ML20076G491
| ML20076G491 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 10/11/1994 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20076G489 | List: |
| References | |
| NUDOCS 9410190362 | |
| Download: ML20076G491 (14) | |
Text
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t UNITED STATES NUCLEAR REGULATORY COMMISSION f
WASHINGTON. D.C. 20565.0001
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT N0.171 TO FACILITY OPERATING LICENSE NO. DPR-16 GPU NUCLEAR CORPORATION ALQ JERSEY CENTRAL POWER & LIGHT COMPANY OYSTER CREEK NUCLEAR GENERATING STATION DOCKET NO. 50-219
1.0 INTRODUCTION
The BWR Owners Group (BWR0G), of which GPU Nuclear Corporation (GPUN/the licensee) is a member, sponsored studies by General Electric (GE) to apply probabilistic analytical methods in order to justify an increase in surveillance test intervals (STIs) and allowable out-of-service times (A0Ts) for various BWR instrumentation. All proposed STI and A0T changes in accordance with these studies rr.sulted in a series of GE Licensing Topical Reports (LTRs) which have been previously reviewed and approved by the NRC staff. Also, A0Ts are clarified in accordance with the most recently approved BWR Owners Group letters which were used in the development of NUREG-1433
" Standard Technical Specification, General Elo.tric Plants, BWR/4." The Technical Specification (TS) changes will permit specified Channel Tests to be conducted quarterly rather than weekly or monthly. However, the approval was conditional based on a list of plant-specific conditions that the licensees should follow.
By letter dated May 12, 1994, as supplemented September 2, 1994, the licensee proposed to revise the TS requirements regarding STIs (Table 4.1.1) and A0Ts (Table 3.1.1) for Reactor Protection System (RPS), Emergency Core Cooling System (ECCS) Actuation, Isolation Actuation and Control Rod Block Instrumentation in accordance with the GE LTRs.
The proposed A0T and STI changes are based on the most recently approved BWR Owners Group letters which were used in establishing A0Ts and STIs for the new Standard Technical Specifications, (NUREG-1433).
Technical changes regarding Channel Calibration requirements for Average Power Range Monitor (APRM) Scram, High Drywell Pressure (for Core Cooling) and Turbine Trip Scram instrumentation are also proposed.
Editorial changes are proposed to correct and clarify TS.
The licensee also provided information regarding required plant-specific evaluation contained in reports MDE-98-0485, " Technical Specification Improvement Analysis for the Reactor Protection System for Oyster Creek Nuclear Generating Station" dated July 1985 and RE-004, " Technical Specification Improvement Analysis for Emergency Core Cooling System Actuation 9410190362 941011 PDR ADOCK 05000219 P
. Instrumentation for Oyster Creek Nuclear Generating Station" dated January 1987. The September 2,1994, letter provided clarifying information that did not change the initial proposed no significant hazards consideration determination.
2.0 DISCVSSION AND EVALUATION The proposed changes reflect Standard Technical Specification (STS) revisions contained in the LTRs which, based upon reliability analyses, support t
increases in STIs and A0Ts for surveillance and repair.
These changes are beneficial in reducing:
(i) potential unnecessary plant scrams, (ii) excessive equipment test cycles, and (iii) the diversion of personnel and resources for unnecessary testing.
The NRC staff has reviewed and approved these LTRs in Safety Evaluation Reports (SERs).
Subsequently, all the LTRs were issued with the corresponding SERs included.
The Oyster Creek Technical Specification requiremerts regarding STIs (Table 4.1.1) and A0Ts (Table 3.1.1) are to be revised for Reactor Protection System (RPS), Emergency Core Cooling System (ECCS) Actuation, Isolation Actuation and Control Rod Block Instrumentation in accordance with the following GE LTRs:
(1)
NEDC-30851P-A " Technical Specification Improvement Analyses for BWR Reactor Protection System," dated March 1988 (2)
NEDC-30936P-A (Parts 1 and 2) "BWR Owners Group Technical Specification Impr eement Methodology (With Demonstration for BWR ECCS Actuation Instrumenution)," dated December 1988 (3)
NE0C-30851P-A (Supplement 2) " Technical Specification Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation," dated March 1989 (4)
NEDC-31677P.A " Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation," dated Jt.ly 1990 l
(5)
NEDC-30851P-A (Supplement 1) " Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation," dated October 1988 (6)
GENE-770-06-1-A " Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," dated December 1992.
