NRC-98-0040, Application for Amend to License NPF-43,requesting to Allow one-time Extension of Surveillance Intervals to Delay Shutdown for Upcoming Sixth Refueling Outage (RFO6) Until 980904
| ML20236F061 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 06/26/1998 |
| From: | Gipson D DETROIT EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20236F064 | List: |
| References | |
| CON-NRC-98-0040, CON-NRC-98-40 NUDOCS 9807020008 | |
| Download: ML20236F061 (96) | |
Text
_ _ - _.. _ _ - _ _ - -
c 7'3 Dougits R. Gipson Senior Vice President, Nurlear Generation fermi 2 6400 North Dixie llwy, Newport, Michigan 48166 Tel: 313.586.5201 Fax: 313.586.41"2 Detroit Edison i
10 CFR 50.92 June 26,1998 hT C-98-0040 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington D C 20555-0001
References:
- 1) Fermi 2 i
NRC Docket No. 50-341 NRC License No. NPF-43
- 2) Detroit Edison Letter to NRC,"One-Time Technical Specification Revision to Allow Extension of the Fermi 2 Operating Cycle," NRC-95-0096, dated September 20,1995
- 3) 1998 Summer Assessment," Reliability of Bulk Electricity Supply in North America," North American Electric Reliability
\\
}k Council, May 1998
Subject:
One-Time Technical Specifications Revision to Allow Extension of the Current Fermi 2 Oneratine Cycle 6 The Detroit Edison Company (Detroit Edison) hereby files an application to amend o\\
the Fermi 2 Technical Specifications to allow extension of the current Operating M
jpycle 6 This application is filed to revise applicable Technical Specifications i
related to system testing, instrumentation calibration, component inspection, j
component testing, response time testing, and Logic System Functional Tests to I
allow a one-time extension of the surveillance intervals. The pmposed extensions are requested to delay shutdown for the upcoming sixth Refuelir,g Outage (RFO6) until September 4,1998.
9807020008 990626 PDR ADOCK 05000341 P
.a A DTE Energy Company
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USNRC L'
NRC-98-0040 Page 2 l-Fermi 2 is scheduled to be in the refueling outage during August of 1998. Detroit Edison normally experiences its maximum peak load in the summer months, particularly during the month of August, and even with the availability of Fermi 2 must purchase additional capacity from extemal sources in order to provide a reliable supply of power to its customers. Additionally,if the Midwest region experiences L-warmer than normal temperatures this summer, available power sources throughout the region could be inadequate to meet the resultant higher demand.
As discussed in Reference 3, the extended outage at the D. C. Cook nuclear units as well as significant potential unavailability of power from the neighboring utilities L
such as Commonwealth Edison, Illinois Power, and Ontario Hydro will make it -
difficult for Detroit Edison to replace the capacity lost when Fermi 2 is unavailable L
this summer. Furthermore, the loss of this additional 1100 MW of Fermi 2 capacity l
can only further negatively impact the overall reliability for the Midwest region as a whole. Additionally, there are major concerns that any extemal power purchases l-made this summer will be curtailed due to " bottlenecks" in the transmission system, L
which has already occurred on several occasions during relatively mild summer days in the month ofJune 1998.
The transmission interconnections between electrical utilities were not built to r
facilitate the delivery of a major portion of a local utility's power demand. The fact that most regional utilities will at times during the summer need to import power could lead to " congestion" at the transmission interconnections during times of peak
[~
' demand. Moreover, as stated previously, curtailments in power imports have already
. taken place in the region during May and June of 1998 as a result of the new North g
L American Electric Reliability Council (NERC) policies and procedures for operating
}
the transmission system ofinterconnected electric utilities, and these curtailments are only expected to increase as temperatures and load increase further during the months ofJuly and August 1998.
Consequently, both the availability of external power and its delivery through the transmission interconnections may become questionable during the peak demand L
period of this summer. Postponing the outage at Fermi 2 until September will significantly alleviate this problem, and will allow Detroit Edison to be able to serve its customers in a much more reliable manner, and will help in improving the reliability for the Midwest region as a whole. Having the 1100 MW of Fermi 2 L
capacity available to serve peak summer load will significantly lessen the Company's L
concerns about the availability of external power and the reliance on transtnission interconnections. Therefore, Detroit Edison is proposing revisions to Fenni 2 Tecl nical Specifications which would allow plant shutdown to be delayed for
- approximately two weeks until September 4,1998.
l-i' L
s USNRC NRC-98-0040 Page 3 l
Similar surveillance extension requests to support postponement of refueling outages have previously been approved for Fermi 2 (Reference 2), River Bend, Nine Mile l
Point 2, Perry, and D.C. Cook. Also, several plants have received approvals to extend their operating cycle permanently to 24 months which allows a maximum t
surveillance interval of 30 months including the 25% allowance for scheduling flexibility. The longest extension expected of 61 days is much less than the l
additional time period allowed by the permanent 24 month cycle. The longest extension request of 61 days, for a two week delay in the outage, represents the need for extending surveillance to envelop the entire duration of the outage. The surveillance being extended to the end of the outage are expected to be completed much earlier during the outage as per the:r planned schedules.
i provides a table which lists the Technical Specifications for which i
extensions are being requested and the date required for extension. The table provides a cross reference to the Attachment 1 enclosure whichjustifies the extension. Additionally, Attachment 1 provides descriptions and discussions for Technical Specifications which require extension. Enclosures I through 30 to provide thejustification for the proposed extensions to the Technical Specifications. Attachment 2 provides the No Significant Hazards Consideration discussion. The revised Technical Specification pages are provided in Attachment 3.
Based on this request Detroit Edison would plan to commence shutdown for the sixth l
refueling outage on September 4,1998. The first surveillance requirement becomes I
overdue on August 23,1998, provided the EDG inspection which becomes overdue on August 19,1998 is performed on-line prior to that date. Therefore, approval of this extension is requested by August 1,1998. Prompt review of this proposal is requested to allow appropriate time for orderly preparations of the outage work required during RFO6.
Detroit Edison has reviewed the proposed TS changes against the criteria of 10 CFR 50.92 and 10 CFR 51.22 for environmental considerations. The proposed changes do not involve a significant hazards consideration, nor significantly change the types or significantly increase the amounts of effluents that may be released offsite, nor significantly increase individual or cumulative occupational radiation exposures.
Based on the foregoing, i>etroit Edison concludes that the proposed TS changes meet the criteria provided in 10 CFR $1.22(c) (9) for a categorical exclusion from the requirements for an Environmental Impact Statement or an Environmental Assessment. In accordance with 10 CFR 50.91, Detroit Edison is providing a copy of this letter to the State of Michigan.
l L____
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.USNRC.
NRC-98-0040 Page 4
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d Detroit Edison requests that the NRC approve and issue these changes by August 1, 1998, with implementation within 30 days following NRC approval. All surveillance tests for which an extension is requested will be completed prior to entry into Operational Condition 3 upon start-up.
1 If you have any questions or require additional information, please contact i
i
- Mr. Norman K. Peterson, Director-Nuclear Licensing at (734) 586-4258.
l-I Sincerely, I-I i
Attachments l
Enclosure l
i l
cc: B. L. Burgess j
. G. A. Harris -
]
A. J. Kugler Regional Administrator, Region III Supervisor, Electric Operators
. Michigan Public Service Commission l
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USNRC NRC-98-0040 Page 5 i
1 i
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I, DOUGLAS R. GIPSON, do hereby affinn that the foregoing statements are based i
on facts and circumstances which are true and accurate to the best of my knowledge and belief.
DOUGLAS R. DIPSON Senior Vice President 9
ILO.1998 before me personally On this day of appeared Douglas R. Gipson, being first[puly sworn and says that he executed the foregoing as his free act and deed.
kAf>Ao L/ WY Motary Public I N08AUE A. ARMETTA i
MetMMES Me l
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Attachment I to NRC-98-0040 Page 1 j
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l ATTACIIMENT 1 FERMI 2 NRC DOCKET NO. 50-341 OPERATING LICENSE NO. NPF-43 REQUEST TO REVISE TECIINICAL SPECIFICATIONS "ONE-TIME TECIINICAL SPECIFICATIONS REVISION TO ALLOW EXTENSION OF TIIE FERMI 2 OPERATING CYCLE" DESCRIPTION AND EVALUATION I
OF TIIE PROPOSED CHANGES i
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i Attachment I to NRC-98-0040 Page 6 l
DESCRIPTION AND EVALUATION OF THE PROPOSED CHANGES (S) 1 1
1-l BACKGROUND:
1 l
l As described in the cover letter, both the availability of external power and its delivery through the transmission interconnections may become questionable during the peak demand period of this summer. Postponing the outage until September will significantly alleviate this problem, and will allow Detroit Edison to be able to meet its power dema.nds without being concerned about availability of extemal power or its delivery during the peak periods. Therefore, Detroit Edison is proposing revisions to Fermi 2 Technical Specifications which would allow plant shutdown to be delayed until September 4,1998.
1 In order to minimize the changes to the individual pages of the Technical Specifications, a revision is proposed for Section 4.0.2 which would replace the reference Tables 4.0.2-1 and 4.0.2-2 showing the affected specifications and new completion date for these surveillance. The Fermi 2 Technical Specifications (TS) require the performance of Operating Cycle surveillance requirements (SRs), including: instrumentation calibration, response time testing and Logic System Functional Tests (LSFT). Examples of the TS to which these SRs apply include the reactor protection system instrumentation, isolation actuation instrumentation, emergency core cooling system (ECCS) actuation instrumentation, control rod block instrumentation, remote shutdown monitoring I
instrumentation, accident monitoring instrumentation, plant systems actuation instrumentation, Appendix R instrumentation and reactor protection system 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 least once every 18 months. TS 4.0.2 allows a 25% extension of the surveillance interval to 22.5 months,
)
if required, to provide some flexibility in cycle lengths. In addition, system testing l
involving valve operability or performance testing, including Primary Containment Isolation Valve actuation and leak testing and AC and DC system surveillance, are required to be performed normally at refueling intervals but at least once every 18 months. With the exception of Primary Containment Isolation Valves tested under the j
requirements of10 CFR 50 Appendix J, TS 4.0.2 allows a 25% extension of the surveillance interval to 22.5 months, if required, to provide some flexibility in cycle lengths.
A one-time change is being requested to c-tend the surveillance intervals for the above cited TS SRs. Extensions are being requested for surveillance which are due prior to the start of the scheduled outage for equipment only needed during plant operations.
l Extensions are being requested for surveillance due prior to the end of the outage for l
1 l
~
l
}
Attachment I to NRC-98-0040 Page 7 l
equipment needed during the outage. The earlier date of September 14,1998 corresponds
{
to the anticipated shutdown date of September 4,1998 including flexibility in shutdown I
l dates and time to cooldown to Operational Condition 4. The later date of October 18, 1998 corresponds to the expected end date of the refueling outage. All smveillance tests for which an extension is requested will be completed prior to entry into Operational
{
Condition 3 upon start-up. The extension for equipment or systems required during Operational Conditions 1,2, and 3 is requested until at least September 14,1998, in order to allow time for cooldown to Operational Condition 4 when a majority of the Limiting Conditions for Operation for the respective SRs are not applicable. This will also allow l
entry into Operational Condition 5 and initiation of core alteration. Though some l
equipment is only required to be operable in Operational Condition 1 or Operational Conditions 1 and 2, rather than in Operational Conditions 1,2, and 3, surveillance for equipment not required to be operable during the outage are grouped together in a table l
extending the surveillance test intervals to September 14,1998. This grouping reduces the complexity of the proposed change and eliminates the need to specify in the Technical Specification precisely on which date the mode changes will occur. The longest extension request of 61 days, for a two week delay in the outage, represents the need for extending surveillance to envelop the entire duration of the outage. The surveillance j
being extended to the end of the outage are expected to be completed much earlier during the outage as per their planned schedules.
l An additional extension is being requested for SRs wherein the applicability extends to Operational Conditions 4,5 or other shutdown situations. These SRs associated with the l
TS are proposed to be extended until the end of the refueling outage, scheduled for October 18,1998, to support the current schedule and system windows already established, to provide defense in depth during the shutdown period and to provide flexibility in outage scheduling. Also, some specific line items that are not required to be operational during Operational Condition 4 or 5 have a requested extension date of the
[
end of the outage. The reason for this is that the surveillance procedure that tests these components also fulfills the surveillance requirements for components that are required to i
be operational during Operational Conditions 4 or 5.
l It is also requested that the "N times 18 months" cumulative surveillance interval for l
various response 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 l
perfomied during RFO6. This re-establislunent of the baseline will ensure that future response time testing intervals, with respect to the cumulative "N times 18 months" interval, will not become late due to the interval extensions that are required for RFO6.
The samejustification provided in this amendment request for the individual response time surveillance interval extensions applies to the "N times 18 months" cumulative surveillance interval extensions because the cumulative surveillance interval would not be extended by more than that being requested for individual response time tests.
Ic
o
' to NRC-98-0040 Page 8 DESCRIPTION:
The following sections provide a description of each of the proposed TS changes cited above. Enclosures to this Attachment describe thejustification for each proposed extension. Evaluations ofinstrument drifVcalibration data and surveillance testing failure experience are included in these justifications where applicable.
The Surveillance Test history data base was reviewed to identify any tests during the last refueling outage which were coded as either equipment failure or " partially complete".
The partially complete category includes not only component failures causing an interruption of the test, but it would also include procedure problems, plant conditions that might have precluded further testing, etc. Based on the 'ist of" failures" identified by this search, the surveillance test records were retrieved and each was evaluate <l. The evaluation categorized the failure modes for the components. Failures that would have no effect on the safety function were eliminated. Where the failure may have impacted the safety function, further evaluations were conducted that considered factors such as: (1) whether a similar failure would be detected during the extended cycle by other more frequently conducted tests (such as functional tests), PM activities, IST or other monitoring activities; (2) whether the failure was caused by a special event, occurrence or maintenance activity which has not occurred during the current cycle; and (3) whether or not the failure could have been time dependent and thus be relevant to the proposed extension.
ASME Code Class Testing (13 Section 4.0.5)
As stated in TS Bases Section 3/4.0, Section 4.0.5 establishes the requirement that inservice inspection of ASME Code Class 1,2, and 3 components and inservice testing of ASME Code Class 1,2, and 3 pumps and valves shall be performed in accordance with a periodically updated version of Section XI of the ASME Boiler and Pressure Vessel Code and Addenda as required by 10 CFR 50.55a. These requirements apply except when relief has been provided in writing by the Commission.
This specification includes a clarification of the frequencies for performing the inservice inspection and testing activities required by Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda. This clarification is provided to ensure consistency in surveillance intervals throughout the TS and to remove any ambiguities relative to the frequencies for performing the required inservice inspection and testing activities.
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. to NRC-98-0040 Page 9 TS Section 4.0.5, requires that surveillance requirements for inservice inspection and testing of ASME Code Class 1,2, & 3 components shall be applicable as follows:
Inservice inspection of ASME Code Class 1,2, and 3 components and a.
inservice testing of ASME Code Class 1,2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).
b.
Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspection i
and testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications:
ASME Boiler and Pressure Vessel Required frequencies Code and applicable Addenda for perfonning inservice terminology for inservice inspection and testing inspection and testing activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days The provisions of TS 4.0.2 are applicable to the above required c.
frequencies for performing inservice inspection and testing activities.
d.
Performance of the above inservice inspection and testing activities shall be in addition to other specified SR.
e.
Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any TS.
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i f.
The Inservice Inspection (NDE) Program for piping identified in l~
NRC Generic Letter 88-01, dated January 25,1988, "NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping", shall be performed in accordance with the staff positions on schedule, methods and personnel, and sample expansion included in this generic letter.
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Attachment I to NRC-98-0040 Page 10 Where specific TS extensions are required for ASME components (e.g., excess flow check valves) these extensions are discussed andjustified with the TS sections. Although not specifically identified in Fermi 2 Technical Specifications, the Code specified 2 year requirement is being treated as a 731 day surveillance frequency and is subject to a 25%
extension. This has already been determined to be acceptable per NUREG-1482, Section 6.2. The surveillance associated with Technical Specification 4.0.5 are associated with the Pump and Valve testing, and in addition to Technical Specification 4.4.8, address piping, component support, system leakage, and weld examination. The ASME requirements for these surveillance specify a refueling or 24 month evaluation frequency. However, since the normal Fermi 2 operating cycle is 18 months, the inspections have been performed on a 18 month cycle. Enclosure 20 provides discussions whichjustify a no impact on plant safety for extending Technical Specification 4.0.5 for Inservice Testing of the Non-Interruptible Air Supply (NIAS) isolation valves, which becomes overdue on October 2,1998, for 17 days to the end of the refueling outage.
Scram Accumulator (TS Section 4.1.3.5.b.2)
As stated in UFSAR Section 4.5.2.2.2.3 the Control Rod Drive (CRD) hydraulic system supplies and controls the pressure and flow to and frons the drives through hydraulic control units (HCUs). The basic components in each HCU are manual, pneumatic, and electrical valves; an accumulator; related piping; electrical connections; Illters; and instrumentation. The scram accumulator stores sufficient energy to fully insect a control rod at lower Reactor Pressure Vessel (RPV) pressures. At higher RPV pressmes, the accumulator pressure is assisted or replaced by RPV pressure.
Technical Specification 4.1.3.5.b.2 requires demonstrating that each control rod scram accumulator is operable at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2) by measuring and recording the time for at least 10 minutes that each individual accumulator check valve maintains the associated accumulator pressure above the alarm set point with no control rod drive pump operating. This item becomes overdue on October 2,1998 and requires an extension of 17 days to the end of the refueling outage. Thejustification for this extension is provided in Enclosure 1.
Standby Liquid Control System (TS Section 4.1.5.d items 1 through 3)
As stated in UFS AR Section 4.5.2.4.2 the Standby Liquid Control System (SLCS),
which is manually initiated from the main control room, injects boron neutron absorber solution into the reactor if the operator believes the reactor cannot be shutdown or kept shutdown with control rods. The SLCS function is a backup to control rods and is to maintain the reactor shutdown under all conditions without control rods. The boron solution tank, the test tank, the two positive-displacement pumps, the two explosive valves, and associated local valves and controls are mounted in the reactor building.
Borated liquid is piped into the Reactor Pressure Vessel (RPV) and discharged near the
Attachment I to NRC-98-0040 Page 11 bottom of the core shroud so it mixes with the cooling water rising through the core. The boron absorbs thermal neutrons and thereby terminates the nuclear fission chain reaction in the fuel.
Technical Specification 4.1.5.d items 1 through 3 state the standby liquid control system shall 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:
1.
