ML13331B034
| ML13331B034 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 04/15/1988 |
| From: | Baskin K Southern California Edison Co |
| To: | |
| Shared Package | |
| ML13316B885 | List: |
| References | |
| NUDOCS 8804220202 | |
| Download: ML13331B034 (13) | |
Text
S S
BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION Application of SOUTHERN CALIFORNIA EDISON
)
COMPANY and SAN DIEGO GAS & ELECTRIC COMPANY )
for a Class 104(b) License to Acquire,
)
DOCKET NO. 50-206 Possess, and Use a Utilization Facility as
)
Part of Unit No. 1 of the San Onofre Nuclear )
Amendment No. 150 Generating Station
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SOUTHERN CALIFORNIA EDISON COMPANY and SAN DIEGO GAS & ELECTRIC COMPANY, pursuant to 10 CFR 50.90, hereby submit Amendment Application No. 150.
This amendment consists of Proposed Change No. 180 to Provisional Operating License No. DPR-13. Proposed Change No. 180 modifies the Technical Specifications incorporated in Provisional Operating License No. DPR-13 as Appendix A.
Proposed Change No. 180 is a request to revise Appendix A Technical Specification Sections 2, 3 and 4 Technical Specifications to be consistent with a replacement upgraded NIS for San Onofre Unit 1, also described therein. The proposed Technical Specifications provide Limiting Conditions for Operation and Surveillance requirements for the upgraded NIS.
In the event of conflict, the information in Amendment Application No. 150 supersedes the information previously submitted.
8804220202 880415 PDR ADOCK 05000206 P
0
-2 Based on the significant hazards analysis provided in the Description of Proposed Change and Significant Hazards Analysis of Proposed Change No. 180, it is concluded that (1) the proposed change does not involve a significant hazards consideration as defined in 10 CFR 50.92, and (2) there is reasonable assurance that the health and safety of the public will not be endangered by the proposed change.
Pursuant to 10 CFR 170.12, the fee of $150 is herewith remitted.
-3 Subscribed on this /_
day of
, 1988.
Respectfully submitted, SOUTHERN CALIFORNIA EDISON COMPANY By: -P/AX )
Kenneth P. Baskin Vice President Subscribed and sworno before me this
/4 day of
//__
Nota Publi'c in and for the County of Los ngeles, State of California My Commission Expires:
lv3/f OFFICIAL SEAL AGNES CPABTREE Notary Public -California Charl es R. Kocher uLOSANGELESCOUNTY James A. Beoletto My Comm. Exp. Sep. 14,1990 Attorneys for Southern California Edison Company By:
Ja es\\A. Be tto
-4 Subscribed on this
/
day of _
1988.
Respectfully submitted, SAN DIEGO GAS & ELECTRIC COMPANY By :
G ar D//otton SenioV ice President Subscribed and sworn to before me this
.;Z day of
/_
_f_
OFFICIAL SEAL STEPHANIE E. HITT NOTARY PUBUIC CAUFORN PRINCiPAL Notary PubS'ic in and for the County of San Diego, State of California My Commission Expires:
I OJO 22I David R. Pigott Samuel B. Casey Orrick, Herrington & Sutcliffe Attorneys for San Diego Gas & Electric Company By:
David R. Pigott
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of SOUTHERN
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CALIFORNIA EDISON COMPANY
)
and SAN DIEGO GAS & ELECTRIC
)
Docket No. 50-206 COMPANY (San Onofre Nuclear
)
Generating Station Unit No. 1
)
CERTIFICATE OF SERVICE I hereby certify that a copy of Amendment Application No. 150 was served on the following by deposit in the United States Mail, postage prepaid, on the 15th day of April, 1988.
Henry J. McGurren, Esq.
Staff Counsel U.S. Nuclear Regulatory Commission Washington, D.C. 20545 David R. Pigott, Esq.
Samuel B. Casey, Esq.
Orrick, Herrington & Sutcliffe 600 Montgomery Street San Francisco, California 94111 L. G. Hinkleman Bechtel Power Corporation P.O. Box 60860, Terminal Annex Los Angeles, California 90060 Michael L. Mellor, Esq.