2.1 A0T AND STI CHANGES The A0T and STI changes to Oyster Creek TS grouped under the applicable topical reports are as follows:
. NEDC-308SlP-A dated March 1988/MDE-98-0485. Rev. I dated July 1985 (1)
Table 3.1.1, Section A.2-A.12 and Table Notes - Note nn is added which provides allowable out-of-service times for repair for the specified scram parameters. Note nn is clarified in accordance with BWROG letter 92102 from C. L. Tully (GE) to B. K. Grimes (NRC), "BWR Owners Group (BWROG) Topical Reports on Technical Specification Improvement Analysis for BWR Reactor Protection Systems - Use for Relay and Solid State Plants (NEDC-30884 and NEDC-30851P)," dated November 4,1992.
Note nn will apply to the following scram parameters:
a.
High-Reactor Pressure - Parameter 2 b.
High Drywell Pressure - Parameter 3 c.
Low Reactor Water Level - Parameter 4 d.
High Water Level in Scram Discharge Volume - Parameter 5.a & 5.b e.
Low Condenser Vacuum - Parameter 6 f.
High Radiation in Main Steamline Tunnel - Parameter 7 g.
Average Power Range Monitor - Parameter 8 h.
Intermediate Range Monitor (IRM) - Parameter 9 1.
Main Steamline Isolation Valve Closure - Parameter 10 j.
Turbine Trip Scram - Parameter 11 k.
Generator Load Rejection Scram - Parameter 12 (2)
Table 3.1.1, Note c - Note c is revised to allow any one APRM to be removed from service for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for surveillance without tripping its trip system.
(3)
Table 4.1.1, Instrument Channel Nos. 1 (Scram Function), 2, 3, 5.b, 11 (APRM Scram Trips) and 13.a - The Channel Test interval is revised to quarterly from weekly or monthly for the scram instrumentation identified below:
a.
High Reactor Pressure - Instrument Channel I b.
High Drywell Pressure (Scram) - Instrument Channel 2 c.
Low Reactor Water Level - Instrument Channel 3 d.
High Water Level in Scram Discharge Volume - Instrument Channel 5.b (analog) e.
APRM Scram Trips - Instrument Channel 11 f.
High Radiation in Main Steamline - Instrument Channel 13.a (4)
Table 4.1.1, Instrument Channel No. 30 and NOTE 1 - This is a new requirement added to ensure that the automatic scram contactors are tested on a weekly basis.
The test of the automatic scram contactors using the subchannel test switches does not have to be performed at each weekly interval if the automatic scram contactors are tested by other means, i.e., by performance of a different required Channel Calibration as discussed below in Item (12) and replaced with the note concerning the weekly test of the automatic scram contactors.
2
. NEDC-30936P-A. (Parts 1 and 21 dated December 1988/RE-004 dated January 1987 (5)
Table 3.1.1, Sections C.1-2, D.1-3, G.1-3 and Table Notes - Note pp is added providing an allowable out-of-service time for repair for the specified parameters as identified below:
a.
ISOLATION CONDENSER INITIATION (Section C)
- 1) High Reactor Pressure - Parameter 1
- 2) Low-Low Reactor Water Level - Parameter 2 b.
CORE SPRAY INITIATION (Section D)
- 1) Low-Low Reactor Water Level - Parameter 1
- 2) High Drywell Pressure - Parameter 2 3)
Low Reactor Pressure (valve permissive) - Parameter 3 c.
AUTOMATIC DEPRESSURIZATION (Section G)
- 1) High Drywell Pressure - Parameter 1
- 2) Low-Low-low Reactor Water Level - Parameter 2
- 3) AC Voltage - Parameter 3 (6)
Table 4.1.1, Instrument Channel Nos.1 (Isolation Condenser Actuation Function), 4 (Isolation Condenser Actuation and Core Spray Actuation Functions), 6 and 9 (Core Spray Actuation and Automatic Depressurization Actuation Functions) - The Channel Test interval is revised to quarterly from monthly for the ECCS Actuation instrumentation indicated below:
a.