Initiating one of the standby liquid control system loops, including an explosive valve, and verifying that a flow path from the pumps to the reactor pressure vessel is available by pumping demineralized water into the reactor vessel. The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch which has been certified by having one charge of that batch successfully fired. Both injection loops shall be tested in 36 months. This item becomes overdue on August 29,1998 and requires an extension of 51 days to the end of the refueling outage. Thejustification for this extension is provided in Enclosure 2.
2.
Demonstrating that the pump relief valve setpoint is less than or equal to 1400 psig and verifying that the relief valve does not actuate during recirculation to the test tank. This item becomes overdue on September 24,1998 and requires an extension of 25 days to the end of the refueling outage. Thejustification for this extension is provided in Enclosure 2.
3.
Demonstrating that all piping between the storage tank and the explosive valve is unblocked by pumping from the storage tank to the test tank and then draining and flushing the piping with demineralized water. This item becomes overdue on 4
August 29,1998 and requires an extension of 51 days to the end of the refueling outage. The justification for this extension is provided in Enclosure 2.
Reactor Protection System (TS Section 4.3.1.1)
I The RPS instrumentation and control initiates an automatic reactor shutdown (scram) if monitored system variables exceed pre-established limits. This action prevents fuel damage and limits system pressure. Technical Specification 4.3.1.1 requires that each reactor protection system instrumentation channel be demonstrated operable by the performance of a Channel Functional test or a Channel Calibration for the operational I
conditions at the frequencies shown in Table 4.3.1.1-1. The following discussions are provided for the RPS signals of concern for this one-time Technical Specification extension.
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Attachment I to NRC-98-0040 Page 12 Item 3 Reactor Pressure As stated in UFSAR Section 7.2.1.1.5.4, a nuclear system pressure increase during reactor operation compresses the steam voids and results in a positive reactivity insertion. This causes increased core heat generation that could lead to fuel failure and system overpressurization. A scram counteracts a pressure increase by quickly reducing core fission heat generation. The nuclear system high-pressure scram setting is chosen slightly above the RPV maximum normal operating pressure to permit normal operation without spurious scram, yet provides a wide margin to the maximum allowable nuclear system pressure.
Item 3 for Table 4.3.1.1-1 requires channel calibration at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This item becomes overdue on September 2,1998 and requires an extension of 13 days to reach September 14,1998. Thejustification for this extension is provided in Enclosure 3.
Item 4 Reactor Vessel Low Water Level - Level 3
)
As stated in UFSAR Section 7.2.1.1.5.4, low water level in the RPV indicates that the fuel is in danger of being inadequately cooled. Decreasing the water level while the reactor is operating at power decreases the reactor coolant inlet subcooling. The effect is the same as raising feedwater temperature. Should f
Reactor Pressure Vessel water level decrease too far, fuel damage could result. A reactor scram protects the fuel by reducing the fission heat generation within the core.
Item 4 for Table 4.3.1.1-1 requires channel calibration at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This item becomes overdue on August 24,1998 and requires an extension of 22 days to reach September 14,1998. Thejustification for this extension is provided in Enclosure 3.
Item 6 Main Steam Line High Radiation As stated in UFSAR Section 7.2.1.1.5.4, high radiation in the vicinity of the main steam lines may indicate a gross fuel failure in the core. When high radiation is detected near the steam line, a scram is initiated to limit the release of fission products from the fuel.
Item 6 for Table 4.3.1.1-1 requires channel calibration at least once per 18 months
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Attachment I to NRC-98-0040 Page 13 (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This item becomes overdue on August 25,1998 and requires an extension of 21 days to reach September 14,1998. The justification for this extension is provided in Enclosure 3.
Item 7 Drywell Pressure As stated in UFSAR Section 7.2.1.1.5.4, high pressure inside the primary containment may indicate a break in the nuclear system process barrier. It is prudent to scram the reactor in such a situation, to minimize the possibility of fuel damage and to reduce energy transfer from the core to the coolant.
Item 7 for Table 4.3.1.1-1 requires channel calibration at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This item becomes overdue on August 25,1998 and requires an extension of 21 days to reach September 14,1998. Thejustification for this extension is provided in Enclosure 3.
Item 11 Reactor Mode Switch In Shutdown As stated in UFSAR Section 7.2.1.1.5.4, when the mode switch is in Shutdown, the reactor is to be shutdown with all control rods inserted. This scram is not con-sidered a protective function because it is not required to protect the fuel or nuclear system process barrier, and it bears no relationship to minimizing the release of radioactive material from any barrier.
Item 11 for Table 4.3.1.1-1 requires channel functional test at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). The functional testing of this component is a portion of the system logic system functional test. This item becomes overdue on October 5,1998 and requires an extension of 14 days to the end of the refueling outage. The justification for this extension is provided in Enclosure 4.
RPS Logic System Functional Tests (LSIT) (TS Section 4.3.1.2)
SR 4.3.1.2 requires a LSFT and simulated automatic actuation of all channels of the RPS at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). The Logic System Functional Test consists of perfonning several different plant procedures which, when combined, test the complete logic system.
The first of these plant procedures becomes overdue on August 24,1998. The longest extension required for any plant procedure supporting the Logic System Functional Test is 22 days. Since some components of the RPS system are required to be functional
Attachment I to I
NRC-98-0040 Page 14 during Operational Conditions 4 and 5 this extension is requested to the end of the refueling outage. Thejustification for this extension is provided in Enclosure 4.
Primary and Secondary Isolation (TS 4.3.2.1)
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The Containment and Reactor Vessel Isolation Control System (CRVICS) includes the l
instrument 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, instrument channel relay contacts opx and deenergize the trip logic and thereby initiate isolation. Once initiated, the CRVICS trip logics seal in and may be reset by the operator only when the initiating conditions return to normal. The following variables provide input to the CRVICS logics for initiation of reactor vessel and containment isolation, as well as initiation trip for other plant functions when predetermined limits are exceeded:
As stated in UFSAR Section 7.3.2.2.7.1, a low-water level in the RPV could indicate that reactor coolant is being lost through a breach in the nuclear system process barrier and that the core has the potential of becoming overheated as the reactor coolant inventory diminishes. Reactor vessel low-water level initiates closure of various Class A valves and Class B valves. The closure of Class A valves is intended to either isolate a breach in any of the lines in which valves are closed, or to conserve reactor coolant by closing off process lines. The closure of Class B valves is intended to prevent the escape of radioactive materials from the primary containment through process lines that are in communication with the primary containment free space. Three RPV low water level isolation trip settings are used to completely isolate the RPV and the primary containment. The level signals are defined as follows:
a.
Level 3 is the highest of the three and also initiates the level scram and isolates RHR shutdown cooling.
b.
Level 2 is the initiation level for the reactor core isolation cooling (RCIC) and HPCI systems and is selected to be less than the volume resulting from a l
void collapse occurring in the event of a scram from full power. Level 2 l
also closes certain containment isolation valves.
c.
Level 1 is selected far enough above the top of the active fuel based on the time required for the RHR'and Core Spray Systems to function in the event of a large break. Level 1 also isolates the MSIVs.
~ Attachment 1 to NRC-98-0040 Page 15.
The following discussions are provided for Isolation Actuation Instrumentation (TS Table 4.3.2.1-1) surveillance interval extensions.
Item 1.a.1 Reactor Vessel Low Water Level-Level 3 Item 1.a.1 for Table 4.3.2.1-1 requires channel calibration at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2).. This item becomes overdue on August 24,1998 and requires an extension of 22 days to reach September 14,1998. Thejustification for this extension is provided in Enclosure 5.
Item 1.a.2 Reactor Vessel Low Water Level-Level 2 Item 1.a.2 for Table 4.3.2.1-1 requires channel calibration at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months i
per TS 4.0.2). This item becomes overdue on August 23,1998 and requires an extension of 23 days to reach September 14,1998. Thejustification for this j
extension is provided in Enclosure 5.
I Item 1.a.3 Reactor Vessel Low Water Level-Level 1 Item 1.a.3 for Table 4.3.2.1-1 requires channel calibration at least once per 18 l
months (with a maximum allowable surveillance interval extension of 4.5 months l
per TS 4.0.2). This item becomes overdue on August 23,1998 and requires an j
extension of 23 days to reach September 14,1998. Thejustification for this j
' extension'is provided in Enclosure 5.
- Item 1.b Drywell High Pressure
' As stated in UFSAR Section 7.3.2.2.7.6, high pressure in the drywell could indicate a breach of the nuclear system process barrier inside the drywell. The i
automatic closure of various Class B valves prevents the release of significant amounts of radioactive material from the primary containment. Item 1.b for Table 4.3.2.1-1 requires channel calibration at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2).
This item becomes overdue on August 25,1998 and requires an extension of 21 days to reach September 14,1998. The justification for this extension is provided in Enclosure 5.
.. to NRC-98-0040 Page 16 Item 1.c.1 Main Steam Line High Radiation As stated in UFSAR Section 7.3.2.2.7.2, high radiation in the vicinity of the main steam lines could indicate a gross release of fission products from the fuel. High radiation near the main steam lines initiates isolation of all main steam lines, main steam line drain and reactor water sample line. Item 1.c.1 for Table 4.3.2.1-1 requires channel calibration at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This item becomes overdue on August 25,1998 and requires an extension of 21 days to reach September 14,1998. The justification for this extension is provided in l.
l Item 1.c.2 Main Steam Line Pressure Low Low MSL pressure indicates that there may be a problem with the turbine pressure regulation, which could result in a low reactor vessel water level condition and the RPV cooling down more than 100 F/hr if the pressure loss is allowed to continue. Item 1.c.2 for Table 4.3.2.1-1 requires channel calibration at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This item becomes overdue on August 25, 1998 and requires an extension of 21 days to reach September 14,1998. The justification for this extension is provided in Enclosure 5.
Item 1.c.3 Main Steam Line Flow High Main steam line high flow could indicate a break in a main steam line. The automatic closure of various Class A valves prevents the excessive loss of reactor coolant and the release of significant amounts of radioactive material from the nuclear system process barrier. On detection of the main steam line high flow, all four main steam lines and the main steam line drain are isolated. Item 1.c.3 for Table 4.3.2.1-1 requires a channel calibration at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2).
This item becomes overdue on August 25,1998 and requires an extension of 21 days to reach September 14,1998. Thejustification for this extension is provided in Enclosure 5.
Item 1.d Main Steam Line Tunnel Temperature - High High temperature in the space in which the main steam lines are located outside the primary containment could indicate a breach in a main steam line. The automatic closure of various Class A valves prevents both the excessive loss of reactor coolant and the release of significant amounts of radioactive material from the nuclear system process barrier. When high temperatures occur in the main l
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Attachment I to NRC-98-0040 Page 17 steam line space, all four main steam lines and the main steam line drain are isolated. Item 1.d for Table 4.3.2.1-1 requires channel calibration at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This item becomes overdue on August 26,1998 and requires an extension of 20 days to reach September 14,1998. Thejustification for this extension is provided in Enclosure 5.
Item 1.e Condenser Pressure High The Condenser Pressure High Function is provided to prevent over-pressurization of the main condenser in the event of a loss of the main condenser vacuum. As condenser pressure increases, an upper limit ensures that the MSIV closure occurs soon enough to prevent release of steam to the Turbine Build.ing through the turbine shaft seals or from rupture of the turbine exhaust diaphragms. Item 1.e for Table 4.3.2.1-1 requires channel calibration at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2).
This item becomes overdue on August 27,1998 and requires an extension of 19 days to reach September 14,1998. Thejustification for this extension is provided in Enclosure 5.
Item 1.f Turbine Building Area Temperature - High High temperature in the space in which the main steam lines are located outside the primary containment could indicate a breach in a main steam line. The automatic closure of various Class A valves prevents both the excessive loss of reactor coolant and the release of significant amounts of radioactive material from the nuclear system process barrier. When high temperatures occur in the main steam line space, all four main steam lines and the main steam line drain are isolated. Item 1.f o Table 4.3.2.1-1 requires channel calibration at least once per r
18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This item becomes overdue on August 27,1998 and requires an extension of 19 days to reach September 14,1998. Thejustification for this extension is provided in Enclosure 5.
Item 1.h ManualInitiation The manual initiation push button channels introduce signals into the isolation logic that are redundant to the automatic protective instrumentation and provide manual isolation capability. There is no specific UFSAR safety analysis that takes credit for this function. It is retained for the overall redundancy and diversity of the isolation function as required by the NRC in the plant licensing basis. Item 1.h for Table 4.3.2.1-1 requires a channel functional test at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5
Attachment I to NRC-98-0040 Page 18 months per TS 4.0.2). This item becomes overdue on August 28,1998 and requires an extension of 18 days to reach September 14,1998. Thejustification for this extension is provided in Enclosure 6.
Item 2.d Reactor water Cleanup System (RWCU) Isolation - Standby Liquid Control System (SLCS) Initiation.
The isolation of the RWCU System is required when the SLCS System has been initiated to prevent dilution and removal of the boron solution by the RWCU System. SLCS System initiation signals are initiated from the two SLCS pump start signals. Item 2.d for Table 4.3.2.1-1 requires a channel functional test at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This item becomes overdue on August 29, 1998 and requires an extension of17 days to reach September 14,1998. The justification for this extension is provided in Enclosure 6.
Item 2.e Reactor Vessel Low Water Level - Level 2 See the Reactor Vessel Low Water Level discussion provided under Primary Containment above. Item 2.e for Table 4.3.2.1-1 requires channel calibration at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This item becomes overdue on August 23, 1998 and requires an extension of 23 days to reach September 14,1998. The justification for this extension is provided in Enclosure 5.
Item 5.a Reactor Vessel Low Water Level - Level 3 See the Reactor Vessel Low Water Level discussion provided under Primary Containment above. Item 5.a for Table 4.3.2.1-1 requires channel calibration at least once per 18 months (with a maximum allowable surveillance interval extension of4.5 months per TS 4.0.2). This item becomes overdue on August 24, 1998 and requires an extension of 22 days to reach September 14,1998. The justification for this extension is provided in Enclosure 5.
Item 5.c RHR Shutdown Cooling Mode Isolation Manual Initiation The manual initiation push button channels introduce signals into the isolation logic that are redundant to the automatic protective instrumentation and provide manual isolation capability. There is no specific UFSAR safety analysis that takes credit for this function. It is retained for the overall redundancy and diversity of the isolation function as required by the NRC in the plant licensing basis. Item 5.c for Table 4.3.2.1-1 requires channel functional test at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5
Attachment I to NRC-98-0040 Page 19 months per TS 4.0.2). This item becomes overdue on August 31,1998 and requires an extension of 15 days to reach September 14,1998. Thejustification for this extension is provided in Enclosure 6.
Item 6.a Reactor Vessel Low Water Level-Level 2 See the Reactor Vessel Low Water Level discussion provided under Primary Containment above. Item 6.a for Table 4.3.2.1-1 requires channel calibration at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This item becomes overdue on August 23, 1998 and requires an extension of 57 days to the end of the refueling outage. The justification for this extension is provided in Enclosure 5.
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Item 6.b Drywell Pressure High l
See the Drywell Pressure High discussion provided under Primary Containment above. Item 6.b for Table 4.3.2.1-1 requires channel calibration at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This item becomes overdue on August 25,1998 and requires an extension of 21 days to reach September 14,1998. Thejustification for this extension is provided in Enclosure 5.
Isolation Logic System Functional Tests (LSFT)(TS Section 4.3.2.2)
SR 4.3.3.2 requires a LSFT and simulated automatic actuation of all channels of the Primary and Secondary Isolation system at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). The Logic System l
Functional Test consists of performing several different plant procedures, which when combined test the complete logic system. The first of these plant procedures becomes -
overdue on August 23,1998. The longest extension required for any plant procedure supporting the Logic System Functional Test is 57 days. Since some components of the primary and secondary containment isolation system are required to be functional during Operational Conditions 4 and 5, this extension is requested to the end of the refueling outage. The justification for this extension is provided in Enclosure 6.
Emergency Core Cooling System (TS Section 4.3.3.1, Table 4.3.3.1-1)
As stated in UFSAR Section 6.3.1 the objective of the emergency core cooling systems (ECCS), in conjunction with the containment, is to limit the release of radioactive materials should a LOCA occur, so that resulting radiation exposures are kept within the limits prescribed in 10 CFR 100. The purp se of the ECCS instrumentation is to initiate appropriate responses from the systems '.o ensure that the fuel is adequately cooled in the event of a design basis accident or 1.asient. For most anticipated operational l
I
s to NRC-98-0040 Page 20 3
l occurrences and Design Basis Accidents (DBAs), a wide range of dependent and I
independent parameters are monitored. The following variables provide input to the l
ECCS logics for initiation of ECCS and ECCS support systems.
I Item 1.a Reactor Vessel Water Level Low - Level 1 i
Previously discussed under Primary and Secondary Containment Instmmentation.
Item 1.a for Table 4.3.3.1-1 requires channel calibration at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This item becomes overdue on August 26,1998 and requires an extension of 54 days to the end of the refueling outage. Thejustification for this extension is provided in Enclosure 7.
l Item 1.b Drywell Pressure High
)
Previously discussed under Primary and Secondary Containment Instrumentation.
Item 1.b for Table 4.3.3.1-1 requires channel calibration at least once per 18 l
months (with a maximum allowable surveillance interval extension of 4.5 months i
per TS 4.0.2). This item becomes overdue on September 5,1998 and requires an
{
extension of 10 days to reach September 14,1998. Thejustification for this extension is provided in Enclosure 7.
Item 1.c Steam Dome Pressure Low Reactor Steam Dome low pressure provides the permissive signal for ECCS injection. This signal prevents opening the injection valves until pressure is within the limitations of the injection system. This prevents the potential ovegressurization of ECCS low pressure piping. Item 1.c for Table 4.3.3.1-1 requires channel calibration at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This item becomes overdue on August 24,1998 and requires an extension of 56 days to the end of the refueling outage. Thejustification for this extension is provided in.
Item 1.d Core Spray ManualInitiation The manual initiation push button channels introduce signals into the ECCS start logic that are redundant to the automatic protective instnunentation and provide manual ECCS start capability. Item 1.d for Table 4.3.3.1-1 requires channel functional tests at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This item becomes overdue on August 28,1998 and requires an extension of 52 days to the end of the refueling outage. The justification for this extension is provided in Enclosure 8.
Attachment I to NRC-98-0040 Page 21 Item 2.a Reactor Vessel Water Level Low - Level 1 Previously discussed under Primary and Secondary Containment Isolation Instrumentation. Item 2.a for Table 4.3.3.1-1 requires channel calibration at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This item becomes overdue on August 26,1998 and requires an extension of 54 days to the end of the refueling outage. The justification for this extension is provided in Enclosure 7.