Thelen, Marrin, Johnson & Bridges Two Embarcadero Center San Francisco, California 94111 Huey Johnson Secretary for Resources State of California 1416 Ninth Street Sacramento, California 95814 Janice E. Kerr, General Counsel California Public Utilities Commission 5066 State Building San Francisco, California 94102
-2 C. J. Craig Manager U. S. Nuclear Projects I ESSD Westinghouse Electric Corporation Post Office Box 355 Pittsburgh, Pennsylvania 15230 A. I. Gaede 23222 Cheswald Drive Laguna Niguel, California 92677 Frederick E. John, Executive Director California Public Utilities Commission 5050 State Building San Francisco, California 94102 Docketing and Service Section Office of the Secretary U.S. Nuclear Regulatory Commission Washington, D.C. 20555
' es A. Beolet o
DESCRIPTION AND SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS OF PROPOSED CHANGE NO. 180 TO THE TECHNICAL SPECIFICATIONS PROVISIONAL OPERATING LICENSE NO. DPR-13 This is a request to revise Sections 1.0, "DEFINITIONS;" 2.1, "REACTOR CORE Limiting Combination of Power, Pressure and Temperature;" 3.5.1, "REACTOR TRIP SYSTEM INSTRUMENTATION;" 3.5.6, "ACCIDENT MONITORING INSTRUMENTATION;"
3.11, "CONTINUOUS POWER DISTRIBUTION MONITORING;" 4.1.1, "OPERATIONAL SAFETY ITEMS;" 4.1.5, "ACCIDENT MONITORING INSTRUMENTATION" and 4.4, "EMERGENCY POWER SYSTEM PERIODIC TESTING" of the Appendix A Technical Specifications for San Onofre Nuclear Generating Station, Unit 1 (SONGS 1).
DESCRIPTION OF CHANGES Technical Specification 1.0 describes the definitions of terms used throughout the specifications. Proposed Change No. 180 adds a definition for ACTUATION LOGIC TEST, adds to the CHANNEL TEST definition and reorders the specification to list the terms in alphabetical order. These revisions are consistent, to the extent possible, with NUREG-0452, "Standard Technical Specifications (STS) for Westinghouse Pressurized Water Reactors." The ACTUATION LOGIC TEST definition is exactly consistent with the STS and is necessary due to the addition of the NIS coincidentor logic to Specification 3.5.1.
The expanded definition to CHANNEL TEST is for consistency with the STS, but the STS title "ANALOG CHANNEL OPERATIONAL TEST" is not adopted due to the current extensive use of the term "CHANNEL TEST" in other specifications not currently being revised. The reordering of Section 1.0 necessitates an administrative change to Table 4.1.2 to renumber the reference to the definition for E.
Technical Specification 2.1 describes the limiting core operating conditions and the maximum safety settings. Among the settings is the limit for a reactor trip on nuclear overpower, as detected by the Nuclear Instrumentation System (NIS).
Proposed Change No. 180 would include a value for the low setting and revise the footnote regarding an adjustment of the nuclear overpower trip high setting setpoint based upon asymmetric power distributions that may be seen during power operation. The inclusion of a value for the nuclear overpower low setting is consistent with the STS and the SRR. This revision does not represent a change from the past setpoint for this NIS range, but recognizes a need to include this value in Table 2.1.
The revision to the nuclear overpower high setting footnote consists of a redefinition of the limit of asymmetric power distribution from 10% to 5%. Asymmetry is defined in plant procedures as the difference between the highest and lowest reading power range percent meters.
Technical Specification 3.5.1 describes the Limiting Conditions for Operation (LCOs) for the reactor trip system instrumentation. This instrumentation is necessary to initiate appropriate plant response to prevent exceeding plant safety limits.
Proposed Change No. 180 proposes to revise 3.5.1 to be consistent with the upgraded NIS being installed at SONGS 1. These revisions are consistent, to the extent possible, with the STS. The addition of the Power Range Dropped Rod Rod Stop and NIS Coincidentor Logic operability, and associated action statements are consistent with the STS and their safety function, as described in the Attachment 3 Safety Review Report (SRR) and NIS Installation Design, (hereafter collectively referred to as
-2 the SRR).