High Reactor Pressure - Instrument Channel I b.
Low-Low Water Level - Instrument Channel 4 c.
Low-Low-Low Water Level - Instrument Channel 6 d.
High Drywell Pressure (Core Cooling) - Instrument Channel 9 NEDC-30851P-A. Sucolement 2 dated March 1989/NEDC-31677P-A dated July 1990 (7)
Table 3.1.1, Sections B.1-6, F.1-2, H.1-2 and L and Table Notes - Note oo is added which provides allowable out-of-service times for repair for the specified isolation actuation parameters.
Note oo is clarified in accordance with GE letter 0G90-579-32A from W. P. Sullivan and J. F.
Klapproth (GE) to M. L. Wohl (NRC), " Implementation Enhancements to 1
Technical Specification Changes Given in Isolation Actuation Instrumentation Analysis," dated June 25, 1990. Note oo will apply to the following isolation actuation instrumentation:
I a.
REACTOR ISOLATION (Section B) 1)
Low-Low Reactor Water Level - Parameter 1
- 2) High Flow in Main Steamline A - Parameter 2
- 3) High Flow in Main Steamline B - Parameter 3
- 4) High Temperature in Main Steamline Tunnel - Parameter 4 5)
Low Pressure in Main Steamline - Parameter 5
- 6) High Radiation in Main Steamline Tunnel - Parameter 6
- . a.
, b.
PRIMARY CONTAINMENT ISOLATION (Section F) 1). High Drywell Pressure - Parameter 1.
a 2)
Low-Low Reactor Water Level - Parameter 2 c.
ISOLATION CONDENSER ISOLATION-(Section H) 1)
High: Flow Steam Line - Parameter 1
'2) 'High Flow Condensate Line - _ Parameter 2 I
d.
CONDENSER VACVUM PUMP ISOLATION (Section L) 1)
High Radiation in Main Steamline Tunnel - Parameter 1 (8)
Table 4.1.1, Instrument Channel Nos. 2 (Primary Containment Isolation Function Common to Scram), 4 (Reactor Isolation and Primary Containment Isolation Functions Common to ECCS), 7, 8,'13a (Reactor Isolation and Condenser Vacuum Pump Isolation Functions Common to Scram), 14 and 15 -
The Channel Test interval is recised to quarterly from weekly or monthly for the following isolation actuation instrumentation:
a.
High Drywell Pressure (Scram) - Instrument Channel 2
[
b.
-Low-Low Water Level - Instrument Channel 4 c.
High Flow in Main Steamline - Instrument Channel 7 1'
d.
Low Pressure in Main.Steamline - Instrument Channel 8 e.
High Radiation in Main Steamline - Instrument Channel 13.a f.
High Radiation in Reactor Building - Instrument Channel 14 g.
High Radiation on Air Ejector Off-Gas - Instrument Channel 15 N10C-30851P-A. Sucolement I dated October 1988/ GENE-770-C6-1-A dated December 1992 t
(9)
Table 4.1.1, Instrument Channel No. 12 - The Channel Test frequency is revised to quarterly from monthly for the Control Rod Block instrumentation below:
a.
APRM Rod Blocks - Instrument Channel 12 TECHNICAL CHANGES (GLOBAL) j (10) Sections 3.1 and 4.1, Bases - The bases for the two TS Instrumentation sections are revised to-reflect the GE LTRs as the basis for A0T and STI 1
changes and will be included as references.
(11)
Table 3.1.1, Global Note * - This note, which appears at the end of. the table and just before the indicated Notes section, describes the Action Required column and provides an allowed out-of-service time. for performing surveillance. The note is revised to allow 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> instead of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance of instrument channels provided at least one operable channel in the same trip system is monitoring the parameter.
The 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> allowance is supported by the LTRs for all instrumentation in Table 3.1.1 except the Diesel Generator Load Sequence Timers and Loss of Power instruments.
Note kk was established for this instrumentation and retains the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> allowance.