Item 2.b Drywell Pressure High Previously discussed under Primary and Secondary Containment Isolation Instrumentation. Item 2.b for Table 4.3.3.1-1 requires channel calibration at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This item becomes overdue on September 5,1998 and requires an extension of 10 days to reach September 14,1998. Thejustification for this extension is provided in Enclosure 7.
Item 2.c Steam Dome Pressure Low Previously discussed under Item 1.c above Item 2.c for Table 4.3.3.1-1 requires channel calibration at least once per 18 months (with a maximum allowaMe surveillance interval extension of 4.5 months per TS 4.0.2). This item b=ames overdue on August 24,1998 and requires an extension of 56 days to the end of the refueling outage. Thejustification for this extension is provided in Enclosure 7.
Item 2.d Reactor Vessel Water Level Low - Level 2 Previously discussed under Primary and Secondary Containment Isolation Instrumentation. Item 2.d for Table 4.3.3.1-1 requires channel calibration at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This item becomes overdue on August 26,1998 and requires an extension of 54 days to the end of the refueling outage. The justification for this extension is provided in Enclosure 7.
Item 2.c Steam Dome Pressure Low Steam Dome Pressure low provides a permissive signal for LPCI loop select logic. The signal delays LPCI loop select logic during a small break loss of coolant accident until reactor pressure decreases to increase sensitivity of the loop selection process.
s Attachment I to NRC-98-0040 Page 22 Item 2.e for Table 4.3.3.1-1 requires channel calibration at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This item becomes overdue on August 23,1998 and requires an extension of 57 days to the end of the refueling outage. Thejustification for this extension is provided in Enclosure 7.
Item 2.h LPCI ManualInitiation The manual initiation push button channels introduce signals into the ECCS start logic that are redundant to the automatic protective instrumentation and provide manual ECCS start capability. Item 2.h for Table 4.3.3.1-1 requires channel functional test at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This item becomes overdue on August 31,1998 and requires an extension of 49 days to the end of the refueling outage. Thejustification for this extension is provided in Enclosure 8.
Item 3.a Reactor Vessel Water Level Low - Level 2 Previously discussed under Primary and Secondary Containment Isolation Instrumentation. Item 3.a for Table 4.3.3.1-1 requires channel calibration at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This item becomes overdue on August 26,1998 and requires an extension of 20 days to reach September 14,1998. Thejustification for this extension is provided in Enclosure 7.
l Item 3.b Drywell Pressure High I
Previously discussed under Primary and Secondary Containment Isolation 1
Instrumentation. Item 3.b for Table 4.3.3.1-1 requires channel calibration at least j
once per 18 months (with a maximum allowable surveillance interval extension of i
4.5 months per TS 4.0.2). This item becomes overdue on September 5,1998 and requires an extension of 10 days to reach September 14,1998. Thejustification l
for this extension is provided in Enclosure 7.
l Item 4.a Reactor Vessel Water Level Low - Level 1 r
Previously discussed under Primary and Secondary Containment Instrumentation.
Item 4.a for Table 4.3.3.1-1 requires channel calibration at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This item becomes overdue on August 26,1998 and requires an extension of 20 days to reach September 14,1998. Thejustification for this extension is provided in Enclosure 7.
Attachment I to NRC-98-0040 Page 23 Item 4.f Reactor Vessel Water Level Low - Level 3 Previously discussed under Primary and Secondary Containment Instmmentation.
Item 4.f for Table 4.3.3.1-1 requires channel calibration at least once per 18 l
months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This item becomes overdue on September 3,1998 and requires an extension of 12 days to reach September 14,1998. Thejustification for this extension is provided in Enclosure 7.
Item 4 h Drywell Pressure - High Bypass Timer One of the signals required for ADS initiation is Drywell Pressure-High.
However, if the event requiring ADS initiation occurs outside the drywell (e.g.,
main steam line break outside containment), a high drywell pressure signal may never be present. Therefore, the Automatic Depressurization System Low Water Level Actuation Timer is used to bypass the Drywell Pressure-High Function after a certain time period has elapsed. Item 4.h for Table 4.3.3.1-1 requires channel calibration at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This item becomes overdue on August 26,1998 and requires an extension of 20 days to the reach September 14,1998. The justification for this extension is provided in Enclosure 7.
l ECCS System Logic System Functional Tests (TS Section 4.3.3.2) l SR 4.3.3.2 requires a LSFT and simulated automatic actuation of all channels of the ECCS system at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). The Logic System Functional Test consists of performing several different plant procedures which, when combined, test the complete logic system. The first of these plant procedures becomes overdue on August 23,1998. The longest extension required for any plant procedure supporting the Logic System Functional Test is 57 days. Since some components of the ECCS system ree required to be functional during Operational Conditions 4 and 5 this extension is 1
requested to the end of the refueling outage. Thejustification for this extension is j
l provided in Enclosure 8.
i
. ECCS System Response Time Testing (TS Section 4.3.3.3) l Technical Specification SR 4.3.3.3 requires the ECCS system response time testing of the functional units to be within the limits 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 least one channel per trip system ch that all channels are tested at least once every N times 18 months where N is 11: total number
.. to NRC-98-0040 Page 24 of redundant channels in a specific trip system. This item becomes overdue on August 28,1998 and requires an extension of 52 days to the end of the refueling outage. The justification for this extension is provided in Enclosure 9.
ATWS Recirculation Pump Trip Instrumentation (T3 Section 4.3.4)
The ATWS-RPT System initiates a recirculation pump trip (RPT), adding negative reactivity following events in which a scram does not (but should) occur, to lessen the effects of an ATWS event. Tripping the recirculation pumps adds negative reactivity from the increase in steam voiding in the core area from core flow decrease. When Reactor Vessel Water Level low or Reactor Steam Dome Pressure high setpoint is l
reached, the recirculation pump drive motor breakers and motor-generator field breakers trip. Both of these initiating parameters have been discussed for the Primary and Secondary Containment Isolation or Reactor Protection System section above.
Item 1 for Table 4.3.4-1 requires channel calibration at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This item becomes overdue on August 26,1998 and requires an extension of 20 days to reach l
September 14,1998. The justification for this extension is provided in Enclosure 10.
Item 2 for Table 4.3.4-1 requires channel calibration at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This item becomes overdue on August 24,1998 and requires an extension of 22 days to reach September 14,1998. The justification for this extension is provided in Enclosure 10.
' ATWS Recirculation Pump Trip System Logic System Functional Test (TS Section 4.3.4.2) l Technical Specification SR 4.3.4.2 requires a LSFT and simulated automatic actuation of l
all channels of the ATWS Recirculation Pump Trip System at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2).
The Logic System Functional Test consists of performing several different plant procedures, which when combined test the complete logic system. The first of these plant procedures becomes overdue on August 24,1998 and requires an extension of 22 days to reach September 14,1998. The justification for this extension is provided in Enclosure 11.
Reactor Core Isolation Cooling System (TS Section 4.3.5)
As stated in UFSAR Section 7.4.1.1, the reactor core isolation cooling (RCIC) system provides core cooling during reactor shutdown by pumping makeup water into the RPV i
in case of a loss of main feedwater flow. The RCIC system is started either automatically upon receipt of a low reactor water level signal (Level 2) or manually by the operator.
l 1
6 Attachment I to NRC-98-0040 Page 25 Water is pumped to the core by a turbine pump driven by reactor steam. High water level I
(Level 8) in the RPV indicates that the RCIC system has performed satisfactorily in providing makeup water to the RPV. All of these parameters have been discussed for the Primary and Secondary Containment Isolation or ECCS sections above.
Item a. for Table 4.3.5.1-1 requires channel c ilibration at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This L
item becomes overdue on August 26,1998 and requires an extension of 20 days to reach September 14,1998. The justification for this extension is provided in Enclosure 12.
Item b. for Table 4.3.5.1-1 requires channel calibration at least once per 18 months (with l
a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This item becomes overdue on August 26,1998 and requires an extension of 20 days to reach September 14,1998. Thejustification for this extension is provided in Enclosure 12.
l Reactor Core Isolation Cooling Logic System Functional Test (TS Section 4.3.5.2)
Technical Specification SR 4.3.5.2 requires a LSFT and simulated automatic actuation of all channels of the Reactor Core Isolation Cooling System at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2).
The Logic System Functional Test consists ofperforming several different plant procedures which, when combined, test the complete logic system. The first of these plant procedures becomes overdue on August 26,1998 and requires an extension of 20 days to reach September 14,1998. The justification for this extension is provided in 3.
Control Rod Block Instrumentation (TS Section 4.3.6)
The scram discharge high water level trip bypass is controlled by the manual operation of l
two keylocked switches, a bypass switch, and the mode switch. The mode switch must be in either the Shutdown or the Refuel position in order to bypass this trip. Four bypass channels emanate from the four banks of the RPS mode switch and are each connected into the RPS logic. This bypass allows the operator to reset the RPS scram relays so that l
the system is restored to operation while the operator drains the scram discharge volume.
l In addition, actuating the bypass initiates a control rod block. Resetting the trip actuators opens the scram discharge volume vent and drain valves. An annunciator in the main control room indicates the bypass condition. Operation of both the Mode Switch and the l
~ Scram Discharge Volume were previously discussed in this submittal.
I Item 5.b for Table 4.3.6-1 requires a channel ftmetional test at least once per 18 months (with a maximum allowable surveillance inters al extension of 4.5 months per TS 4.0.2).
This item becomes overdue on October 5,1998 and requires an extension of 14 days to i
I
_ _ ___ - _ -__ ___ _O
Attachment I to NRC-98-0040 Page 26 the end of the refueling outage. The justification for this extension is provided in
'r 4.
Item 7 for Table 4.3.6-1 requires a channel functional test at least once per 18 months
. (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2).
This item becomes overdue on October 5,1998 and requires an extension of 14 days to the end of the refueling outage. The justification for this extension is provided in 4.
Post Accident Monitoring (TS Section 4.3.7.5)
As discussed in UFSAR Section 7.5.1.4 information readouts are provided to accommodate events up to and including a LOCA. These readouts are designed from the standpoint of operator action, information, and event tracking requirements, providing
- assurance that requirements for all other credible events or incidents will be covered. The i
L process instrumentation proddes information to the operator for his use in monitoring reactor conditions after a LOCA.
l I'
Technical Specification SR 4.3.7.5, Table 4.3.7.5-1, Items 1,2.a,2.b,12, and 16 require a channel calibration of the instrumentation at least once per 18 months (with an allowable surveillance interval extension of 4.5 months per TS 4.0.2).
I
. Item I becomes overdue on August 24,1998 and requires an extension of 22 days to reach September 14,1998. Thejustification for this surveillance interval extension is
.provided in Enclosure 15.
Item 2.a becomes overdue on September 3,1998 and requires an extension of 12 days to reach September 14,1998. The justification for this extension is provided in Enclosure 15.
l Item 2.b becomes overdue on August 24,1998 and requires an extension of 22 days to l-reach September 14,1998. The justification for this extension is provided in Enclosure L
15.
Item 12 becomes overdue on September 4,1998 and requires an extension of 11 days to reach September 14,1998. Thejustification for this extension is provided in Enclosure
-15.
Item 16, the primary containment isolation valve position verification, consists of performing several different plant procedures which, when combined, verify all containment isolation valves. The first of these plant procedures becomes overdue on August 28,1998 and requires an extension of18 days to reach September 14,1998. The justification for this extension is provided in Enclosure 15.
I t-L 1
Gb Attachment I to l
Page 27 Feedwater Turbine Trip System (TS Section 4.3.9.1) i As stated in Technical Specification Bases Section 4.3.9, the feedwater/ main turbine trip l
system actuation instrumentation is provided to initiate action of the feedwater system / main turbine trip system in the event of a high reactor vessel water level due to failure of the feedwater controller under maximum demand.
Technical Specification Section 4.3.9.1 requires each feedwater/ main turbine trip system actuation instrumentation channel be demonstrated operable by the performance of Channel Calibrations for the operational cenditions and at the frequencies shown in Table 4.3.9.1-1. Table 4.3.9.1-1 requires these tests at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2) for Reactor Vessel High Water Level-Level 8. The Item a becomes overdue on August 31, 1998 and requires an extension of 15 days to reach September 14,1998. Thejustification for this extension is provided in Enclosure 16.
Feedwater Turbine Trip System Logic System Functional Test (TS Section 4.3.9.2)
Technical Specification Section 4.3.9.2 requires a Logic System Functional Test and simulated automatic operation of all channels at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This item becomes overdue on August 31,1998 and requires an extension of 15 days to reach September 14,1998. The justification for this extension is provided in Enclosure 17.
Alternative Shutdown System (TS Section 4.3.11)
As stated in Technical Specification Bases Section 4.3.11, the operability of the alternative shutdown system ensures that a fire will not preclude achieving safe i
shutdown. The alternative shutdown system instrumentation is independent of areas where a fire could damage systems normally used to shutdown the reactor. Thus, the system capability is consistent with General Design Criterion 3 and Appendix R to 10 CFR 50. Technical Specification Section 4.3.11.1 requires each alternative shutdown
)
instrumentation channel be demonstrated operable by the performance of Channel l
Calibrations for the operational conditions and at the frequencies shown in Table 4.3.11.1-1. Table 4.3.11.1-1 requires these tests at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2).
Items 7 (Reactor Water Level) and 8 (Reactor Pressure) will become overdue on September 1,1998 and require an extension of 14 days to reach September 14,1998. The justification for these extensions is provided in Enclosure 18.
l l
l
Attachment I to NRC-98-0040 Page 28 Safety Relief Valves (TS Section 4.4.2)
The safety valve ftmetion of the safety relief valves operate to prevent the reactor coolant system from being pressurized above the Safety Limit of 1325 psig in accordance with the ASME Code. A total of 11 operable safety relief valves is required to limit reactor pressure to within ASME III allowable values for the worst case upset transient.
SR 4.4.2.2.b requires the low-low set function pressure actuation instrumentation be demonstrated operable by performance of a Channel Calibration, Logic System Functional Test and simulated automatic operation of the entire system at least once per 18 months. This item becomes overdue on August 24,1998 and requires an extension of 22 days to reach September 14,1998. Thejustification for this extension is provided in 9.
Emergency Core Cooling System Operation (TS Section 4.5.1)
As stated in Technical Specification Bases Section 3/4.5.1 and 3/4.5.2, the core spray system (CSS), together with the LPCI mode of the RHR system, is provided to assure that the core is adequately cooled following a loss-of-coolant accident and provides adequate core cooling capacity for all break sizes up to and including the double-ended reactor recirculation line break, and for smaller breaks following depressurization by the ADS. The CSS is a primary source of emergency core cooling after the reactor vessel is depressurized and a source for flooding of the core in case of accidental draining. The surveillance requirements provide adequate assurance that the CSS will be operable when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown.
SR 4.5.1.c.1 requires that the emergency core cooling systems 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) for the CSS, the LPCI system, and the HPCI system by performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence and, verifying that each automatic valve in the flow path actuates to its correct position. Actual injection of coolant into the reactor vessel may be excluded from this test. SR 4.5.1.c.1 for CSS, LPCI and HPCI becomes overdue on August 24,1998 and requires an extension of 56 days to the end of the refueling outage. Thejustification for this extension is provided in Enclosure 21.
SR 4.5.1.d.2.a requires that the emergency core cooling systems be demonstrated operable at least once per 18 months (with a maximum allowable surveillance iturval extension of 4.5 months per TS 4.0.2) for the ADS by performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence, but excluding actual valve actuation. SR 4.5.1.d.2.a for ADS becomes overdue
j
. to NRC-98-0040 Page 29 on August 26,1998 and requires an extension of 20 days to reach September 14,1998.
Thejustification for this extension is provided in Enclosure 21.
Primary Containment Isolation Valves (TS Section 4.6.3)
As stated in Technical Specification Bases Section 3/4.6.3, the operability of the primary containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment and is consistent with the requirements of GDC 54 through 57 of Appendix A of 10 CFR Part 50. Containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that the release of radioactive material to the environment will be consistent with the assumptic as used in the analyses for a LOCA.
SR 4.6.3.2 requires each primary containment automatic isolation valve be demonstrated operable during cold shutdown or refueling 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 a containment isolation test signal each automatic isolation valve actuates to its isolation position. The isolation valve operability verification consists ofperforming several different plant procedures which, when combined, verify operability of all isolation valves. The first of these plant procedures becomes overdue on August 23,1998 and requires an extension of 23 days to reach September 14,1998. Thejustification for this extension is provided in Enclosure 22.
Secondary Containment (TS Section 4.6.5)
As stated in Technical Specification Bases Section 3/4.6.5, secondary containment is designed to minimize any ground level release of radioactive material which may result from an accident. The reactor building and associated structures provide secondary containment during normal operation when the drywell is sealed and in service. At other times, the drywell may be open and, when required, secondary containment integrity is specified.
SR 4.6.1.d.1, and d.2: SR 4.6.1.d.1 requires, at least once per 18 months, that one standby gas treatment subsystem to be verified to draw down the secondary containment to greater than or equal to 0.25 inch of vacuum water gauge in less than 567 seconds at a flow rate ofless than or equal to 3800 cfm. In addition SR 4.6.1.d.2 requires, at least once per 18 months, operation of one standby gas treatment subsystem for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and maintaining greater than or equal to 0.25 inch of vacuum water gauge in the secondary containment at a flow rate not exceeding 3000 cfm. These items become overdue on September 30,1998, and require an extension of 19 days to the end of the refueling outage. Thejustification for these extensions is provided in Enclosure 23.
l
)
i
, to NRC-98-0040 f
Page 30 '
I SR 4.6.5.2.b requires each secondary containment ventilation system automatic isolation damper shown in Technical Specification Table 3.6.5.2-1 be demonstrated operable during cold shutdown or refueling 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 a containment isolation test signal each isolation damper actuates to its isolation position.
This item becomes overdue on October 2,1998 and requires an extension of 17 days to l
the end of the refueling outage. Thejustification for this extension is provided in i
' 3.
Service Water Systems (TS Section 4.7.1)
As stated in Technical Specification Bases Section 3/4.7.1, the operability of the service water systems ensures that sufficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions.
L SR 4.7.1.2.b requires the emergency equipment cooling water system be demonstrated operable at least once per 18 months (with a maximum allowable smveillance interval extension of 4.5 months per TS 4.0.2) by verifying that each automatic valve servicing
)
non-safety-related equipment actuates to its isolation position and the associated EECW pump automatically starts on an automatic actuation test signal. This item becomes l
overdue on August 30,1998 and requires an extension of 50 days to the end of the l
refueling outage. Thejustification for this extension is provided in Enclosure 24.
l SR 4.7.1.3.b requires the emergency equipment service water system 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 the EESW pump automatically starts upon receipt of an actuation test signal. This item becomes overdue on August 30,1998 and requires an extension of 50 days to the end of the refueling outage. Thejustification l
for this extension is provided in Enclosure 24.
i SR 4.7.1.4.b requires each of the diesel generator service water subsystems 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 each DGSW pump starts automatically upon receipt of a start signal for the associated diesel generator.
