated August 2,1993) l which supponed extension of the interval for LSFT from 18 to 24 months. This l
was based on the small probability of relay or contact failure relative to mechanical l
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E Attached to: RBG-39552 Page 80 i
component failure probability and, therefore, the increase in LSFT interval represented no significant change in the ovemil safety system unavailability; e) Response time testing interval extension for Isolation Actuation and ECCS Actuation Instrumentation channels is acceptable based on the BWR Owner's Group i
Licensing Topical Report NEDO-32013P (hiarch 1992) which providedjustification for elimination of response time testing and, therefore, provides a suitable argument for extending the interval for a short period of time; f) Response time testing interval extension for RPS Instrumentation channels is acceptable because: i) there are redundant sensors that can initiate the scram function; ii) one-out-of-two redundancy exists in every individual instrument channel within each trip function; iii) several redundant and diverse instmment channels are provided which can detect and generate a scram signal; iv) the failure probability is a small fraction of the total control rod insertion (scram) failum probability; v) failure of instrumentation in the sluggish mode is a small fraction of its overall failure modes; and vi) NRC Safety Evaluation Report dated August 2,1993 (Peach Bottom Atomic Power Station, Units 2 and 3 docket) has previously provided approval of the extension of the RPS response time testing surveillance interval from 18 to 24 months.
g) Response time testing intenral extension for hiain Steam Line Isolation due to l
high radiation is acceptable because of NRC acceptance (dated hiay 15,1991) of the BWR Owner's Group LTR which justified removal of the hiSIV and mactor scram functions of the Main Steam Line Radiation hionitors, and this document's applicability to RBS.
I h) Response time testing interval extension of the hiain Steam Line Isolation due to conditions other than high radiation is acceptable because i) redundancy and diversity exist in individual instmment channels within a trip function; ii) instmmentation response time is a small fraction of the overall response time of the actuating device; iii) instrumentation failure probability is an very small portion of the total hiSIV failure probability; and, iv) failure of instrumentation in the sluggish responding mode is a small fraction of its overall failure modes.
i) RWCU Containment Isolation Valve actuation interval extension is acceptable based on: i) redundancy provided in the design of the penetrations; ii) the periodic l
testing of the valves performed during power operation; and, iii) the short period of time the interval is being extended.
I j) Battery and Cha ger inspection and load test interval extensions are acceptable based on: i) the past history of testing for the batteries and chargers has been ' good; ii) the nominal of weekly pilot cell data has indicated no degradation; and, iii) the
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short period of time the interval is being extended is small.
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Attached to: RBG-39552 Page 81 k) The chanp to TS 4.8.2.2 is editorial and administrative only to achieve consistency in the TS. This meets the meets the example of a change not likely to involve significant hazards considerations as cited in the Statements of t
Consideration published in the Federal Register (48 FR 14864).
- 1) The Electrical Protection Assembly (EPA) calibration interval extension is acceptable based on the inherent lack of drift for the EPAs and accuracy of the system logic.
m) The reestablishment of the baseline for the "N times 18 month" cumulative j
surveillance interval for response time testing is acceptable in that the extension of the cumulative interval would not be for more than the individual extension requested and justified herein.
Therefore, from the above it is shown that the proposed change will not significantly inemase the probability or consequences of an accident pmviously i
evaluated.
i 2.
The proposed change would not create the possibility of a new or different kind of accident from any accident previously evaluated because:
The proposed TS change requests a one-time extension of the surveillance intervals for the instrument calibration, instrument channel LSFT and response time testing, RWCU containment isolation valve actuation, battery and charger inspection and load testing, and EPA calibration. The proposed changes do not necessitate a physical alteration to the plant (no new or different type of equipment will be installed). In that the requested extension durations are small as compared to the overall interval allowed by TS, drift data supports extension of calibration intervals, NRC and industry evaluations support extension of LSFT, industry evaluations and redundancy in system design support extension of response time testing, past testing and periodic testing provides confidence of no effect on equipment availability by extending the surveillance interval, the change does not create the possibility of a new or different kind of accident from any accident i
previously evaluated.
The change to TS 4.8.2.2 is editorial and administrative only to achieve consistency in the TS. This meets the meets the example of a change not likely to involve significant hazards considerations as cited in the Statements of Consideration published in the Federal Register (48 FR 14864).
In addition, the requested reestablishment of the baseline at RF-5 for the "N times s
18 months" cumulative surveillance interval for response time testing is acceptable in that the cumulative surveillance interval will not be extended by more than that l
which is proposed for individual response time tests during RF-5. The individual
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Attached to: kBG-39552 Page 82 response time test surveillance interval extensions have been justified herein. The justification for individual response time test surveillance interval extensions applies to the cumulative surveillance interval extension which is requested and will be granted by allowing reestablishment of the baseline of Oc "N times 18 months" i
surveillance interval to the response time tevir.g dates for those response time tests to be performed during RF-5. The proposed changes do not necessitate a physical alteration to the plant (no new or different tyne of equipment will be installed).
Therefore, the change does not create the possibility of a new or different kind of accident.
3.
The pmposed change wil! not involve a significant reduction in the margin of
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safety because:
i The proposed TS change requests a one-time extension of the surveillance inten als for the instmment calibration, instmment channel LSFT and response time testing, RWCU containment isolation valve actuation, battery and charger inspection and
-l load testing, and EPA calibration. In that the requested extension durations are i
small as compared to the overall interval allowed by TS, drift data supports t
extension of calibration intervals, NRC and industry evaluations support extension of LSFT, industry evaluations and redundancy in system design support extension of response time testing, past testing and periodic testing provides confidence of no effect on equipment availability by exterding the surveillance interval, the change does not involve a significant reduction in the margin of safety.
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The change to TS 4.8.2.2 is editocial and administrative only to achieve consistency in the TS. This meets the meets the example of a change not likely to involve significant hazards considerations as cited in the Statements of Consideration published in the Federal Register (48 FR 14864).
l In addition, the requested reestablishment of the baseline at RF-5 for the "N times 18 months" cumulative surveillance interval for response time testing is acceptable in that the cumulative surveillance interval will not be extended by more than that l
which is proposed for individual Irsponse time tests during RF-5. The individual response time test surveillance interval extensions have been justified herein. The justification for individual response time test sun'eillance interval extensions applies
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to the cumulative surveillance interval extension which is requested and will be granted by allowing reestablishment of the baseline of the "N times 18 months" l
surveillance interval to the response time testing dates for those response time tests i
10 be perfomied during RF-5. Therefore, the change does not involve a significant reduction in the margin of safety.
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