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.g i (12) Table 4.1.1, NOTE 1 and Figure 4.1.1 - Current NOTE 1 and Figure 4.1.1 are deleted since surveillance intervals will be based on the reliability analyses contained in the GE LTRs and not on the methodology described in the current TS bases which uses Figure 4.1.1 which allows surveillance Channel Test intervals to be adjusted between monthly and quarterly.
The licensee has not used this method of adjusting Channel Test intervals.
2.2 CHANNEL CALIBRATION INTERVAL CHANGE The channel calibration interval changes proposed to the Oyster Creek TS are as follows:
(1)
Table 4.1.1, Instrument Channel 11.(APRM Scram Trips) - The Channel Calibration interval for the APRM Scram Trips is currently weekly. This interval is the same as the current Channel Test interval. The change of the Channel Test interval to quarterly is proposed and supported by the LTRs. Accordingly, an extension of the Channel Calibration Test interval to quarterly is appropriate. The licensee has demonstrated that drift data of the affected instrumentation remained within the existing allowance in the instrument setpoint calculation when l
considered over the extended period.
(2)
Table 4.1.1, Instrument Channels 9 (High Drywell Pressure for Core Cooling) and 23 (Turbine Trip Scram) - A quarterly Channel Calibration interval is added for instrument channels 9 and 23.
Currently, a Channel Calibration interval is not specified for the High Drywell Pressure instruments.
Since they are calibrated on a quarterly interval, it is appropriate to include this surveillance requirement in the TS. Also, the Channel Calibration interval is' not specified for instrument channel 23.
Since this trip parameter senses turbine stop valve position via limit switches and its switch adjustment methods are similar to Main Steamline Isolation Valve Scram instrumentation (Instrument Channel 10), it is appropriate to include this surveillance requirement in the TS.
2.3 EDITORIAL CHANGES The editorial changes proposed to the Oyster Creek TS are as follows:
(1)
Capitalization of Definitions - In Sections 3.1 and 4.1, the definitions, where they appear in specifications, bases,-tables and table notes, are capitalized to highlight the fact they are terms with specific meanings.
This is consistent with STS convention.
The definitions are contained in TS Section 1.0.
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. (2)
Table 3.1.1 Heading - The table column which currently reads Min. No. of Instrument Channels Per Operable Trip Systems is incorrect in that the word Systems should not be plural. A separate column specifies the required number of Trip Systems.
The subject column specifies the number of instrument channels for each Trip System and, therefore, the word System in this column is properly singular.
(3)
Table 3.1.1, Sections A.7, B.6 and L.1 - Minor grammatical changes are made to ensure that this instrumentation (High Radiation in Main Steamline Tunnel) is consistently identified where it appears in its functions of scram (Section A.7) and isolation (Sections B.6 and L.1).
Since this instrumentation is common to both scram and isolation functions, it needs to be consistently displayed to preclude confusion when new Note oo to Table 3.1.1 is applied.
(4)
Table 3.1.1, Section C - The word " plant" is deleted from the Action Required since it is superfluous.
This allows the Action Required to refer directly to the TS definition for PLACE IN COLD SHUTDOWN CONDITION. Also, in order to clarify the function achieved, the heading is revised to Isolation Condenser " Initiation".
(5)
Table 3.1.1, Section C.2 - The word " level" is added to the C.2 description of the Isolation Condenser Initiation variable on Low-Low Reactor Water Level.
This is for clarification purposes and makes this description consistent with other reactor water level instrument descriptions in the table.
(6)
Table 3.1.1, Section D - The description of the function in this section is clarified to indicate that instrumentation under Section D initiates Core Spray.
Therefore, the function description is revised to read Core Spray Initiation.
(7)
Table 3.1.1, Section J.1 - The requirement for operability in the Startup mode was inadvertently omitted in License Amendment 72.
The table is revised to restore that omission.
(8)
Table 3.1.1, Section J.4, Notes and Associated Footnote - Note gg is deleted since it concerns a 1985 licensing condition which is no longer in effect.
This note expired at the end of the Cycle mid-10 outage.
Oyster Creek is currently operating in Cycle 14.