This item becomes overdue on August 27,1998 and requires an extension of 53 days to l
the end of the refueling outage. Thejustification for this extension is provided in
' 4.
Control Room Emergency Filtration System (TS Section 4.7.2)
As stated in Technical Specification Bases Section 3/4.7.2, the operability of the control room emergency filtration system ensures that (1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and
l O to NRC-98-0040 i
t Page 31 instrumentation cooled by this system and (2) the control room will remain habitable for operations personnel during and following all design basis accident conditions.
Continuous operation of the system with heaters operable for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> during each 31-i day period is sufficient to reduce the buildup of moisture on the charcoal adsorbers. The operability of this system in conjunction with control room design provisions is based on l
limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent. This limitation is consistent with the requirements of General Design Criterion 19 of Appendix A,10 CFR Part 50.
l l
SR 4.7.2.1.c, Items 1,2,3 require that the control room emergency filtration system be l
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:
I 1.
Verifying that the system satisfies the in-place penetration testing acceptance criteria ofless than 1.0% and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c, and C.5.d of Regulatory Gaide 1.52, Revision 2, March 1978, while operating the system at a flow rate of I800 l
cfm 10% through the makeup filter and 3000 cfm 10% through the j
l recirculation filter.
l 2.
Verifying within 31 days aller removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory l
Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets l
the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 1.0%.
l 3.
Verifying a system flow rate of 3000 cfm
- 10% during system operation when tested in accordance with ANSI N510-1980.
These items become overdue on September 21,1998 and require an extension of 28 days to the end of the refueling outage. Thejustification for this extension is provided in 5.
SR 4.7.2.1.e Items 1,2 & 4 require that the control room emergency filtration system be demonstrated operable at least once per 18 months (with a maximum allowable j
surveillance interval extension of 4.5 months per TS 4.0.2) by:
l
- 1. Verifying that the pressure drop across the recirculation train and across the makeup train combined HEPA filters and charcoal adsorber banks is less than 8 inches and 6 inches water gauge respectively while operating the system at a flow rate of 3000 cfm
- 10% through the recirculation filter train and 1800 cfm
- 10% through the makeup filter train.
l l
l
Attachment I to NRC-98-0040 Page 32 This item (SR 4.7.2.1.e.1) becomes overdue on September 21,1998 and requires an extension of 28 days to the end of the refueling outage. The justification for this extension is provided in Enclosure 25.
2.
Verifying that the system will automatically switch to the recirculation mode of operation on each of the below actuation test signals and verifying that on any one of the below recirculation mode actuation test signals, the system automatically switches to the recirculation mode of operation, the isolation valves close within 5 seconds and the control room is maintained at a positive pressure of at least 0.125 inch water gauge relative to the outside atmosphere during system operation at a flow rate less than or equal to 1800 cfm through the emergency makeup air filter:
a) Control center inlet radiation monitor, b) Fuel pool ventilation exhaust radiation monitor.
c) Low reactor water level.
d) High drywellpressure.
This item (SR 4.7.2.1.e.2) becomes overdue on October 2,1998 and requires an extension of 17 days to the end of the refueling outage. Thejustification for this extension is provided in Enclosure 25.
4.
Verifying that each of the emergency makeup inlet air heaters dissipate 12.0
- 2.0 kW when tested in accordance with ANSI N510-1980. This item (SR 4.7.2.1.e.4) becomes overdue on September 21,1998 and requires an extension of 28 days to the end of the refueling outage. Thejustification for this extension is provided in Enclosure 25.
Reactor Core Isolation Cooling System (TS Section 4.7.4)
SR 4.7.4.c.1 requires a system functional test of the Reactor Core Isolation Cooling (RCIC) System to be performed. This function test includes simulated automatic actuation and restart and verification that each automatic valve in the flow path actuates to its correct position. The test is required to be performed to demonstrate RCIC System operability at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). This item becomes overdue on August 26,1998 and requires an extension of 20 days to reach September 14,1998. Thejustification for the extension of this item is provided in Enclosure 26.
5 Attachment I to NRC-98-0040 Page 33 Snubbers (TS Section 4.7.5) l l
As stated in Technical Specification Bases Section 3/4.7.5: All snubbers are required operable to ensure that the structural integrity of the reactor coolant system and all other safety-related systems is maintained during and following a seismic or other event initiating dynamic loads. Snubbers excluded from this inspection program are those installed on non-safety-related systems and then only if their failure or failure of the system on which they are installed would have no adverse effect on any safety-related system.
SR 4.7.5.e requires each snubber be demonstrated operable by performing functional tests. At least once per 18 months during shutdown, a representative sample of snubbers shall be tested using one of the provided sample plans. The sample plan shall be selected prior to the test period and cannot be changed during the test period. The NRC Regional Administrator shall be notified in writing of the sample plan selected prior to the test period or the sample plan used in the prior test period shall be implemented. Detroit Edison has elected to again use the 10% sample plan.
This item becomes overdue on October 4,1998 and requires an extension of 15 days to the end of the refueling outage. Thejustification for this extension is provided in 7.
A.C. Sources, D.C. Sources And Onsite Power Distribution Systems (TS Section 4.8.1)
As stated in Technical Specification Bases Sections 3/4.8.1,3/4.8.2 and 3/4.8.3 the operability of the A.C. and D.C. power sources and associated distribution systems during operation ensures that sufficient power will be available to supply the safety-related equipment required for (1) the safe shutdown of the facility and (2) the mitigation and control of accident conditions within the facility. The minimum specified independent and redundant A.C. and D.C. power sources and distribution systems satisfy the requirements of General Design Criteria 17 of Appendix A to 10 CFR 50.
The surveillance requirements for demonstrating the operability of the diesel generators are in accordance with the recommendations of Regulatory Guide 1.9, " Selection of Diesel Generator Set Capacity for Standby Power Supplies", December 1979; Regulatory Guide 1.108, " Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants", Revision 1, August 1977; and Regulatory Guide 1.137, " Fuel-Oil Systems for Standby Diesel Generators", Revision 1, October 1979.
5
. to NRC-98-0040 Page 34 Emergency Diesel Generators SR 4.8.1.1.2.e.1 requires each of the diesel generators 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 subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for this class of standby service. This item becomes overdue on August 19,1998 and requires an extension of 61 days to the end of the refueling outage. The justification for this extension is provided in Enclosure 28. This SR may be performed on-line prior to RFO6.
In that case this SR will not require an extension.
SR 4.8.1.1.2.e.2 requires each of the diesel generators 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 the diesel generator capability to reject a load of greater than or equal to 1666 kW while maintaining engine speed less than the nominal speed plus 75% of the difference between nominal speed and the overspeed trip setpoint or 115% of nominal speed, whichever is lower. This item becomes overdue on August 27,1998 and requires an extension of 53 days to the end of the refueling outage. The justification for this extension is provided in Enclosure 28. This SR may be performed on-line prior to RFO6. In that case this SR will not require an extension.
SR 4.8.1.1.2.e.3 requires each of the diesel generators 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 the diesel generator capability to reject a load of 2850 kW without tripping. The generator voltage shall not exceed 4784 volts during and following the load rejection. This item becomes overdue on August 27,1998 and requires an extension of 53 days to the end of the refueling outage. The justification for this extension is provided in Enclosure 28. This SR may be performed on-line prior to RFO6. In that case this SR will not require an extension.
SR 4.8.1.1.2.e.4.a requires each of the diesel generators 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 simulating a loss-of-offsite power by itself, and verifying deenergization of the emergency busses and load shedding from the emergency busses.
This item becomes overdue on August 30,1998 and requires an extension of 50 days to the end of the refueling outage. Thejustification for this extension is provided in 8.
SR 4.8.1.1.2.e.4.b requires each of the diesel generators be demonstrated operable at least l
once per 18 months (with a maximum allowable surveillance interval extension of 4.5 l
months per TS 4.0.2) by simulating a loss-of-offsite power by itself, and verifying the diesel generator starts on the auto-start signal, energizes the emergency busses with
)
i pennanently connected loads within 10 seconds, energizes the auto-connected loads
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Attachment I to I
Page 35 through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads. After energization, the steady-state voltage l
and frequency of the emergency busses shall be maintained at 4160
- 420 volts and 60
- l.2 Hz during this test. This item becomes overdue on September 29,1998 and requires an extension of 20 days to the end of the refueling outage. Thejustification for this extension is provided in Enclosure 28.
I SR 4.8.1.1.2.e.5 requires each of the diesel generators 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 ECCS actuation test signal, without loss-of-offsite power, the diesel generator starts on the auto-start signal and operates on standby for greater than or equal to 5 minutes. The generator voltage and frequency shall be 4160
- 420 volts and 60
- 1.2 Hz within 10 seconds after the auto-start signal; the steady-state generator voltage and frequency shall be maintained within these limits during this test.
This item becomes overdue on August 26,1998 and requires an extension of 54 days to the end of the refueling outage. Thejustification for this extension is provided in 8.
SR 4.8.1.1.2.e.6.a requires each of the diesel generators 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 simulating a loss-of-offsite power in conjunction with an ECCS actuation test signal, and verifying deenergization of the emergency busses and load shedding from the emergency busses. This item becomes overdue on August 30,1998 and requires an extension of 50 days to the end of the refueling outage. Thejustification for this extension is provided in Enclosure 28.
SR 4.8.1.1.2.e.6.b requires each of the diesel generators 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 simulating a loss-of-offsite power in conjunction with an ECCS actuation test signal, and verifying the diesel generator starts on the auto-start signal, energizes the emergency busses with permanently connected loads within 10 seconds, energizes the auto-connected shutdown loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads.
After energization, the steady-state voltage and frequency of the emergency busses shall be maintained at 4160
- 420 volts and 60
- 1.2 Hz during this test. This item becomes overdue on August 26,1998 and requires an extension of 54 days to the end of the refueling outage. Thejustification for this extension is provided in Enclosure 28.
SR 4.8.1.1.2.e.7 requires each of the diesel generators be demonstrated operable at least l
once per 18 months (with a maximum allowable surveillance interval extension of 4.5 l
months per TS 4.0.2) by verifying that all automatic diesel generator trips, except l
l overspeed, generator differential, low lube oil pressure, crankcase overpressure, and failure to start are automatically bypassed for an emergency start signal. This item
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Attachment I to -
l NRC-98-0040 l
Page 36 becomes overdue on August 27,1998 and requires an extension of 53 days to the end of the refueling outage. Thejustification for this extension is provided in Enclosure 28.
1 SR 4.8.1.1.2.e.8 requires each of the diesel generators be demonstrated operable at least i
once per 18 months (with a maximum allowable surveillance interval extension of 4.5 l
months per TS 4.0.2).by verifying the diesel generator operates for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
During the first 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> of this test, the diesel generator shall be loaded to greater than or equal to an indicated 2500-2600 kW and during the remaining 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of this test, the diesel generator shall be loaded to an indicated 2800-2900 kW. The generator voltage and frequency.shall be 4160 420 volts and 60
- 1.2 Hz within 10 seconds after the start signal; the steady-state generator voltage and frequency shall be maintained within these limits during this test. Within 5 minutes after completing this 24-hour test, the surveillance requires performance of Surveillance Requirement 4.8.1.1.2.a.4. This item (SR 4.8.1.1.2.e.8) becomes overdue on August 24,1998 and requires an extension of 56 days to the end of the refueling outage. Thejustification for this extension is provided in 8. This SR may be performed on-line prior to RFO6. In that case this SR will not require an extension.
i' SR 4.8.1.1.2.e.9 requires each of the diesel generators be demonstrated operable at ieast once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2) by verifying that the auto-connected loads to each diesel generator do not exceed the 2000-hour rating of 3100 kW. This item becomes overdue on October 14,1998 and requires an' extension of 5 days to the end of the refueling outage. The justification for this extension is provided in Enclosure 28.
SR 4.8.1.1.2.e.10 requires each of the diesel generators be demonstrated operable at least l
once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2) by verifying the diesel generator's capability to synchronize with the offsite power source while the generator is loaded with its emergency loads upon a l-simulated restoration of offsite power, transfer its loads to the offsite power source, and be restored to its standby status. This item becomes overdue on October 14,1998 and requires an extension of 5 days to the end of the refueling outage. Thejustification for this extension is provided in Enclosure 28.
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SR 4.8.1.1.2.e.11 requires each of the diesel generators be demonstrated operable at least i
once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2) by verifying that the automatic load sequence timer is operable with
- the interval between each load block within 10% ofits design interval. This item becomes overdue on September 26,1998 and requires an extension of 23 days to the end of the refueling outage. The justification for this extension is provided in Enclosure 28.
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, to NRC-98-0040 Page 37 SR 4.8.1.1.2.e.12 items a, b, and c require each of the diesel generators be demonstrated operable at least once per 18 months (with a maximum allowable surveillance inte val extension of 4.5 months per TS 4.0.2) by verifying that the following diesel generator lockout features prevent diesel generator starting only when required:
a) 4160-volt ESF bus lockout.
b)
Differential trip.
c)
Shutdown relay trip.
l These items become overdue on August 27,1998 and require an extension of 53 days to the end of the refueling outage. Thejustification for this extension is provided in 8.
Battery (TS Section 4.8.2) 4 The surveillance requirements for demonstrating the operability of the unit batteries are in accordance with the recommendations of Regulatory Guide 1.129 " Maintenance Testing i
and Replacement of Large Lead Storage Batteries for Nuclear Power Plants," February 1978, and IEEE Std 450-1972, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Generating Stations and Substations."
j Verifying average electrolyte temperature above the minimum for which the battery was sized, total battery terminal voltage on float charge, connection resistance values and the performance of battery service and discharge tests ensures the effectiveness of the i
charging system, the ability to handle high discharge rates and compares the battery capacity at that time with the rated capacity.
SR 4.8.2.1.c.3 requires that each of the required 130-volt batteries and chargers be i
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 the 4
L
. resistance of each cell-to-cell and terminal connection is less than or equal to 150 x 10 L
ohm. This item becomes overdue on August 25,1998 and requires an extension of 55 days to the end of the refueling outage. Thejustification for this extension is provided in j 9.
l SR 4.8.2.1.d requires that each of the required 130-volt batteries and chargers shall 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 either:
1.
The battery capacity is adequate to supply and maintain in operable status l
all of the actual emergency loads for the design duty cycle (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) when l
the battery is subjected to a battery service test, or
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Attachment I to NRC-9&0040 Page 38 2.
The battery capacity is adequate to supply a dummy load of the following profile while maintaining the battery terminal voltage greater than or equal l
to 105 or 210 volts, as applicable:
i a)
Batteries 2PA and 2PB greater than or equal to 710 amperes during l
the initial 6 seconds of the test.
b)
Batteries 2PA and 2PB greater than 182 amperes during the next 42 seconds of the test.
c)
Batteries 2PA and 2PB greater than or equal to 54 amperes during the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of the test.
d)
Batteries 2PA and 2PB greater than or equal to 480 amperes during the last 6 seconds of the test.
This item becomes overdue on August 25,1998 and requires an extension of 55 days to the end of the refueling outage. Thejustification for this extension is provided in 9.-
Electrical Equipment Protective Devices (TS Section 4.8.4)
' As stated in Technical Specification Bases Sections 3/4.8.4: Primary containment electrical penetrations and penetration conductors are protected by either de-energizing circuits not required during reactor operation or demonstrating the operability of primary and backup overcurrent protection circuit breakers by periodic surveillance.
SR 4.8.4.2.a.1.a requires each of the primary containment penetration conductor overcurrent protective devices shown in Technical Specification Table 3.8.4.2-1 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 performing a channel calibration of the associated 4.16-kV circuit protective relays. This item becomes overdue on August 26,1998 and requires an extension of 20 days to reach September 14, 1998. The justification for this extension is provided in Enclosure 30.
SR 4.8.4.2.a.1.b requires each of the primary containment penetration conductor overcurrent protective devices shown in Technical Specification Table 3.8.4.2-1 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) by an integrated system functional test which includes simulated automatic actuation of the system and verifying I
that each relay and associated circuit breakers and overcurrent control circuits function as designed. This item becomes overdue on August 26,1998 and requires an extension of 20 days to reach September 14,1998. The justification for this extension is provided in 0.
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, to NRC-98-0040 Page 39 SR 4.8.4.5.a requires that the Standby Liquid Control System isolation devices (circuit breakers) be demonstrated operable by performance of a channel calibration of the associated protective relays and a channel functional test of each breaker at least once per 18 months (with a maximum allowable extension of 4.5 months per TS 4.0.2). The channel functional test includes simulated actuation of the system and verification that each relay and associated circuit breaker and overcurrent circuit functions as designed.
This item becomes overdue on August 29,1998 and requires an extension of 51 days to reach the projected end of RFO6. Thejustification for the extension of this item is provided in Enclosure 30.
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Attachment I to L
NRC-98-0040 Page 40 -
ENCLOSURE 1 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION l
TECHNICAL SPECIFICATION SR 4.1.3.5 -
SCRAM ACCUMULATOR Justification for extending SR 4.1.3.5.b.2 l
f The scram accumulator check valves are only required to maintain pressure assuming that j
no control rod drive pump is operating. A review of check valve perfonnance for the last refueling outage has indicated no failures to maintain pressure. Therefore, the probability that the check valve would fail at a time when no control rod drive pump was available is considered low. In addition, the requested extension is for a SR which will become due l
when the unit will be in refueling. During refueling all rods are required to be inserted, L
except for a single rod which is permitted to be withdrawn as a special test' exception. A single rod can be withdrawn under the one-rod-out-interlock provided shutdown margin requirements have been satisfied. In that case, failure of the associated accumulator has no consequence. For multiple rod withdrawal, the associated cells must be defueled, eliminating the need for accumulator operability. Therefore, accumulator operability is not required for refueling activities.
l h
Based on the above discussion, including the historical failure review and evaluation, the requested extension isjustified.
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Page 41 l-ENCLOSURE 2 j
JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION '
TECHNICAL SPECIFICATION SR 4.1.5 STANDBY LIQUID CONTROL SYSTEM OPERABILITY Justification for extending Items d.1, d.2, and d.3 An evaluation of the system functional testing performed during the last refueling outage l
indicated no failures to meet acceptance criteria. In addition, explosive charge continuity and valve position' verification are performed monthly and pump testing is performed i
quarterly in accordance with the plant surveillance program. The explosive valves i
' (squibs) are purchased in lots with samples tested prior to installation. The triggers for both explosive valves have sufficient service life for the requested extension.