Therefore, the proposed Dropped Rod Rod Stop and NIS Coincidentor Logic specifications are acceptable. The revision in the Table 3.5.1-1 Table Notation to the P-7 definition is for consistency with the SRR. The revision of Action 2 to now require that channels be placed in a tripped condition within I hour and revise the bypass option is proposed because the new NIS will enable SONGS 1 to perform this action in a time frame consistent with that assumed in the STS, since this action statement is now narrowly limited to the nuclear overpower trip, and there is no overpower trip bypass capability, only a "test" capability. There is also an addition to Action 3 to allow bypass of the intermediate range channel for surveillance testing.
Technical Specification 3.5.6 requires that certain accident monitoring instrumentation be operable and requires appropriate action in the event of channel inoperability. In Proposed Change No. 180, Table 3.5.6-1 is revised to include the two new wide range neutron flux level indications that will be included in the NIS upgrade. These changes are consistent with the STS.
Technical Specification 3.11 describes the incore axial offset (IAO) limits and actions to be taken in the event these limits are exceeded or the measurement system is inoperable. Proposed Change No. 180 proposes to revise Specification 3.11 to be consistent with the NIS upgrade described in the attached SRR. The NIS will now be used to measure IAO, instead of the previously dedicated IAO system, with the NIS as a backup. Therefore, the proposed revisions will revise the component operability requirements from Specification 3.11 and require that in the event of unavailability of IAO measurement capability reactor power will be reduced. The NIS operability is required in Specification 3.5.1 and the requirement for incore-excore correlations is required in Specification 3.10. Additionally, consistent with current practice, an additional backup method using operable NIS channels and hand calculations shall be employed in the event that the selected NIS channels go out of "correlation" as described in Specification 3.10 or the excore axial offset signal processing equipment fails and these components can no longer be used to determine IAO.
The title of Specification 3.11 is revised to be more consistent with its content.
Specification 4.1 describes the surveillance requirements for various SONGS 1 systems and instrumentation, among which the Reactor Trip System Surveillance is included. The existing specification is presented in a tabular format that is not consistent with existing Table 3.5.1-1 or the STS. Proposed Change No. 180 proposes to revise Specification 4.1 to be consistent with the STS, with regards to the Reactor Trip System Instrumentation specified in Specification 3.5.1. A new Table 4.1.1 is proposed with STS type of frequencies and terminology. The remaining items are moved, in their current format, to a new Table 4.1.3, except that a new item for pressurizer pressure input to safety injection actuation is added, as the previous pressurizer.
pressure surveillance of Table 4.1.1 included a test of this function. The surveillances required for this functional element are consistent with the requirements for this element in Specification 4.1.4. Additionally, the existing steam generator level functional element of Table 4.1.1 is limited, by the new Table 4.1.1, to steam flow/feed flow mismatch, as steam generator level monitoring surveillances are adequately covered by Specifications 4.1.5, "Accident Monitoring Instrumentation," and 4.1.8, "Auxiliary Feedwater
-3 Instrumentation."
For all functional units in proposed Table 4.1.1, STS surveillance intervals and limits are specified where appropriate. In many cases the new Table 4.1.1 requires surveillance where none was previously included in the technical specifications. The surveillances and frequencies are a codification into the technical specifications of existing plant practices and are consistent with the STS. The Modes for which surveillance is required is not included, since the provisions of Specification 4.0.1 are an adequate complement to 4.1.1. There are no requirements similar to the STS for an incore-excore calibration, since the current provisions of SONGS 1 specification 3.10, and as proposed in Proposed Change No. 171 to the technical specifications continue to be adequate. The footnote that applies to the startup test frequency is revised to be 31 days. This revision is consistent with the NRC's review of Westinghouse Owner's Group (WOG) work documented in NRC letter from Harold R. Denton to L. D. Butterfield dated July 24, 1985. Specifically, the NRC Staff stated:
based on insights gained during the review of HCAP-10271, the staff has concluded that the technical specification footnote which applies to the start-up test frequency for these channels can be modified. Currently, the footnote indicates that the test must be made within 7 days before start-up. The NRC has concluded that this footnote should be changed to require the test within 31 days before start-up. For a plant experiencing frequent shutdowns or difficulty starting-up, this change would make the frequency of testing during periods when these channels are required to be operable comparable to the testing frequency for the channels which must be operable at power."