(9)
Table 3.1.1, Note t - A typographical error in this note currently allows sensors to be "... operable or bypassed...".
The note should correctly read
... inoperable or bypassed..." since this note would be unnecessary to allow operability.
This clarifies that Core Spray instrumentation operability is not required if its associated and supported Core Spray System is inoperable.
(10) Table 3.1.1, Note y - All references to Note y in Table 3.1.1 were deleted by License Amendment 75.
This note is no longer applicable and is deleted.
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(11) Table 3.1.1, Sections M.1, M.2, M.3, N.a and b and Table Notes - Note kk is added to indicate that the allowable cut-of-service time for surveillance of the Diesel Generator Load Sequence Timers and Loss of l
Power instrumentation does not change and remains 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
The GE LTRs do not address this instrumentation.
In addition, Section M.3 is revised to clarify that the load sequence timer is associated with the Reactor Building Closed Cooling Water Pumps.
(12)
Section 4.1. Specification - A minor editorial change to Specification revises "...as per definitions..." to "...using the definitions...
This change does not alter the meaning or intent of the Specification and is purely grammatical.
(13) Table 4.1.1, NOTE 2 - This note refers to "...Section 2.3 Specifications (1) (a) and (2) (a)...
These specifications are not currently identified in this fashion.
The correct identification of these specifications is "... A.1 and A.2..." and NOTE 2 is revised accordingly.
This change is a correction and ensures consistency between NOTE 2 and Section 2.3.
(14)
Table 4.1.1, Instrument Channels 18, 20 and 25 - The surveillance interval is currently displayed as "l/20" which means once every 20 months.
This is revised to 1/20 mo to clarify that the interval is 20 months and to ensure consistency with the nomenclature used in Table 4.1.1.
(15) Table 4.1.1, Instrument Channels 23, 24 and 26 - The surveillance interval is currently indicated as "Every 3 months." Since the nomenclature used in Table 4.1.1 is 1/3 mo for this interval, the surveillance frequency description for the subject instrument channels is revised to conform with the nomenclature.
This change does not alter the interval and is editorial.
(16) Table 4.1.1, Instrument Channels 28a and 28b - The Channel Check interval for these instrument channels is specified as " daily." The nomenclature indicated in the table legend shows "l/d" as the description of this interval. To ensure consistency, the Channel Check interval for the subject instrument channels is revised to conform with the legend. This change does not alter the interval and is editorial.
(17) Table 4.1.1, NOTES - The table notes are moved from the first page of the table to a more appropriate location at the end of the table.
(18) Table 4.1.1, legend - The table legend establishes nomenclature used in Table 4.1.1 to specify surveillance intervals.
The following definitions are added to the legend: 1/mo = Once per month, 1/20 mo -
Once every 20 months, and 1/24 mo - Once every 24 months.
In addition, the table legend defining the once every 18 month interval is deleted since this interval no longer appears in Table 4.1.1.
. (19) Sections 3.1 and 4.1, Amendment Nos. on pages - As part of the pagination process for this change request, the previous license amendment numbers displayed on the bottom of each page (where applicable since not all pages have amendment numbers) have been revised to ensure that the numbers accurately reflect associated changes to the pages.
2.4 JUSTIFICATION FOR THE PROPOSED CHANGES GPUN has determined that the generic analyses performed by GE for the BWR Owners Group for revised A0Ts and STis for RPS, ECCS Actuation, Isolation Actuation and Rod Block instrumentation are applicable to the Oyster Creek Nuclear Generating Station.
GPUN has completed plant-specific evaluations required by the NRC SERs which approved the LTRs for use by individual facilities. As stated in the SERs, three issues must be addressed to apply the RPS LTR (NEDC-30851P-A) and two issues must be addressed to apply the other LTRs to an individual facility when specific TS are considered for revision.
These issues, the licensee's treatment and the staff's evaluation of them are as follows:
(1)
Confirm the applicability of the generic analyses to the specific facility a.
The generic study in NEDC-30851P-A provides a technical basis to modify the STIs and A0Ts of the RPS.
The generic study also provides additional analyses of different RPS configurations to support the application of the generic basis on a plant-specific basis.