This SR tests the relief valves that provide system overpressure protection fmm the discharge of the positive displacement pumps. The SLCS is designed with two redundant loops. If one relief valve lifted at too low a pressure, the check valve in that discharge
. line would prevent the other pump's flow from recycling back to the storage tank. In p
addition, the current TS surveillance frequency significantly exceeds the ASME XI/OM-1 l
' requirements. The OM-1 requirement is that all valves of a type be tested within 10 years l
with a minimum of 20% tested within any 48 months. The Fermi 2 IST program requires l2 a 5 year period on these valves. The SR delineates the testing requirements to ensure L
operability of the SLCS.
L The SLCS.is a backup to the control rods. The system is also designed with a redundant loop. In addition, functional testing of the SLCS pump is performed on a quarterly basis throughout the operating cycle and the charges in the explosive valves are monitored for L
circuit continuity in the control room. An alann sounds when the circuit is opened.
I' During the functional testing, system pressure is raised to 1215 psig. Since the SLCS l
relief valves are exposed to this pressure, any significant relief valve setpoint drift would be detected during the performance of this more frequent test performed during the operating cycle. The overall impact on system availability, if any, of extending the operating cycle on this one-time basis is small. This conclusion is based on the fact that more frequent testing is performed, there are control room alarms which verify circuit continuity in the explosive valves, and active component redundancy.
~ A review of the history of the SR results indicates that previous testing of the relief valves was performed in situ with a wide tolerance because of the difficulty in determining the lifting pressure with a positive displacement pump. The function of the SLCS important to reactor' safety is to inject sodium pentaborate at high reactor pressure.
Failure of the relief valve on high has no impact on the operability of the SLCS to
. perform its safety function. Further review of the calibration history indicates that the valves have not required adjustment since initial plant startup prior to this bench test.
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, to NRC-98-0040 Page 42 This result demonstrates that the relief valves can be installed for a significant time period without having an adverse effect on the SLCS operability. No additional failures were identified by the historical review.
The temperature of the sodium pentaborate solution and the temperature of the SLCS pmnps suction piping is verified once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The minimum tank and piping temperature requirement for operation is 48 degrees Therefore, it is highly unlikely that the piping would become blocked.
The proposed one-time Technical Specification SR extension has little or no effect on the Standby Liquid Control System availability since the routine surveillance which include pump testing, temperature monitoring, and explosive valve continuity verification provide assurance of system operability. Therefore, the requested extension isjustified.
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1 e to NRC-98-0040 Page 43 ENCLOSURE 3 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.1.1 Table 4.3.1.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION Justification for extending Item 3,4 and 7 j
The Reactor Vessel Dome Pressure, Reactor Vessel Level - Level 3 and Drywell Pressure setpoints were analyzed using the GE Instrument Setpoint Methodology, NEDC-31336.
The specific calculation for the Reactor Vessel Dome Pressure instrument :hannels was submitted to the NRC during the power uprate review that was accepted by NRC Safety Evaluation Report dated September 9,1992 on Fermi-2 docket. The GE hstrument Setpoint Methodology, NEDC-31336, has been accepted by NRC Safety Evaluation Report dated July 18,1995 on the Perry Nuclear Power Plant docket. 'n addition, Rosemount published a report in February 1990, "30 Month Stability Specification For Rosemount 1152,1153,1154 Pressure Transmitters"(Rosemount Report D89000126, Revision A) which was accepted by NRC Safety Evaluation Report daat August 2,1993 on Peach Bottom Atomic Power Station docket. This report supported the extension of l
the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drin allowance.
The Fenni 2 design calculations contain a drift allowance for Rosemount 1153 transmitter equal to at least three consecutive six month vendor drin intervals. The previously published model 1153 six month drin interval value is of the same order of magnitude as the currently published thirty month drin interval value. Therefore, the calculations contain at least 30 months drift allowance. Associated loop trip units are checked on a quarterly basis. This check confirms that the trip units are maintained within allowances. As a result, the existing Fermi-2 design calculation includes an adequate allowance for drin and the extension has no potential impact on the plant safety analyses,i.e., no analytic limit changes are required. Plant calibration data was also reviewed and found to be within the design allowances. Therefore, the extension of the surveillance interval is justified.
Justification for extending Item 6 For the Main Steam Line (MSL) Radiation - High instruments, correct operation is confirmed by a quarterly channel functional check which does not include the detector, and a channel check every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. These tests will identify any potential drin of these instruments (except for the detector and cable) during this period of..me. The quarterly and twice daily surveillance tests, monitoring by plant staff, and the fact that the ion chamber detectors and cable are not considered susceptible to drin ensure that there is no potential impact on the subject instrument availability by allowing the additional time interval of source calibration of the existing radiation monitors.
I 1
.. to NRC-98-0040 Page 44 '
The surveillance allowance for the MSL Radiation monitoring instrument is 30 percent of span. A review of the surveillance tests indicates that drift for the MSL Radiation monitors was 19 percent for the worst case condition. A further review of the MSL radiation monitors calibration history indicates that the detectors have never been adjusted to c.ompensate for drift. Therefore, the drift history for these components documents the ability.of the MSL monitors to perform for multiple cycles without adjustment and with performance remaining within expected bounds.
- A historical search of the 18 month surveillance tests for the MSL radiation monitors for the last refueling outage was performed. The search criteria was to identify failed or partially failed tests, each failed or partially failed test was to be reviewed and evaluated.
The purpose of this evaluation was to demonstrate that the increased calibration surveillance interval would not increase the period an instrument would be unavailable.
One failure was identified in this review. However the detailed evaluation determined
- that this failure (test cable for a spare monitor) would not have affected device performance. Therefore, the requested extension isjustified.
The instruments used for MSL Radiation High are identical to the instruments approved by the NRC for a permanent.24 month cycle extension 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.
l Based on the above discussion, the NRC's prior approval of extensions for identical
)
. instrumentation and the fact that historical data for the radiation monitors shows that i
recalibration has not been required for a period of time greater than 24 months, it is
)
concluded that the impact on instrument availability, if any, is small as a result of the
. one-time surveillance interval extension and the requested extension is justified.
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Attachment I to NRC-98-0040 Page 45 ENCLOSURE 4 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.1.2 LOGIC SYSTEM FUNCTIONAL TEST AND SIMULATED AUTOMATIC OPERATION REACTOR PROTECTION SYSTEM INSTRUMENTATION The equipment and components used in the design of systems requiring LSFT was chosen based on reliability as demonstrated by years of service in both the nuclear and non-nuclear industries. A review of the surveillance test history for the subject SR was performed to detect evidence of excessive random equipment or component failure rates and no such evidence was found. Based on this review and the redundant equipment in each of the subject systems, it was concluded that the impact of reducing the LSFT frequency on system availability is insignificant. This conclusion was supported by two independent studies. The first study was completed for the BWR Owners Group in 1989 and the second was completed for the Nuclear Regulatory Commission (NRC) in 1988. An evaluation of LSFT completed by General Electric Co. for the BWR Owners Group (BWR Owners Group Report EAS 25-0489, Evaluation of Logic System Functional Testing methods, July 1989) has shown that circuit unavailability due to removal from service for testing at power, is the largest contributor to total circuit unavailability. This study analyzed the effect of different surveillance intervals on various basic logic configurations with the net result being that for most logic configurations changing from a 6 month to a refuel cycle surveillance improves the total circuit availability. With the exception of ADS, in no case is the unavailability increased by an appreciable amount. This conclusion is based on the determination that for the subject system the unavailability is only increased appreciably after increasing the test interval beyond 90 months. For ADS, unavailability caused by logic system failure (assuming a test interval of 36 months) remains two orders of magnitude less than currently accepted unavailability caused by valve actuation solenoid failure. Therefore, the additional unavailability due to increased test interval is insignificant in comparison to the existing system unavailability.
A historical search of the 18 month surveillance tests for these components for the last refueling outage was performed. The search criteria was to identify failed or partially failed tests, each failed or partially failed test was to be reviewed and evaluated. The purpose of this evaluation was to demonstrate that the increased functional test I
surveillance interval would not increase the period a function would be unavailable. No failures oflogic system comlanents were identified in this review.
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:
y Attachment I to NRC-98-0040 Page 46
" 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 Fermi 2, therefore, the surveillance interval extension isjustified.
The Reactor Mode Switch as well as most manual actuation and manual inhibit switches are considered to be system logic components since their failure modes and, generally, failure history are more closely related to circuit components than standard mechanical components. Based on this classification thejustification for the Reactor Mode Switch is presented as follows.
Table 4.3.1.1-1 Item 11 This test verifies the Reactor Mode Switch Shutdown' Scram and Scram Bypass logic associated with the RPS function properly. The reactor mode switch scram function is not required to protect the fuel or nuclear boundaries. The RPS functions independently from the mode switch. Based on the above discussion and the reliability oflogic it is
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concluded that the impact, if any, on the mode switch availability, is small as a result of o
i this change. A review of the history of the SR results demonstrates that there is no evidence of any failures which would invalidate this conclusion.
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Page 47 ENCLOSURE 5 l
JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.2.1 Table 4.3.2.1-1 ISOLATION ACTUATIONINSTRUMENTATION Justification for extending Items 1.a.1,1.a.2,1.a.3,1.b,1.c.3,2.e,5.a,6.a, and 6.b J
The Reactor Vessel Level - Level 1,2,3 and 8 and drywell pressure setpoints were analyzed using the GE Instrument Setpoint Methodology, NEDC-31336. The method of calculation is similar to the calculations submitted to the NRC during the power uprate review that was accepted by NRC Safety Evaluation Report dated September 9,1992 on Fermi-2 docket. The GE Instrument Setpoint Methodology, NEDC-31336, has been accepted by NRC Safety Evaluation Report dated July 18,1995 on the Perry Nuclear
)
Power Plant docket. In addition, Rosemount published a report in February 1990,"30 Month Stability Specification For Rosemount 1152,1153,1154 Pressure Transmitters" (Rosemount Report D89000126, Revision A) which was accepted by NRC Safety
)
Evaluatien Report dated August 2,1993 on Peach Bottom Atomic Power Station docket.
This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance.
i The Fermi 2 design calculations contain a drin allowance for Rosemount 1153 transmitter equal to at least three consecutive six month vendor drift intervals. The previously published model 1153 six month drift interval value is of the same order of i
magnitude as the currently published thirty month drin interval value. Therefore, the calculations contain at least 30 months drift allowance. Associated loop trip units are l
checked on a quarterly basis. This check confirms that the trip units are maintained '
within allowances. As a result, the existing Fermi-2 design calculation includes an j
adequate allowance for drift and the extension has no potential impact on the plant safety analyses, i.e., no analytic limit changes are required. Plant calibration data was also l
reviewed and found to be within the design allowances. Therefore, the extension of the surveillance interval is justified.
Justification for extending Item 1.c.1 For the Main Steam (MSL) Line Radiation - High instruments, correct operation (except the detector) is confirmed by a quarterly channel functional check and a channel check every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. These tests will identify any potential drift of these instruments (except for the detector and cable) during this period of time. The quarterly and twice daily surveillance tests, monitoring by plant staff, and the fact that the ion chamber detectors and cable are not considered susceptible to drift ensure that there is no potential impact on the subject instrument availability by allowing the additional time inteival of source calibration of the existing radiation monitors.
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~ Attachment I to NRC-98-0040 -
1 Page 48 The surveillance allowance for the MSL radiation instrument is 30 percent of span. A review of the surveillance tests indicates that drift for the MSL radiation monitors was 19 percent for the worst case condition. A further review of the MSL radiation monitors calibration history indicates that the detectors have never been adjusted to compensate for drift. Therefore, the drift history for these components documents the ability of the MSL monitors to perform for multiple cycles without adjustment and with performance remaining within expected bounds.
I The instruments used for MSL Radiation High are identical to the instruments approved j
- by the NRC for a permanent 24 month cycle extension 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.
Based on the above discussion, the NRC's prior approval ofidenticalinstrumentation and the fact that historical data for the radiation monitors shows that recalibration has not been required for a period of time greater than 24 months, it is concluded that the impact on instrument availability, if any, is small as a result of the one-time surveillance interval extension and the requested extension is justified.
Justification for extending Items 1.c.2 and 1.e The Main Steam Line Pressure and Condenser Pressure setpoints were analyzed for an extended surveillance interval using the GE Instrument Setpoint Methodology, NEDC-31336. The method of calculation is similar to the calculations submitted to the NRC during the power uprate review that was accepted by NRC Safety Evaluation Report dated September 9,1992 on Fermi-2 docket. The GE Instrument Setpoint Methodology, NEDC-31336 has been accepted by NRC Safety Evaluation Report dated July 18,1995 on the Perry Nuclear Power Plant docket.
. The evaluation conservatively considered six consecutive 6 months drift intervals for the Rosemount 1151 transmitters. Therefore the calculations contain at least 36 months drift l
. allowance. Associated loop trip units are checked on a quarterly basis. This check confirms that the trip units are maintained within allowances. As a result, the existing Fermi-2 setpoint includes an adequate allowance for drift and the extension has no potential impact on the plant safety analyses, i.e., no analytic limit changes are required.
l Plant calibration data was also reviewed and found to be within the design allowances.
Therefore, the extension of the surveillance interval is justified.
I o
Attachment I to NRC-98-0040 Page 49 Justification for extending Item 1.d and 1.f The Main Steam Line Tunnel and Turbine Building Area Temperature setpoints were analyzed for an extended surveill uce interval using the GE Instrument Setpoint Methodology, NEDC-31336. The method ofcalculation is similar to the calculations
-l submitted to the NRC during the power uprate review that was accepted by NRC Safety -
l Evaluation Report dated September 9,1992 on Fermi-2 docket. The GE setpoint methodology has been accepted by NRC Safety Evaluation Report dated July 18,1995 on
' the Perry Nuclear Power Plant docket.
The Main Steam Line Tunnel and Turbine Building Area Temperature calibration procedure performs an operability. check of the RTD's and calibrates the remaining loop components (Tech Spec Definition 1.4 states "... Calibration ofinstrument channels with
- resistance temperature detectors (RTD) or Thermocouple sensors shall consist of-verification of operability of the sensing element and adjustment, as necessary, of the L
remaining adjustable devices in the channel."). The loop operability is also verified in l-shiftly channel checks. Associated loop trip units are checked on a quarterly basis. This
]
check confirms that the trip units are maintained within allowances. As a result, the Fermi-2 setpoint includes an adequate allowance for drift and has no potential impact on plant safety analyses, i.e., no analytic limit changes are required.' Plant calibration data was reviewed and found to be within the design allowances. Therefore, the extension of '
the surveillance interval is justified.
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Attachment I to NRC-98-0040 Page 50 ENCLOSURE 6 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.2.2 LOGIC SYSTEM FUNCTIONAL TEST AND SIMULATED AUTOMATIC OPERATION ISOLATION ACTUATION INSTRUMENTATION SR 4.3.2.1 Justification for extending SR 4.3.2.2:
. 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' 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."
Justification for extending Table 4.3.2.1-1 Items 1.h,2.d, and 5.c The Reactor Mode Switch as well as most manual actuation and manual inhibit switches are considered to be system logic components since their failure modes and generally failure history are more closely related to circuit components than standard mechanical.
1 components. Based on this classification thejustification for the extension of the surveillance for manual initiation and manual inhibit switches is identical to the above discussion.
The evaluation above is applicable to Fermi 2. A historical search of the 18 month l
. surveillance tests for these components for the last refueling outage was performed. The l
search criteria was to identify failed or partially failed tests for further review and evaluation.
L The purpose of this evaluation was to demonstrate that the increased functional test l
surveillance interval would not increase the period a function would be unavailable. No L
problems were identified by this review. Based on the previous discussion, the historical
. failure review and evaluation, and the small impact on safety function availability, the l
requested extension isjustified.
' to NRC-98-0040 Page 51 ENCLOSURE 7 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.3.1 Table 4.3.3.1-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION Justification for extending Item 1.a,1.b,1.c,2.a,2.b,2.c,2.d,2.e,3.a,3.b,4.a,4.f
. The Reactor Vessel Level - Levels 1,2, and 8, Drywell Pressure, Reactor Vessel Dome Pressure, RHR (LPCI Mode) Riser Differential Pressure, and Recirculation Pump Differential Pressure setpoints were evaluated using the GE Instrument Setpoint Methodology, NEDC-31336. The method of calculation is similar to the calculations submitted to the NRC during the power uprate review that was accepted by NRC Safety Evaluation Report dated September 9,1992 on the Fermi-2 docket. The GE Instrument Setpoint Methodology, NEDC-31336, has been accepted by NRC Safety Evaluation
' Report dated July 18,1995 on the Perry Nuclear Power Plant docket. In addition, Rosemount published a report in February 1990,"30 Month Stability Specification For Rosemount 1152,1153,1154 Pressure Transmitters"(Rosemount Report D89000126, Revision A) which was accepted by NRC Safety Evaluation Report dated August 2,1993
[
on the Peach Bottom Atomic Power Station docket. This report supported the extension t
'of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance.
' The Fermi 2 design calculations contain a drift allowance for Rosemount 1153 transmitters equal to at least three consecutive six month vendor drift intervals. The previously published model 1153 six month drift interval value is of the same order of magnitude as the currently published thirty month drift interval value. Therefore, the
- calculations contain at least 30 months drift allowance. Associated loop trip units are checked on a quarterly basis. This check confirms that the trip units are maintained within allowances. As a result, the existing Fermi-2 design calculation includes an adequate allowance for drift and the extension has no potential impact on the plant safety analyses, i.e., no analytic limit changes are required. Plant calibration data was also l
reviewed and found to be within the design allowances. Therefore, the extension of the surveillance interval is justified.
Justification for extending Item 4.h The ECCS Drywell Pressure - High Bypass Timer setpoint was analyzed for an extended surveillance interval using the GE Instrument Setpoint Methodology, NEDC-31336. The method of calculation is similar to the calculations submitted to the NRC during the power uprate review that was accepted by NRC Safety Evaluation Report dated September 9,1992 on the Fermi-2 docket. The GE Instrument Setpoint Methodology, NEDC-31336 has been accepted by NRC Safety Evaluation Report dated July 18,1995 on the Perry Nuclear Power Plant docket.
i l
l
.-e
,- to NRC-98-0040 Page 52 A setpoint evaluation was perfonned that analyzed an accuracy / drift allowance of two times the vendor accuracy and found the results to be acceptable. Vendor performance specifications do not define a separate drift allowance and the setpoint is not considered to be dependent upon the surveillance interval; however, an evaluation was performed assuming that the increased interval would degrade the setpoint repeatability. As a result,
{
the existing Fermi-2 setpoint includes an adequate allowance for drift. Plant calibration data was also reviewed and found to be within the design allowances. Therefore, the extension of the surveillance interval is justified.
A historical search of the 18 month surveillance tests for these instruments for the last f
refueling outage was perfonned. The search criteria was to identify failed or partially thiled tests for further review and evaluation. The purpose of this evaluation was to demonstrate that the increased surveillance interval would not increase the period a function would be unavailable. No problems were identified by this review. Based on the previous discussion, i
the historical failure review and evaluation, and the small impact on safety function 1
availability, the requested extension is justified.
I I
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. to NRC-98-0040 Page 53 ENCLOSURE 8 l
JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.3.2 i
LOGIC SYSTEM FUNCTIONAL TEST AND SIMULATED AUTOMATIC I
OPERATION EMERGENCY CORE COOLING SYSTEM l
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, l
smveillance 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 l
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 i
safety system unavailability."
Justification for extending Table 4.3.3.1 1 Items 1.d, and 2.h l
Most manual actuation and manual inhibit switches are considered to be system logic l
components since their failure modes and generally failure history are more closely related to circuit components than standard mechanical components. Based on this classification, the justification for the extension of the surveillance for manual initiation and manual inhibit switches is identical to the above discussion.
l The evaluation above is applicable to Fermi 2. A historical search of the 18 month surveillance tests for these instruments for the last refueling outage was perfonned. The search criteria was to identify failed or partially failed tests for further review and evaluation.
The purpose of this evaluation was to demonstrate that the increased functional test surveillance interval would not increase the period a function would be unavailable. No problems were identified by this review. Based on the previous discussion, the historical failure review and evaluation, and the small impact on safety function availability, the requested extension isjustified.
l l
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Attachment I to NRC-98-0040 Page 54 ENCLOSURE 9 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECH ~NICAL SPECIFICATION SR 4.3.3.3 RESPONSE TIME TESTING EMERGENCY CORE COOLING SYSTEM This response time testing applies to the ECCS system response times. For the Core Spray System, High Pressure Coolant Injection and Low Pressure Coolant Injection, pump, valve and flow testing is performed on a more frequent basis in accordance with the Inservice Testing Program during the operating cycle to ensure that these assumed ECCS functions are available. Although these tests do not ensure the response times for the ECCS initiation, this testing in combination with Emergency Diesel Generator Testing would indicate any significant system slow responses.
The extension would have no substantial measurable effect on plant safety because:
There are redundant ECCS systems powered by different methods which a.
can perform the required safety functions.
b.
The response time failure probability is a very small fraction of the total ECCS failure probability.
i i
c.
Failure of ECCS components in the sluggish responding mode does not invalidate the components' ability to perform its safety function.
Extension of the Peach Bottom Atomic Power Station Units 2 and 3 surveillance intervals for response time testing for ECCS systems from 18 to 24 months was accepted by the NRC in the Safety Evaluation Report dated August 2,1993.
A historical search of the 18 month surveillance for response time testing for the last refueling outage was performed. The search criteria was to identify failed or partially failed tests for further review and evaluation. The purpose of this evaluation was to demonstrate that the increased surveillance interval would not increase the period a function would be unavailable. No ECCS response time test failures were identified in this review.
i Based on the abovu, a one-time extension of the ECCS response time testing surveillance intervalisjustified.
l l
i Attachment I to NRC-98-0040 Page 55 ENCLOSURE 10 l
JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION l
TECHNICAL SPECIFICATION SR 4.3.4 Table 4.3.4-1 ATWS RECIRCULATION PUMP TRIP INSTRUMENTATION Justification for extending Item 1 The Reactor Vessel Level - Levels 1,2, and 8 setpoints were analyzed using the GE Instrument Setpoint Methodology, NEDC-31336. The method of calculation is similar to the calculations submitted to the NRC during the power uprate review that was accepted by NRC Safety Evaluation Report dated September 9,1992 on Fermi-2 docket. The GE I
Instrument Setpoint Methodology, NEDC-31336, has been accepted by NRC Safety
)
Evaluation Report dated July 18,1995 on the Perry Nuclear Power Plant docket. In addition, Rosemount published a repo 1 in February 1990, "30 Month Stability Specification For Rosemount 1152,1153,1154 Pressure Transmitters"(Rosemount Report D89000126, Revision A) which was accepted by NRC Safety Evaluation Report dated August 2,1993 on the Peach Bottom Atomic Power Station docket. This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drin allowance.
The Fermi 2 design calculations contain a drift allowance for Rosemount 1153 transmitters equal to at least three consecutive six month vendor drift intervals. The previously published model 1153 six month drift interval value is of the same order of magnitude as the currently published thirty month drift interval value. Therefore, the calculations contain at least 30 months drin allowr.nce. Associated loop trip units are checked on a quarterly basis. This check confirms that the trip units are maintained j
within allowances. As a result, the existing Fermi-2 design calculation includes an adequate allowance for drin and the extension has no potential impact on the plant safety analyses, i.e., no analytic limit changes are required. Plant calibration data was also reviewed and found to be within the design allowances. Therefore, the extension of the surveillance interval is justified.
l
)
Justification for extending Item 2 The high Reactor Vessel pressure setpoints were analyzed using the GE Instrument l
Setpoint Methodology, NEDC-31336. The specific calculation for Reactor pressure instrument channels was submitted to the NRC during the power uprate review that was act epted by NRC Safety Evaluation Report dated September 9,1992 on Fermi-2 docket.
TF e GE Instrument Setpoint Methodology, NEDC-31336, has been accepted by NRC Snfety Evaluation Report dated July 18,1995 on the Perry Nuclear Power Plant docket.
in addition, Rosemount published a report in February 1990,"30 Month Stability l
Specification For Rosemount 1152,1153,1154 Pressure Transmitters"(Rosemount Report D89000126, Revision A) which was accepted by NRC Safety Evaluation Report
' Attachment 1 to NRC-98-0040 Page 56 dated August 2,1993 on the Peach Bottom Atomic Power Station docket. This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance.
The Fermi 2 design calculations contain a drift allowance for Rosemount 1153 I-transmitter equal to at least three consecutive six month vendor drin intervals. The previously published model 1153 six month drift interval value is of the same order of magnitude as the currently published thirty month drift interval value. Therefore, the calculations contain at least 30 months drift allowance. Associated loop trip units are checked on a quarterly basis..This check confirms that the trip units are maintained
-. within allowances. As a result, the existing Fermi-2 design calculation includes an L
. adequate allowance for drift and the extension has no potential impact on the plant safety analyses, i.e., no analytic limit changes are required. Plant calibration data was also l
. reviewed and found to be within the design allowances. Therefore, the extension of the surveillance interval is justified.
A historical search of the 18 month surveillance tests for these instruments for the last refueling outage was perfonned. The search criteria was to identify failed or partially failed tests for further review and evaluation. The purpose of this evaluation was to demonstrate that the increased calibration surveillance interval would not increase the period a component would be unavailable. No failures of components associated with the extended interval were identified. ; Based on the previous discussion, the historical failure review and evaluation and evaluation, and the small impact on safety function availability the requested extension is justified.
l l
. Attachment I to
)
{
Page 57 ENCLOSURE 11'
- JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.4.2 l
LOGIC SYSTEM FUNCTIONAL TEST AND SIMULATED AUTOMATIC l
OPERATION ATWS RECIRCULATION PUMP TRIP INSTRUMENTATION l
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 l_
safety system unavailability."
A historical search of the 18 month surveillance tests for these components for the last refueling outage was performed. The search criteria was to identify failed or partially failed l
tests for further review and evaluation. The pmpose of this evaluation was to demonstrate that the increased surveillance interval would not increase the period a component would be unavailable. No failures of components associated with the extended interval were identified.
Based on the previous discussion, the historical failure review and evaluation, and the small impact on safety function availability the requested extension isjustified.
l l
f l
3 Attachment.1 to -
- Page 58 ENCLOSURE 12 l
JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.5.1 Table 4.3.5.1-1 REACTOR CORE ISOLATION COOLING INSTRUMENTATION i
- Justification for extending Item a and b The Reactor Vessel Level - Levels 1,2, and 8 setpoints were analyzed using the GE I
Instrument Setpoint Methodology, NEDC-31336. The method of calculation is similar to l
the calculations submitted to the NRC during the power uprate review that was accepted by NRC Safety Evaluation Report dated September 9,1992 on Fermi-2 docket. The GE
- Instrument Setpoint Methodology, NEDC-31336, has been accepted by NRC Safety
. Evaluation Report dated July 18,1995 on the Perry Nuclear Power Plant docket. In addition, Rosemount published a report in February 1990,"30 Month Stability Specification For Rosemount 1152,1153,1154 Pressure Transmitters"(Rosemount Report D89000126, Revision A) which was accepted by NRC Safety Evaluation Report L
dated August 2,1993 on the Peach Bottom Atomic Power Station docket. This report l
supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance.
V L
The Fermi 2 design calculations contain a drift allowance for rasemount 1153 transmitter equal to at least three consecutive six month vendor drift intervals. The L
previously published model 1153 six month drift interval value is of the same order of magnitude as the currently published thirty month drift interval value. Therefore, the calculations contain at least 30 months drift allowance. Associated loop trip units are checked on a quarterly basis. This check confirms that the trip units are maintained within allowances. As a result, the existing Fermi-2 design calculation includes an i
adequate allowance for drift and the extension has no potential impact on the plant safety j
analyses, i.e., no analytic limit changes are required. Plant calibration data was also 1
l reviewed and found to be within the design allowances. Therefore, the extension of the l
surveillance interval is justified.
A historical search of the 18 month surveillance tests for these instruments for the last refueling outage was performed. The search criteria was to identify failed or partially failed tests for further review and evaluation. The purpose of this evaluation was to demonstrate that the increased calibration surveillance interval would not increase the period a component L'
would be' unavailable. No failures of components associated with the extended interval were identified. Based on the previous discussion, the historical failure review and evaluation, and the small impact on safety function availability, the requested extension is justified.
I 1
j
!M - to NRC-98-0040 Page 59 i
L ENCLOSURE 13 l
JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.5.2 LOGIC SYSTEM FUNCTIONAL TEST ANi> SIMULATED AUTOMATIC OPERATION
- REACTOR CORE ISOLATION COOLING 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' reliabilities are not dominated by the reliabilities of the logic system, but by that l
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 l
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."
A historical search of the 18 month surveillance tests for these components for the last -
l
- refueling outage was performed. The search criteria was to identify failed or partially failed tests for further review and evaluation. The purpose of this evaluation was to demonstrate
- that the increased surveillance interval would not increase the period a component would be l
unavailable. No failures of components associated with the extended interval were identified.
Based on the previous discussion, the historical failure review and evaluation, and the small impact on safety function availability, the requested extension is justified.
L l
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-_____--__________-._m_
_-____..____._._m._
t=
. Attachment 1 to.
. NRC-98-0040 Page 60 ENCLOSURE 14 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.6 Table 4.3.6-1 CONTROL ROD BLOCK INSTRUMENTATION Justification for extending Item 5.b and 7 '
The Reactor Mode Switch, as well as most manual actuation and manual inhibit switcaes, are considered to be system logic components since their failure modes and, generally, failure history are more closely related to circuit components than standard mechanical components (i.e., pumps and valves). Based on this classification, the justification for the l
extension of the surveillance for manual initiation and manual inhibit switches is identical i
to the Logic System Functional Test discussion provided in Enclosure 16. A review of the surveillance history during RFO5 for these components did not indicate any failures.-
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Attachment I to NRC-98-0040 a
- Page 61 ENCLOSURE 15
- JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION
' TECHNICAL SPECIFICATION SR 4.3.7.5 Table 4.3.7.5-1 i
POST ACCIDENT MONITORING INSTRUMENTATION _
Justification for extending Item 1 The Accident Monitoring Reactor Pressure instrumentation was analyzed using the GE Instrument Setpoint Methodology, NEDC-31336. The GE Instrument Setpoint Methodology, NEDC-31336, has been accepted by NRC Safety Evaluation Report dated July 18,1995 on the Perry Nuclear Power Plant docket. In addition, Rosemount
- published a report in February 1990,"30 Month Stability Specification For Rosemount 1152,1153,1154 Pressure Transmitters"(Rosemount Report D89000126, Revision A) which was accepted by NRC Safety Evaluation Report dated August 2,1993 on the l
Peach Bottom Atomic Power Station docket. This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance.
The Fermi 2 design calculations contain a drift allowance for Rosemount 1152 transmitter equal to at least three consecutive six month vendor drift intervals. The previously published model 1152 six month drift interval value is of the same order of
. magnitude as the currently published thirty month drift interval value.- Therefore, the calculations contain at least 30 months drift allowance. The transmitter provides input to recorders. Channel check is performed monthly. This check confirms that the indicating channel performs within allowances. As a result, the existing Fermi-2 design calculation
)
includes an adequate allowance for drift and the extension has no potential impact on the plant ' safety analyses.' Plant calibration data was also reviewed and found to be within the design allowances. Therefore, the extension of the surveillance interval isjustified.
Justification for extending Ite:n 2.a l_
The Reactor Fuel Zone indication is supplied by the same instruments which supply the l
l-Reactor Vessel Level - Levels 1,2, and 8 setpoints this transmitters drift was analyzed l
using the GE Instrument Setpoint Methodology, NEDC-31336. The GE Instrument
)
Setpoint Methodology, NEDC-31336, has been accepted by NRC Safety Evaluation l
Report dated July 18,1995 on the Perry Nuclear Power Plant dr
+ In addition, Rosemount published a report in February 1990, "30 Month Sta..ty Specification For
- Rosemount 1152,1153,1154 Pressure Transmitters"(Rosemount Report D89000126, Revision' A) which was accepted by NRC Safety Evaluation Report dated August 2,1993
- on the Peach Bottom Atomic Power Station docket. This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a l
reduction in the drift allowance.
Attachment I to NRC-98-0040 Page 62 The Fermi 2 design calculations contain a drift allowance for Rosemount 1153 transmitter equal to at lead three consecutive six month vendor drift intervals. The previously published model i153 six month drift interval value is of the same order of magnitude as the currently published thirty month drift interval value. Therefore, the calculations contain at least 30 months drift allowance. The transmitter provides input to recorder / indicators. Channel check is performed monthly. This check confirms that the indicating channel performs within allowances. As a result, the existing Fermi-2 design calculation includes an adequate allowance for drift and the extension has no potential impact on the plant safety analyses. Plant calibration data was also reviewed and found to be within the design allowances. Therefore, the extension of the surveillance interval isjustified.
Justification for extending Item 2.b The Reactor Vessel Water Level instrumentation was analyzed using the GE Instrument Setpoint Methodology, NEDC-31336. The GE Instrument Setpoint Methodology, NEDC-31336, has been accepted by NRC Safety Evaluation Report dated July 18,1995 on the Perry Nuclear Power Plant docket. In addition, Rosemount published a report in February 1990,"30 Month Stability Specification For Rosemount 1152,1153,1154 Pressure Transmitters"(Rosemount Report D89000126, Revision A) which was accepted by NRC Safety Evaluation Report dated August 2,1993 on the Peach Bottom Atomic Power Station docket. This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance.
The Fermi 2 design calculations contain a drift allowance for Rosemount 1153 transmitter equal to at least three consecutive six month vendor drift intervals. The previously published model 1153 six month drift interval value is of the same order of magnitude as the currently published thirty month drift interval value. Therefore, the calculations contain at least 30 months drift allowance. The transmitter provides input to recorders. Channel check is performed monthly. This check confirms that the indicating channel performs within allowances. As a result, the existing Fermi-2 design calculation includes an adequate allowance for drift and the extension has ao potential impact on the plant safety analyses. Plant calibration data was also reviewed and found to be within the design allowances. Therefore, the extension of the surveillance interval is justified.
Justification for extending Item 12 For the Containment High Range Radiation Monitor (CliRRM) instruments, correct operation is confirmed by a monthly channel functional check (except the detector).
These tests will identify any potential drift of these instruments (except for the detector and cable) during this period of time. The quarterly and twice daily surveillance tests, monitoring by plant staff, and the fact that the ion chamber detectors and cable are not considered susceptible to drift ensure that there is no potential impact on the subject i
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' Attachment I to L
NRC-98-0040 Page 63.
11 instrument availability by allowing the additional time interval of source calibration of l
l the existing radiation monitors.
The surveillance allowance for the CHRRM instrument is 20 percent of span. A review of the surveillance tests indicates that the drift for the CHRRM was 23 percent for the worst case condition. This dria was isolated to a single data point and when removed as an outlier the remaining data points indicated a dria of 18 percent. A further review of
. the CHRRM calibration history indicates that the detectors have never been adjusted to
. compensate for drin. Therefore, the drin history for these components documents the
)
ability of the CHRRM to perform for multiple cycles without adjustment and with
. performance remaining within expected bounds.
l' The instruments used for CHRRM are identical to the instruments approved by the NRC for a permanent 24 month cycle extension in the NRC Safety Evaluation Report (dated i
L August 2,1993) related to extension of the Peach Bottom Atomic Power Station, Unit j
' Numbers 2 and 3, surveillance intervals from 18 to 24 months.
Based on the abovn discussion, the NRC's prior approval ofidentical instrumentation and the fact that historical data for the radiation monitors shows that recalibration has not been required for a period of time greater than 24 months, it is concluded that the impact
- on instrument availability, if any, is small as a result of the one-time surveillance interval extension, and the requested extension isjustified.
Justification for extending Item 16
' Limit switches, which provide'the position indication for these valves, are mechanical devices that require mechanical adjustment only; dria is not applicable to these devices.
Therefore, an increase in the surveillance interval to accommodate one time extended operating cycle does not affect the limit switches with respect to drift.
A h_istorical search of the 18 month surveillance tests for these instruments for the last j
refueling outage was performed. The search criteria was to identify failed or partially failed tests. Each failed or partially failed test was to be reviewed and evaluated. The purpose of this evaluation was to demonstrate that the increased calibration surveillance interval would not increase the period an instrument would be unavailable. No problems were identified by l
r this review. Based on the previous discussion, the historical failure review and evaluation, and the small impact on safety ftmetion availability, the requested extension is justified.
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Attachment ' to NRC-98-0040 Page 64 i
ENCLOSURE 16 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION l
TECHNICAL SPECIFICATION SR 4.3.9.1 Table 4.3.9.1-1 FEEDWATER TURBINE TRIP SYSTEM INSTRUMENTATION 4
l Justification for extending Item a l
The Reactor Vessel Level - Level 3 and 8 setpoints were analyzed using the GE Instrument Setpoint Methodology, NEDC-31336. The method of calculation is similar to l
the calculations submitted to the NRC during the power uprate review that was accepted by NRC Safety Evaluation Report dated Septemoer 9,1992 on the Fermi-2 docket. The GE Instrument Setpoint Methodology, NEDC-31336, has been accepted by NRC Safety Evaluation Report dated July 18,1995 on the Perry Nuclear Power Plant docket. In addition, Rosemount published a report in February 1990, "30 Month Stability Specification For Rosemount 1152,1153,1154 Pressure Transmitters"(Rosemount Report D89000126, Revision A) which was accepted by NRC Safety Evaluation Repon dated August 2,1993 on the Peach Bottom Atomic Power Station docket. This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the dria allowance.