A plant-specific evaluation for modifying the STIs and A0Ts of the RPS in the TS of Oyster Creek has been performed by GE and is contained in the plant-specific evaluation report MDE-98-0485, Technical Specification Improvement Analysis for the Reactor Protection System for Oyster Creek Nuclear Generating Station. The evaluation utilized the generic basis and additional analyses documented in LTR NEDC-30851P-A.
The results indicated that the RPS configuration for Oyster Creek has several differences compared to the RPS configuration in the generic evaluation.
The NRC staff has reviewed NEDC-30851P-A and MDE 0485 and verified that the generic analysis is applicable to Oyster Creek.
The differences between the RPS at Oyster Creek and the generic plant analyzed in NEDC-30851P-A are discussed in (3) below.
b.
GE Report NEDC-30936P-A provides an acceptable generic basis for supporting plant-specific TS changes that extend ECCS STIs and A0Ts for test and repair.
The plant-specific evaluation contained in GE Report RE-004, Technical Specification l
Improvement Analysis for the Emergency Core Cooling System Actuation Instrumentation for Oyster Creek Nuclear Generating Station, followed the procedures of NEDC-30936-A to identify and evaluate the differences between the Oyster Creek ECCS configuration and the ECCS configuration used in the generic
e,
, analysis.
The results of the staff review indicate that while the ECCS configuration for Oyster Creek has two minor discrepancies, the discrepancies and their impact do not affect the applicability of the TS changes developed by the generic ef forts of these LTRs. Therefore, the generic analysis in NEDC-30936P-A is applicable to Oyster Creek.
c.
LTR NEDC-30851P-A, Supplement 2, Appendix A identifies GPUN as a participant in the BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation evaluation.
The staff reviewed the licensee's evaluation and verified that the generic analysis is applicable to Oyster Creek.
d.
LTR NEDC-31677P-A, Appendix E identifies GPUN as a participant in the BWR Isolation Actuation Instrumentation evaluation.
The staff reviewed the licensee's evaluation and verified that the generic analysis is applicable to Oyster Creek.
e.
LTR NEDC-30851P-A, Supplement 1, Appendix B identifies GPUN as a participant in the BWR Control Rod Block Instrumentation evaluation.
The staff reviewed the licensee's evaluation and verified that the generic analysis is applicable to Oyster Creek.
f.
LTR GENE-770-06-1-A identifies the application of changes to STIs and A0Ts for Selected Instrumentation Technical Specifications to all BWR plants.
The staff reviewed the licensee's evaluation and verified that this LTR is applicable to Oyster Creek.
This LTR was found applicable only to the limited extent that it identified A0Ts for Control Rod Block Instrumentation which NEDC-30851P-A, Supplement I did not explicitly address.
(2)
Demonstrate, by use of current drift information provided by the equipment vendor or plant-specific data, that the drift characteristics for instrumentation used in RPS, ECCS, Isolation and Rod Block instrument channels in the plant are bounded by the assumptions used in the LTRs when the functional test interval is extended from weekly or monthly to quarterly.
The staff generic SER of May 27, 1987 on GE LTR NEDC-30844 and NEDC-30851P states the NRC's requirement for confirmation of instrument setpoint drift allowance.
By a letter to the BWR Owners Group from C.
l Rossi (NRC) dated April 27, 1988, the NRC requested licensees to confirm j
that the setpoint drift which could be expected under the extended STIs l
has been studied and either (i) has been shown to remain within the existing allowance in the RPS and ESFAS instrument setpoint calculation l
or (ii) that the allowance and setpoint have been adjusted to account for the additional expected drift.
No additional information need to be provided for staff review.
However, records showing the actual setpoint calculation and supporting data should be retained onsite for possible future staff audit.
The licensee has demonstrated that drift data of i
1
. the affected instrumentation remained within the existing allowance in the RPS and ESFAS instrument setpoint calculation when considered over the extended period.
(3)
Confirm that the differences between the parts of the RPS that perform the trip functions in the plant and those of the base case plant were included in the plant-specific analysis done using the procedures of Appendix K of NEDC-30851P or present plant-specific analyses to demonstrate no appreciable change in RPS availability or public risk.