The Fermi 2 design calculations contain a drin allowance for Rosemount 1153 transmitter equal to at least three consecutive six month vendor drin intervals. The previously published model 1153 six month drift interval value is of the same order of magnitude as the currently published thiny month drin interval value. Therefore, the calculations contain at least 30 months drin allowance. Associated loop trip units are checked on a quarterly basis. This check confirms that the trip units are maintained within allowances. As a result, the existing Fermi-2 design calculation includes an adequate allowance for drin and the extension has no potential impact on the plant safety analyses, i.e., no analytic limit changes are required. Plant calibration data was also reviewed and found to be within the design allowances.
A historical scarch of the 18 month surveillance tests for these instruments for the last refueling outage was performed. The search criteria was to identify failed or partially failed tests for further review and evaluation. The purpose of this evaluation was to demonstrate that the increased calibration surveillance interval would not increase the period an instrt. ment would be unavailable. No problems were identified by this review.
Based on the previous discussion, the historical failure review and evaluation, and the small impact on safety ftmetion availability the requested extension isjustified.
s Attachment I to NRC-98-0040 l
Page 65 ENCLOSURE 17 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.9.2 l
LOGIC SYSTEM FUNCTIONAL TEST AND SIMULATED AUTOMATIC l
OPERATION i
FEEDWATER TURBINE TRIP SYSTEM l^
j As stated in the NRC Safety Evaluation Report (dated August 2,1993) related to L
extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
l
" Industry reliability studies for boiling water reactors (BWRs), prepared by the L
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 ftmetional test interval represents no significant change in the overall safety system unavailability."
L The evaluation above is applicable to Fermi 2. A historical search of the 18 month surveillance tests for these instruments for the last refueling outage was performed. The l
search criteria was to identify failed or partially failed tests for further review and evaluation.
The pmpose of this evaluation was to demc,nstrate that the increased surveillance interval l
would not increase the period an instrument would be unavailable. No problems were l
identified by this review. Based on the previon.; discussion, the historical failure review and evaluation, and the small impact on safety function availability the requested extension is justified.
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Page 66 i
ENCLOSURE 18 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.11.1 Table 4.3.11.1-1 ALTERNATIVE SIIUTDOWN SYSTEM Justification for extending Item 7 Alternate Shutdown Reactor Water Level Indication was analyzed using the GE Instrument Setpoint Methodology, NEDC-31336. The GE Instrument Setpoint
)
Methodology, NEDC-31336, has been accepted by NRC Safety Evaluation Report dated July 18,1995 on the Perry Nuclear Power Plant docket. In addition, Rosemount published a report in February 1990, "30 Month Stability Specification For Rosemount i152,1153,1154 Pressure Transmitters"(Rosemount Report DS9000126, Revision A) which was accepted by NRC Safety Evaluation Report dated August 2,1993 on the Peach Bottom Atomic Power Station docket. This report suppoded the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drin allowance.
The Fermi 2 design calculations contain a drift allowance for Rosemount 1153 transmitters equal to at least three consecutive six month vendor drin intervals. The previously published model 1153 six month drin interval value is of the same order of magnitude as the currently published thirty month drin interval value. Therefore, the calculations contain at least 30 months drift allowance. The transmitter provides input to an indici Channel check is performed monthly. This check confirms that the indicati.s. channel perfonns within allowances. As a result, the existing Fermi-2 design calculation includes an adequate allowance for drift and the extension has no potential i
impact on the plant safety analyses. Plant calibration data was also reviewed and found l
to be within the design allowances. Therefore, the extension of the surveillance interval isjustified.
Justification for extending Item 8 The Alternative Shutdown System Reactor Pressure instrumentation was analyzed using the GE Instrument Setpoint Methodology, NEDC-31336. The GE Instrument Setpoint Methodology, NEDC-31336, has been accepted by NRC Safety Evaluation Report dated July 18,1995 on the Perry Nuclear Power Plant docket. In addition, Rosemount i
published a report in February 1990, "30 Month Stability Specification For Rosemount 1152,1153,1154 Pressure Transmitters"(Rosemount Report D89000126, Revision A) which was accepted by NRC Safety Evaluation Repon dated August 2,1993 on the Peach Bottom Atomic Power Station docket. This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drin allowance.
- l' Attachment I to
' NRC-98-0040 s
Page 67 The Fermi 2 design calculations contain a drift allowance for Rosemount 1153
- transmitters equal to at least three consecutive six month vendor drift intervals. The -
previously published model 1153 six month drift interval value is of the same order of magnitude as the currently published thirty month drift interval value. Therefore, the
. calculations contain at least 30 months drift allowance. The transmitter provides input to
- a recorder. Channel check is performed monthly. This check confirms that the indicating channel performs within allowances. As a result, the existing Fermi-2 design calculation includes an adequate allowance for drift and the extension has no potential impact on the plant safety analyses, i.e., no analytic limit changes are required. Plant calibration data was also reviewed and found to be within the design allowances.
A historical search of the 18 month surveillance tests for these instruments for the last refueling outage was performed. The search criteria was to identify failed or partially failed tests for further review and evaluation. The purpose of this evaluation was to demonstrate that the increased calibration surveillance interval would not increase the period a component would be unavailable. No failures were identified by this review.
Based on the previous discussion, the historical failure review and evaluation, and the
- small impact on safety function availability the requested extension isjustified.
}
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- NRC-98-0040 Page 68 ENCLOSURE 19 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.4.2.2 SAFETY RELIEF VALVES Justification' for extending SR 4.4.2.2.b The purpose of this test is to perform a Logic System Functional test, and simulated
. automatic operation of the entire low-low set function pressure actuation instrumentation.
l
-The low-low set logic is designed with redundancy and single-failure criteria; that is, no L
single electrical failure will (1) prevent any low-low set valve from opening, and (2) cause inadvertent seal-in oflow-low set logic.
Based on the above discussion, the logic design of the low-low pressure set provides redundancy and reliability which ensures the function will remain available during the
. extended operating cycle.
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' 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 L
tested on a more frequent 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 l
safety system unavailability."
L
.- A historical search of the 18 month surveillance tests for these surveillance requirements
' for the last refueling outage was performed. The search criteria was to identify failed or partially failed tests for further review and evaluation. The purpose of this evaluation l
.~was to demonstrate that the increased surveillance interval would not increase the period L
a component would be unavailable. No problems were identified by this review. Based on the previous discussion, the historical failure review and evaluation, and the small impact on safety function availability the requested extension is justified.
C
- Attachment 1 to NRC-98-0040 l
Page 69 ENCLOSURE 20 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.0.5 ASME CODE INSPECTION PROGRAM (ISI/IST) l i
Discussion for SR 4.0.5 l
ASME Section XI specifies periodic leak testing of system isolation valves. The Fermi 2 Inservice Inspection Program requires testing of the Non-Interruptible Air Supply (NIAS) isolation valves on an 18 mor.'.h basis. In the past, the isolation valves have consistently met the leak testing criteria. It addition, the extension requested is during the period when the plant will be in Operational Condition 4 or 5 and the number of required systems that NIAS supports (e.g., SGTS, Secondary Containmen.t Integrity, and Control Room Emergency Filtration System) is minimized.
When the reactor is shutdown, these systems e e only required during core alteration, operations within the potential for draining the reactor vessel, and movement of fuel in secondary containment. Since, these activities will most likely be completed by the time the surveillance becomes overdue on October 2,1998, the potential impact on safety is minimal. Therefore, the extension of 17 days for this surveillance is justified.
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4 Attachment I to 4
NRC-98-0040 Page 70 ENCLOSURE 21 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.5.1 i
EMERGENCY CORE COOLING SYSTEM OPERATION Justification for extending SR 4.5.1.c.1 For the Core Spray System, a pump, valve and flow test is performed on a more frequent basis in accordance with the Inservice Testing Program during the operating cycle to l
ensure that this assumed ECCS function is available. Although this test does not ensure the operability of the entire Core Spray logic, it does ensure the functional ability of the Com Spray Pumps, and a majority of the Core Spray System to produce the required ECCS flow at the required pressure. The portion of the system not tested on the more frequent basis is equivalent to the logic system and the testing would be equivalent to a logic system functional test.
l 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, i
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 the Fermi 2 surveillance interval extension. A historical search of the 18 month surveillance tests for these components for the last refueling outage was performed. The search criteria was to identify failed or partially failed tests for further review and evaluation. The purpose of this evaluation was to demonstrate that the increased surveillance interval would not increase the period a component would be unavailable. No failures were identified by this review. Based on the previous discussion, the historical failure review and evaluation, and the small impact on safety function availability the requested extension is justified.
l l
For the Low Pressure Coolant Injection (LPCI) System, a pump, valve and flow test is performed on a more frequent basis in accordance with the Inservice Testing Program during the operating cycle to ensure that this assumed ECCS function is available.
Although this test does not ensure the operability of the entire LPCI logie, it does ensure the functional ability of the LPCI Pumps, and a majority of the LPCI System to produce w _ ---
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t the required ECCS flow at the required pressure. The portion of the system not tested on the more frequent basis is equivalent to the logic system and the testing would be I
equivalent to a logic system functional test.
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' 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 I
is small relative to the probability of mechanical component failure, increasing the l
logic system functional test interval represents no significant change in the overall l
safety system unavailability."
i A historical search of the 18 month surveillance tests for these components for the last refueling outage was performed. The search criteria was to identify failed or partially failed I
tests for further review and evaluation. The purpose of this evaluation was to demonstrate that the increased surveillance interval would not increase the period a component would be unavailable. No failures of the components were identified by this review. Based on the previous discussion, the historical failure review and evaluation, and the small impact on safety function availability the requested extension is justified.
For the High Pressure Coolant Injection (HPCI) System, a pump, valve and flow test is performed on a more frequent basis in accordance with the Inservice Testing Program during the operating cycle to ensure that this assumed ECCS function is available.
Although this test does not ensure the operability of the entire HPCI logic, it does ensure the functional ability of the HPCI Pump, and a majority of the HPCI System to produce the required ECCS flow at the required pressure. The portion of the system not tested on the more frequent basis is equivalent to the logic system and the testing would be equivalent to a logic system functional test.
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' reliabilities are not dominated by the reliabilities of the logic system, but by that l
of the mechanical components, (e.g., pumps and valves), which are consequently l
tested on a more frequent basis. Since the probability of a relay or contact failure
- 4
' Attachment I to NRC-98-0040 Page 72 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 the Fermi 2 surveillance interval extension. A I
historical search of the 18 month surveillance tests for these components for the last refueling outage was performed. The search criteria was to identify failed or partially failed tests for further review and evaluation.. The purpose of this evaluation was to demonstrate that the j
increased surveillance interval would not increase the period a component would be unavailable. No failures were identified by this review. Based on the previous discussion,-
the historical failure review and evaluation, and the small impact on safety function availability the requested extension is justified.
Justification for extending SR 4.5.1.d.2.a
- The logic design of the Automatic Depressurization System (ADS) provides redundancy and reliability which ensures the function will remain available during the extended j
. operating cycle. SR 4.5.1.d.2.a requires a simulated automatic actuation of the system, which excludes actual valve operation. No extension in the surveillance interval is needed or requested for SR 4.5.1.d.2.b, in which the valves are actually opened.
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' 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." -
4 L
The evaluation above is applicable to the Fermi 2 surveillance interval extension. A historical search of the 18 month surveillance tests for these components for the last refueling outage was performed. The search criteria was to identify failed or partially failed tests for further review and evaluation. The purpose of this evaluation was to demonstrate that the I
increased surveillance' interval would not increase the period a component would be j
unavailable. No failures were identified by this review. Based on the previous discussion, l
. the historical failure review and evaluation, and the small impact on safety function availability the requested extension is justified.
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NRC-98-0040 Page 73 ENCLOSURE 22 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECIINICAL SPECIFICATION SR 4.6.3 PRIMARY CONTAINMENT ISOLATION VALVE FUNCTIONAL TEST Justification for extending SR 4.6.3.2 Most Primary Containment Isolation Valves (PCIVs) are cycled on a more frequent basis in accordance with the ISI/IST program during the operating cycle to ensure that the valves are capable of performing their isolation function. Although this test does not l
ensure the operability of the entire PCIVs Logic, it does ensure the functional ability of i
the PCIVs as assumed in the offsite dose calculations. The remaining portions of the Isolation valve circuit that is not tested during the ISI/IST program is equivalent to a logic system and the testing of these components would be equivalent to a logic system functional test.
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:
l
" Industry reliability studies for boiling water reactors (BWRs), prepared by the BWR Owners Group (NEDC-30936P) show that the overall safety systems' reliability 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 Fermi 2. A historical search of the 18 month surveillance tests for these valves for the last refueling outage was performed. The search criteria was to identify failed or partially failed tests for further review and evaluation. The purpose of this evaluation was to demonstrate that the increased surveillance interval would not increase the period a component would be unavailable. No problems were identified by this review. Based on the previous discussion, the historical failure review and evaluation, and the small impact on safety function availability the requested extension isjustified.
Attachment I to NRC-98-0040 Page 74 ENCLOSURE 23 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.6.5 SECONDARY CONTAINMENT
(
Justification for extending SR 4.6.5.1.d.1 and d.2 For Standby Gas Treatment System (SGTS) components, the fans and HEPA filters are verified to be operable on a more frequent basis in accordance with TS 4.6.5.3. The integrity of the secondary containment boundary and ability to maintain a vacuum is also verified on a more frequent basis in accordance with SR 4.6.5.1.a. Based on a historical review, past testing of the SGTS for SR 4.6.5.1.d.1 and d.2 has shown no indication of degradation or failure. Concerns during this testing have only been raised when the atmospheric pressure transmitters were affected by wind gusts that produced spikes in the differential pressure recorder. Ilowever, these spikes are not representative of bulk l
differential pressure conditions between the reactor building and atmosphere. The drawdown of secondary containment has consistently been accomplished within the 567 second time requirement of SR 4.6.5.1.d.1 and a 0.25 inch water gauge vacuum has been consistently maintained during previous tests. Based on the success of past testing, there is no reason to believe that degradation of components that are not able to be detected during other tests will occur. Therefore, the extension of the surveillance interval for the short period of time requested (19 days) is justified.
Justification for extending SR 4.6.5.2.b Secondary containment dampers are cycled on a more frequent basis in accordance with the Inservice Testing Program during the operating cycle to ensure that the dampers are capable of performing their safety function. Although this test does not ensure the operability of the entire circuit logic, it does ensure the functional ability of the dampers to actuate and establish the canditions in the secondary containment assumed in the offsite dose calculations. The remaining portions of the Isolation damper circuit that is not tested during the ISI/IST program is equivalent to a logic system and the testing of these components would be equivalent to a logic system functional test.
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, 1
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 i
l l
l
- to NRC-98-0040 Page 75 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."
A historical search of the 18 month surveillance tests for these valves for the last refueling outage was performed. The ecc.rch criteria was to identify failed or partially failed tests for further review and evaluation. The purpose of this evaluation was to demonstrate that the increased surveillance interval would not increase the period a component would be unavailable. No failures were identified by this review. Based on the previous discussion, the historical failure review and evaluation, and the small impact on safety function availability the requested extension is justified.
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I Attachment I to NRC-98-0040 Page 76 ENCLOSURE 24 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECIINICAL SPECIFICATION SR 4.7.1 SERVICE WATER SYSTEMS Justification for extending SR 4.7.1.2.b For the Emergency Equipment Cooling Water (EECW) System, a pump, valve and flow test is performed on a more frequent basis in accordance with the Inservice Testing Program during the operating cycle to ensure that this support safety function is available.
Although this test does not ensure the operability of the entire EECW System logic; it does ensure the functional ability of the EECW Pump, and a majority of the EECW System to produce the required EECW cooling flow to supported systems.
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' reliabilities are not dominated by the reliabilities of the logic system, but by that l
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."
A historical search of the 18 month surveillance tests for this system, for the last refueling outage was performed. The search criteria was to identify failed or partially failed tests for further review and evaluation. The purpose of this evaluation was to demonstrate that j
l the increased surveillance interval would not increase the period a component would be I
unavailable. No failures were identified by this review. Based on the previous discussion, historical failure review and evaluation, and the small impact on safety function availability the requested extension isjustified.
1 Justification for extending SR 4.7.1.3.b For the Emergency Equipment Service Water (EESW) System, a pump, valve and flow test is performed on a more frequent basis in accordance with the Inservice Testing Program during the operating cycle to ensure that this support safety function is available.
l Although this test does not ensure the operability of the entire EESW System logic,it I
does ensure the functional ability of the EESW Pump, and a majority of the EESW System to produce the required EESW cooling flow to supported systems.
l l
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' Attachment I to
-.t NRC-98-00401 Page 77 i
1 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, 1
surveillance intervals from 18 to 24 months:
)
i
" Industry reliability studies for boiling water reactors (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 3
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 Fermi 2. A historical search of the 18 month surveillance tests for this system, for the last refueling outage was performed. The search criteria was to identify failed or partially failed tests for further review and evaluation.
The purpose of this evaluation was to demonstrate that the increased surveillance interval 3
would not increase the period a component would be unavailable. No failures were identified by this review. Based on the previous discussion, the historical failure review and evaluation, and the small impact on safety function availability the requested i
extension isjustified.
' Justification for extending SR 4.7.1.4.b For the Diesel Generator Service Water (DGSW) System, a pump, valve and flow test is performed on a more frequent basis in accordance with the Inservice Testing Program during the operating cycle to ensure that this support safety function is available.
Although this test does not ensure the operability of the entire DGSW System logic, it does ensure the functional ability of the DGSW Pump, and a majority of the DOSW i
~ System to produce the required DOSW cooling flow to supported systems.
i 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:
i
" 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 l
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
~'
. Attachment I to l.
. NRC-98-0040 -
PageL78 logic system functional test interval represents no significant change in the overall l'
safety system unavailability."
I
~ The evaluation above is applicable to Fermi 2. A historical search of the 18 month
(
surveillance tests for this system, for the last refueling outage was performed. The search l.
criteria was to identify failed or partially failed tests for further review and evaluation. The purpose of this evaluation was to demonstrate that the increased surveillance interval would not increase the period a component would be unavailable. No failures were identified by
. this review that would adversely effect the system's ability to perform its safety function.