GE Report MDE-98-0485, Revision 1, " Technical Specification Improvement Analysis for the Reactor Protection System for Oyster Creek Nuclear Generating Station", provides a plant-specific evaluation to determine whether the generic study contained in LTR NEDC-30851P-A is applicable to Oyster Creek to review ST!s and add A0Ts in TS for the RPS instrumentation.
This report utilizes the procedures in Appendix K of NEDC-30851P-A to identify and evaluate the differences between the parts of the RPS that perform the trip functions at Oyster Creek and those analyzed in the generic study.
GPUN performed an evaluation of NEDC-30851P-A and MDE-98-0485 to verify that the generic analysis is applicable to Oyster Creek and concluded that it remains applicable with discrepancies addressed as follows:
a.
MDE-98-0485, Appendix A,Section II, Part B.1 - This section lists the RPS sensors and identifies type, total number and number per RPS channel. The number given for MSIV position is 4 total and 3/RPS channel.
The correct numbers are 8 total and 2/RPS channel.
Oyster Creek has separate limit switches which initiate scram at 10% closure for each MSIV.
The generic model contains four instrument channels per trip system as shown in NEDC-30851P-A, Table 7.4.
With two sensors per channel and two channels per trip system, the number of sensors per trip system for MSIV position at Oyster Creek is four.
b.
MDE-98-0485, Section 3, Item k; Appendix A,Section II, Part B.1 and Section 111, Part B.1 - The plant-specific report indicates bistable switches are used to monitor reactor pressure and reactor water level.
Subsequent to the preparation of the report, these sensors were replaced with analog trip units and transmitters. Therefore, there is no longer a difference between j
the generic model and Oyster Creek in this area.
I c.
MDE-98-0485, Appendix A,Section II, Parts C.3, 0.1 and F.1 and Section III, Part 0.1 - Oyster Creek uses GE Type CR205 automatic scram contactors while one GE Type CR205 and one GE Type CR305 are used for the manual scram contactors.
The generic model used GE Type CR105 for automatic and manual scram contactors.
There is no significant reliability effect because the common cause failure rate of GE Type CR105, CR205 and CR305 contactors are the same and the use of diverse contactor models tends to improve reliability for the manual scram function.
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. d.
MDE-98-0485, Section 3, Item m; Appendix A,Section II, Part G.2 and Section III, Part G.1 - Calibration frequency requirements for analog trip units were incorporated into the TS.
The-quarterly calibration interval for analog trip units and annual calibration interval for transmitters is within the range of the sensitivity study documented in NEDC-30851P-A.
e.
MDE-98-0485, Section 3, Item m; Appendix A,Section II, Parts G.4 and G.6 - Although Oyster Creek TS do not provide specific A0Ts for the inoperable instrument channels or trip systems, the bases in TS 3.1 indicate that prompt action is taken to trip the channel or trip system to compensate for the inoperable condition when it involves one trip system.
This is done immediately.
However, when both trip systems are involved the Action Required in TS Table 3.1.1 is initiated immediately since tripping both scram systems will initiate a reactor trip which is undesirable.
f.
MDE-98-0485, Section 3, Item 3; Appendix A,Section II, Part G.7 and Section III, Part G.1 - TS currently allow a channel to be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillance without placing the trip system in the tripped condition.
This has been changed from a 1-hour allowance and is the nominal original A0T for surveillance considered in NEDC-30851P-A.
g.
MDE-98-0485, Appendix A,Section II, Part H.1 - Flux sensors, radiation sensors, and analog sensors (high reactor pressure, low reactor water level, low-low reactor water level, and high water level) are generally rot included in the Channel Tests at Oyster Creek as instrument loop test sWtches_ provide the means for logic testing. These differences are consistent with the TS definition for Channel Test and have no effect with respect to the generic model since they are within the range of the sensitivity study performed in NEDC-30851P-A.
h.
MDE-98-0485, Appendix A,Section II, Parts H.2 and H.3 - When an individual sensor channel.is in repair, the logic channel is tripped, and if the individual sensor channel is 'in test, the sensor is temporarily inoperable but the logic channel is not necessarily tripped.