Based on the previous discussion, the historical failure review and evahiation, and the small impact on safety function availability the requested extension is justified.
L l.
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Page'79 l
i ENCLOSURE 25 l
JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION l
TECHNICAL SPECIFICATION SR 4.7.2 CONTROL ROOM EMERGENCY FILTRATION SYSTEM
' Justification for extending SR. 4.7.2.1.c.1, c.2, c.3, e.1, e.4 The Control Room Emergency Filtration System (CREFS) provides a suitable 1
environment for continuous personnel occupancy and ensures the operability of control I
room equipment and instruments under accident conditions. Proper operation of the system (i.e., operation with proper flow rates for makeup and exhaust) verifies the operability of the control room pressure boundary. The sysem is normally in standby -
- condition, thus gross plugging or fouling of the HEPA filters and charcoal adsorbers will be minimized.. In addition, the CREFS has redundant filter trains and fans which will ensure system availability in the event of a failure of one of the system components. TS 4.7.2.1.b requires operability of the main control room fans and verification of flow
- through the HEPA filter and charcoal adsorbers for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every 31 days. During this
)
testing, performance of the CREFS is demonstrated. This test would identify significant i
failures affecting CREFS operability, including failures to automatically. initiate.
Furthermore, the CREFS system is normally in standby, and as required by TS SR
!~
4.7.2.1.c, if there is any condition such as painting or maintenance on the filter train which would impact the operability of the CREFS system, the test to verify the operability of the CREFS system will be performed. Therefore, it is concluded that because of the redundancy in the CREFS and the other required SRs the impact, if any, of postponing these surveillance tests on system availability is small. A review of the
- history of the SR results for RF05 demon' trates that there is no evidence of any failures s
j-
[
. which would invalidate this conclusion. Therefore, this extension isjustified.
Justification for extending SR 4.7.2.1.e. 2 p
The logic used to initiate the CREFS has been previouslyjustified to allow its testing to be extended based on reliability studies presented in the BWR Owners Group topical report NEDC-30936P and has been accepted by the NRC (Reference NRC Safety Evaluation Report, dated August 2,1993, for the Peach Bottom Atomic Power Station, Units 2 and 3).
The automatic actuation surveillance interval isjustified to be extended based on the redundancy of the system involved, and the Inservice Testing performed during the I-Koperating cycle.
i
- The evaluation above is applicable to Fermi 2. A historical search of the 18 month surveillance tests for this system, for the last refueling outage was performed. The search i
t t.
I
. Attachment 1 to NRC-98-0040 Page 80 l-criteria was to identify failed or partially failed tests for further review and evaluation.
l The purpose of this evaluation was to demonstrate that the increased surveillance interval l
would not increase the period a component would be unavailable. No failures were identified by this review.. Based on the previous discussion, the historical failure review
(
and evaluation, and the small impact on safety function availability the requested -
. extension isjustified.
l
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NRC-98 0040 Page 81 ENCLOSURE 26 l
JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.7.4 l
REACTOR CORE ISOLATION COOLING SYSTEM FUNCTION TEST Justification for extending SR 4.7.4.c.1 The Reactor Core Isolation Cooling (RCIC) System pump, valve and flow test is performed on a more frequent basis in accordance with the Inservice Testing Program during the operating cycle to ensure that this assumed core cooling function is available.
Although this testing does not ensure the operability of the entire RCIC logic, it does
[
ensure functional capability of the RCIC pump and a majority of the RCIC System to produce the required RCIC flow at the required pressure. The portion of the system not tested on the more frequent basis is equivalent to a logic functional test.
As stated in the NRC Safety Evaluation Report (dated August 2,1993) related to the l
extension of the Peach Bottom Atomic Power Station, Unit Number 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 l
tested on a more frequent basis. Since the probability of a relay or contact failure j
is small compared to the probability of mechanical component failure, increasing i
the logic system functional test interval represents no significant change in the overall safety system unavailability."
I The evaluation above is applicable to the Fermi 2 surveillance interval extension. A l
historical search of the 18 month surveillance tests for these components for the last three refueling outages was performed. The search criteria was to identify failed or partially i
failed tests for further review and evaluation. The purpose of this evaluatien was to i
demonstrate that the increased surveillance interval would not increase the period a
[
component would be unavailable. No failures were identified by this review. Based on the previous discussion, the hhtorical failure review and evaluation, and the small impact on safety function availability, the requested extension of 54 days is justified.
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ENCLOSURE 27 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION I
TECHNICAL SPECIFICATION SR 4.7.5 s
SNUBBERS l
Justification for extending SR 4.7.5.e
- Technical Specification 4.7.5.e requires functional testing of snubbers at least once per 18 months, during plant shutdowns. Fermi has chosen to use the ten percent (10%) sample plan per T. S. section 4.7.5.c.1, therefore, SRs 4.7.5.e.2 and 4.7.5.e.3 do not apply.
ASME XI, subsection IWF-5000 requires that snubbers shall be tested each inspection period. The ASME Code does not specify a time frame for testing.
The primary snubber failure modes experienced at Fermi have been attributed to
. temperature and vibration induced degradation. Snubber degradation is, therefore, not L
simply time dependent, but rather a function of system operating time and time at -
L elevated temperatures for snubbers in high temperature areas (e.g., Drywell, steam l
tunnel).
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The reactor startup from RF05 began in November 1996. Between November 1996 and l'
the end of 1997, the reactor had 7.6 effective months of operation at elevated temperature and operating conditions (based on generator on-line time). Assuming continuous operation for the remainder of the cycle from January 1 to September 4,1998, an additional 9 months of operation is added which provides a total of 16.6 effective months of plant operation for Cycle 6. Based on the fact that this effective time at elevated a
temperature and operating conditions is less than the 18 months presently allowed by
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Technical Specifications and that previous snubber failures, as cited above, are attributed to system operating time and time at temperature, the requested extension will have no impact on safety. Therefore, the requested extension isjustified.
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I Attachment I to NRC-98-0040 Page 83 ENCLOSURE 28 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.8.1 DIESEL GENERATORS Justification for extending SR 4.8.1.1.2.e.1 During the operating cycle, the diesel generators are subjected to operational testing every 31 days and fast start testing every 184 days. This testing provides confidence ofdiesel generator operability and the capability to perform its intended function. This testing during the past cycles has not indicated any degradation of the diesel which would negate the extension of the surveillance as requested. In addition, the past internal inspections conducted in accordance with TS SR 4.8.1.1.2.e.1 have not revealed any degradation which would necessitate replacement ofintemal components, although, as a matter of course, several components have been replaced as preventive measures. Therefore, the impact, if any, on system reliability will be small from the one time extension.
Historical testing and surveillance testing during operations have proven the ability of the Diesel engines to start and operate under various load conditions. There is no reason to believe, based on diesel generator history and testing that the extension of this surveillance will have any effect on reliability. Therefore, this extension isjustified.
However, this SR may be performed on-line prior to RFO6. In that case this SR will not require an extension.
Justification for extending SR 4.8.1.1.2.e.2, e.3,4.a,4.b,5,6.a,6.b, and 7 The design of the offsite power to the plant essential busses provides a decreased likelihood that a total loss of offsite power will occur. However, if a total loss of offsite power were to occur and operation of the diesel generators was required, the requested extension would have minimal impact on the system failure probability. The extension of the surveillance interval for the diesel generator logic testing has in itself the same rationale for extension as LSFTs on other systems / components. Since the failure probability of the logic (relays, contacts, etc.) is reasoned, as documented in NEDC-30936P, to be less than the failure probability for the mechanical equipment (pumps, valves, etc.), the extension of the surveillance interval for the logic has minimal impact on the failure to function. And, since the mechanical components (diesel generators) are tested on a more frequent basis (i.e., monthly and 184 day by SR 4.8.1.1.2.a ), the probability of failure to function is further minimized.
Historical testing and surveillance testing during operations have proven the ability of the Diesel engines to start and operate under various load conditions. There is no reason to believe, based on diesel generator history and testing that the extension of this surveillance will have any effect on reliability. Therefore, this extension is justified.
I Attachment I to NRC-98-0040 Page 84 However, SRs 4.8.1.1.2.e.2 and e.3 may be performed on-line prior to RFO6. In that case these SR will not require an extension.
Justification for extending SR 4.8.1.1.2.e.8 During the operating cycle, the diesel generators are subjected to operational testing every i
31 days and fast start testing every 184 days. This testing provides confidence of diesel l
generator operability and the capability to perform its intended function. This testing during the past cycles has not indicated any degradation of the diesel which would negate the extension of the surveillance as requested. In addition, the past intemal inspections l
conducted in accordance with TS SR 4.8.1.1.2.e.1 have not revealed any degradation which would necessitate replacement ofinternal components, although, as a matter of l
course, several components have been replaced as preventive measures.
Historical testing and surveillance testing during operations have proven the ability of the Diesel engines to start and operate under various load conditions. There is no reason to l
believe, based on diesel generator history and testing that the extension of this 1
surveillance will have any effect on reliability. Therefore, this extension isjustified.
However, this SR may be performed on-line prior to RFO6. In that case this SR will not require an extension.
Justification for extending SR 4.8.1.1.2.e.9 Auto connected loads for the diesel generators have not substantially changed since RFO5. Any changes to diesel generator loading have been evaluated in accordance with the appropriate section of the Diesel Generator loading calculation. The tested loads I
during RFO5 combined with any changes since the refueling outage do not exceed the rating requirements of Regulatory Guide 1.9 Rev. 2. There is no reason to believe, based on diesel generator history, the Fermi 2 diesel loading calculation and testing that the extension of this surveillance will have any affect on availability.
Historical testing and surveillance testing during operations have proven the ability of the Diesel engines to start and operate under various load conditions. There is no reason to believe, based on diesel generator history and testing that the extension of this I
surveillance will have any effect on reliability. Therefore, this extension is justified.
Justification for extending SR 4.8.1.1.2.e.10,11,12 i
The design of the offsite power to the plant essential busses provides a decreased likelihood that a total loss of offsite power will occur. However, if a total loss of offsite l
power were to occur and operation of the diesel generators was required, the requested extension would have minimal impact on the system failure probability. The extension of
Attachment I to NRC-98-0040 Page 85 i
the surveillance interval for the diesel generator logic tecting has in itself the same rationale for extension as LSFTs on other systems / components. Since the failure probability of the logic (relays, contacts, etc.) is reasoned, as documented in NEDC-30936P, to be less than the failure probability for the mechanical equipment (pumps, valves, etc.), the extension of the surveillance interval for the logic has minimal impact on the failure to function. Also, since the mechanical components (diesel generators) are tested on a more frequent basis (i.e., monthly and 184 day by SR 4.8.1.1.2.a ), the probability of failure to function is further minimized. A historical review of the load sequencer operation indicated that the timing for the load sequencer for both divisions has been found to ahvays be in tolerance. Based on this good performance, surveillance extension to the end of the outage is recommended. Therefore, the extensions cf the surveillance intervals to reach the end date of RFO6 has minimal impact on the failure probability.
Historical testing and surveillance testing during operations have proven the ability of the Diesel engines to start and operate under various load conditions. There is no reason to believe, based on diesel generator history and testing that the extension of this surveillance will have any affect on reliability. Therefore, this extension isjustified.
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- to NRC-98-0040 Page 86 ENCLOSURE 29 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION l
TECHNICAL SPECIFICATION SR 4.8.2 BATTERIES Justification for extending 4.8.2.1.c.3 Test results for both RF03 and RF04 indicated no significant corrosion which affected battery connection resistance. All resistance measurements were far below the maximum allowed. In addition, if corrosion is observed during the 90 day surveillance, a battery connection resistance test is required to verify resistance is below the Tech Spec limit of 150 microhms.
An evaluation of any failures identified for RFO5 was performed to determine if the failure modes contained time based elements which could impact the one-time extension of the surveillance interval. This evaluation did not identify any failures. Based on the previous discussion, the historical failure review and evaluation, and the small impact on safety function availability the requested extension is justified.
Justification for extending SR 4.8.2.1.d The Fermi 2 division 1 and 2 batteries were capacity tested in May and June respectively of 1986 and April and May respectively of 1991. Capacity factors were greater than 100% for all batteries tested. Based on these capacity tests and the service tests during the other refueling outages, he service life has been determined by extrapolation. Based on the extrapolated rate of degradation for the worst case battery, the batteries will not reach 90% of the ma.qufacturer's rating until their 15th year of service (2001). Based on this, and the fact that the capacity test bounds the service test, the indicated surveillance extensions arejustified.
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Attachment I to NRC-98-0040 j
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ENCLOSURE 30 1
JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.8.4 BREAKER AND CONDUCTOR PROTECTION Justification for extending SR 4.8.4.2.a.l.a l
The Primary Containment Penetration Conductor Overcurrent Protective Devices for the reactor recirculation pumps penetration protection (overcurrent relays) were also evaluated for a 30 month period using a GE extrapolation method. The evaluated drift for these devices was found to be within Technical Specification requirements. Therefore, this one time surveillance interval extension isjustified for the Primary Containment Penetration Conductor Overcurrent Protective Devices.
For the channel calibration an evaluation of any failures identified for RF05 was performed to determine if the failure modes contained time based elements which could impact the one-time extension of the surveillance. The only failure identified by this review was associated with excessive instrument drift and was evaluated as a part of the instrument drift evaluation. Based on the previous discussion, the historical failure review and evaluation, and the small impact on safety function availability the requested extension isjustified Justification for extending SR 4.8.4.2.a.l.b A historical search review of the 18 month surveillance tests for SR 4.8.4.2.a.l.b for the
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last refueling outage was performed. The search criteria was to identify failed or partially I
failed tests. Each failed or partially failed test was to be reviewed and evaluated. The purpose of this evaluation was to demonstrate that the increased calibration surveillance interval would not increase the period a component would be unavailable. No failures were identified by this review. There were no failures of the breakers to open during these surveillance tests.
Excluding the breakers, the integrated system test of the primary containment penetration I
conductor overcurrent protective devices is essentially a Logic System functional test therefore the following discussion is provided.
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:
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" 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 l
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Attachment I to NRC-98-0040 Page 88 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 unavailabili y."
t Based on the surveillance performance history review and the conclusions above regarding logic system functional testing, extension of this surveillance interval is justified.
Justification for extending SR 4.8.4.5.a 1
A historical search of the 18 month surveillance tests for the calibration and functional testing of the Standby Liquid Control System protective relays and circuit breakers was i
performed. The search was to identify failed or partially failed tests. Each failed or partially failed test was to be reviewed and evaluated. The purpose of this evaluation was to demonstrate that the increased surveillance interval would not increase the period a component would be unavailable. The review and evaluation of calibration history concluded that the increased interval would have no effect on the availability of the safety function, and that the extension wasjustified. In addition, review of the channel functional test history of the circuit concluded that no failures have been previously
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experienced, and the function will not be affected by the short increase in the increase in j
the surveillance interval. Therefore, based on the above, the proposed extension is justified.
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l ATTACIIMENT 2 l
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FERMI 2 NRC DOCKET NO. 50-341 NRC LICENSE NO. NPF-43 REQUEST TO REVISE TECIINICAL SPECIFICATIONS "ONE-TIME TECHNICAL SPECIFICATIONS REVISION TO ALLOW EXTENSION OF TIIE FERMI 2 OPERATING CYCLE" 10 CFR 50.92 SIGNIFICANT IIAZARDS CONSIDERATION L
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' to NRC-98-0040 Page 2 10 CFR 50.92 SIGNIFICANT IIAZARDS CONSIDERATION Hasis For Significant IIazards Determination Detroit Edison has concluded that the proposed changes to the Fermi 2 TS, to facilitate a one-time extension in the Fermi 2 operating cycle, do not involve a Significant Hazards Consideration. In support of this determination, an evaluation of each of the three standards set forth in 10 CFR 50.92 is provided below.
1.
The proposed TS chances do not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed TS changes involve a one-time only change in the surveillance testing intervals to facilitate a one-time only change in the Fermi 2 operating cycle. The l
proposed TS changes do not physically impact the plant nor do they impact any design or functional requirements of the associated systems. That is, the proposed TS changes do not significantly degrade the performance or increase the challenges of any safety systems assumed to function in the accident analysis. The proposed
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TS changes affect only the frequency of the surveillance requirements and do not impact the TS surveillance requirements themselves. In addition, the proposed TS changes do not introduce any new accident initiators since no accidents previously '
i evaluated have as their initiators anything related to the change in the frequency of i
surveillance testing. Also, the proposed TS changes do not significantly affect the availability of equipment or systems required to mitigate the consequences of an
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accident because of other, more frequent testing or the availability of redundant systems or equipment. Furthermore, a historical review of surveillance test results supports the abave conclusions. Therefore, the proposed TS changes do not significantly increase the probability or consequences of an accident previously evaluated.
2.
The oronosed TS chances do not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed TS changes involve a one-time only change in the surveillance testing intervals to facilitate the one-time only change in the Fermi 2 operating cycle. The proposed TS changes do not introduce any failure mechanisms of a different type than those previously evaluated since there are no physical changes being made to the facility. In addition, the surveillance test requirements themselves will remain unchanged. Therefore, the proposed TS changes do not create the possibility of a new'or different kind of accident from any previously evaluated.
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NRC-98-0040 Page 3 3.
The proposed TS chances do not involve a significant reduction in a marnin of safety.
I Although the proposed TS changes will result in an increase in the interval between some surveillance tests, the impact, if any, on system availability is small based on other, more frequent testing or redundant systems or equipment, and there is no L
evidence of any time dependent failures that would impact the availability of the -
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systems. Therefore, the assumptions in the licensing basis are not impacted, and the proposed TS changes do not significantly reduce a margin of safety.
Based on the above, Detroit Edison has determined that the proposed amendment does not involve a significant hazards consideration.
ENVIRONMENTAL IMPACT Detroit Edison has reviewed the proposed Technical Specification changes against the criteria of 10 CFR 51.22 for environmental considerations. The proposed change does not involve a significant hazards consideration, nor significantly change the types or significantly increase the amounts of effluents that may be released offsite. In addition, the proposed changes will reduce occupational radiation exposure and so do not involve a significant increase in individual or cumulative occupational radiation exposures. Based on the foregoing, Detroit Edison concludes that the proposed Technical Specifications do meet the criteria given in 10 CFR 51.22(c)(9) for a categorical exclusion from the requirements for an Environmental Impact Statement.
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j l-Based on the evaluation above: 1)'there is reasonable assurance that the health 'and safety of the public will not be endangered by operation in the proposed manner, and 2) such l
activities will be conducted in compliance with the Commission's regulations and l
proposed amendments will not be inimical to the common defense and security or to the l'
health and safety of the public.
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