These conditions are permitted by current TS and have no effect when comparing Oyster Creek to the generic model since they are within the range of the sensitivity study performed in NEDC-30851P-A.
i.
Oyster Creek utilizes a High Recirculation Flow Scram which is not identified in MDE-98-0485.
Two recirculation flow converters, one for each RPS trip system initiates a scram when recirculation flow exceeds the trip setpoint.
This parameter does not serve as a scram sensor for any of the more severe initiating events as defined in NEDC-30851P-A and consequently the addition of another scram initiation signal will have no
a>
. l significant impact on RPS failure frequency.
Elimination of this scram parameter would have no effect on the Oyster Creek relationship to the generic model.
J.
MDE-98-0485, Section 3, Item p; Appendix A,Section II, Part H.4 and Section III, Part G.2 - The number of scram contactor actuations currently experienced during Channel Tests at Oyster Creek differs from those assumed in the generic model and identified in the plant-specific report.
The differences are described as follows:
1)
APRM Channel Test results in 6 actuations per scram contactor in each automatic trip logic channel 2)
MSIV Closure Channel Test causes 2 actuations per scram contactor in each automatic trip logic channel 3)
Turbine Control Valve fast closure Channel Test consists of 2 actuations per scram contactor in each automatic trip logic channel since the s40% power Turbine trip Scram Bypass switches are also tested during this surveillance 4)
The High Recirculation Flow Scram test is performed quarterly and consists of 2 actuations per scram contactor in each automatic trip logic channel 5)
IRM Front Panel test is performed weekly and initiates 2 actuations per scram contactor in each automatic trip logic channel whenever the reactor is not in the RUN MODE.
The above differences have no significant effect since they remain within the range of the sensitivity study performed in NEDC-30851P-A.
The total number of automatic scram contactor actuations is estimated to be approximately 406 actuations per contactor per year.
This exceeds the estimate in NEDC-30851P-A of 272165 actuations per year but has no significant impact on RPS reliability.
k.
The backup scram valves are de-energized to trip.
There is no significant effect on RPS reliability caused by this because the operation of the backup scram valves is controlled by the scram contactors.
The differences between the RPS at Oyster Creek and the generic model are bounded by the analysis contained in the RPS LTR (NEDC-30851P-A).
As discussed in the above, the staff concludes that the generic analysis remains applicable to Oyster Creek.
.a.,,
I 2.4
SUMMARY
The proposed changes extend STIs and A0Ts for instrumentation and have been justified using probabilistic analytical methods.
The affected instrumentation is' associated with the RPS, ECCS, Isolation Actuation and Control Rod Block Instrumentation. The changes have been the subject of generic Licensing Topical Reports which the NRC has reviewed and approved.
GPU Nuclear has addressed the implementation of the generic Technical
. Specification changes identified in the LTRs on a plant-specific basis.
The i
staff has reviewed the LTRs and the plant-specific reports and concludes that' l
the generic analyses are applicable to Oyster Creek. The changes also include editorial changes which are corrections to ensure consistent use of nomenclature, correction of typographical errors, reformatting of the instrumentation tables, and deletion of a note which is no longer. applicable.
The licensee performed the required plant-specific analysis and justified the application of generic analysis to the Oyster Creek plant-specific design.
The information for setpoint drift supports the conclusion that instrument drift is not a concern in extending the functional test interval. from monthly to quarterly.
Therefore, the staff has'found the proposed changes to the Oyster Creek Technical Specifications acceptable.
3.0 STATE CONSULTATION
I In accordance with the Commission's regulations, the New Jersey State official was notified of the proposed issuance of the amendment. The State official had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a i
facility component located within the restricted area as defined in 10 CFR l
Part 20 and changes surveillance requirements.
The NRC staff has determined that the amendment involves no significant increase in the amounts,.and no significant change in the types, of any effluents.that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued'a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (59 FR 32228). Accordingly, the amendment meets the eligibility criteria for i
categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR. 51.22(b) no environmental impact statement or environmental assessment need be i
prepared in connection with the issuance of the amendment.
l
5.0 CONCLUSION
+
The Commission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
S. Rhow Date: October 11, 1994
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