The remaining surveillance specifications are consistent with the existing and/or STS surveillance intervals and types. In no case were the existing surveillance intervals relaxed.
Technical Specification 4.1.5 describes the surveillance intervals and types for the accident monitoring instrumentation specified to be operable by the LCOs of Specification 3.5.6. Proposed Change No. 180 revised Table 4.1.5-1 to be consistent with the revision to 3.5.6 and the STS for this type of equipment.
Technical Specification 4.4 describes the surveillance requirements for the emergency power system at SONGS 1. Specifically, part D.2.e describes a performance discharge test for station batteries. Due to DC loads added by the upgraded NIS, Proposed Change No. 180 proposes to revise Specification 4.4 to require that the performance discharge test be performed to an acceptance criterion of 85% of the manufacturers rating for 125 Volt Battery Bank No. 1.
EXISTING TECHNICAL SPECIFICATIONS See Attachment 1 PROPOSED TECHNICAL SPECIFICATIONS See Attachment 2
-4 SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS As required by 10 CFR 50.91(a)(1), this analysis is provided to demonstrate that the proposed license amendment to implement technical specifications associated with a Nuclear Instrumentation System replacement at the San Onofre Nuclear Generating Station Unit 1 (SONGS 1) represents a no significant hazards consideration. In accordance with the three factor test of 10 CFR 50.92(c), implementation of the proposed license amendment was analyzed using the following standards and found not to:
- 1) involve a significant increase in the probability or consequences for an accident previously evaluated; or
- 2) create the possibility of a new or different kind of accident from any accident previously evaluated; or 3) involve a significant reduction in a margin of safety.
This change is proposed due to the implementation of an upgraded NIS at SONGS 1. The details and safety review of the NIS modification are provided as Attachments 3 and 4 to this change. Accordingly, as can be seen by an examination of the technical specification revisions proposed herein and in the SRR, the proposed revisions are consistent with the intent of the Technical Specifications to delineate operability and surveillance requirements for safety systems. A majority of the specifications are proposed because they are made possible because of the implementation of this modern NIS design. The new NIS allows for testing and operation consistent with the basis of the STS and other previously reviewed NRC positions.
Accordingly, the revisions to Technical Specifications 1.0, "Definitions;" the balance of 2.1, "Reactor Core-Limiting Cobination of Power, Pressure and Temperature;" the balance of 3.5.1, "Reactor Trip System Instrumentation;"
4.1.1, "Operational Safety Items;" and 4.1.5, "Accident Monitorning Instrumentation," are considered an upgrade of the manner in which SONGS 1 is operated and are consistent with the STS. The remaining changes to the asymmetric power limit, the power range-dropped rod rod stop, the incore axial offset monitoring limits and battery surveillance are proposed due to the inappropriateness of the existing specifications when the NIS upgrade is considered. Therefore, only these four changes will be further evaluated for significant hazards consideration.
The revision of the asymmetric power limit defined in Technical Specification 2.1 from 10% to 5% is necessary due to a reorientation of the power range neutron flux detectors of the new NIS. The relocation of one power range excore detector from thimble 4 to thimble 3 alters the radial separation of the power range detectors from 90 degrees between each so that the detector in thimble 3 is 45 degrees from the adjacent detector in one direction and 135 degrees from the adjacent detector in the opposite direction. This relocation potentially changes what the detectors will sense given an asymmetric power distribution and thus the resulting value for asymmetry.
As part of the review performed by Hestinghouse, documented in Section 5.0 of the Attachment 3 SRR, an evaluation of asymmetric power distribution has shown that with the relocated power range instruments, certain cases with an asymmetry of > 10%, which now require tech spec action, would show an asymmetry of < 10% but greater than 5%. Accordingly, the technical specification allowance of up to 10% asymmetry is reduced to allow up to 5%
0 0
-5 asymmetry. The asymmetry limit in the Basis is also changed to 5%. It is noted that this change really should only apply to the one area of the core for which there is less detector "coverage," but this specification will be conservatively applied to all asymmetric radial power distributions, regardless of their location.
This revision does not alter any previously reviewed accidents, create any new accidents or reduce any safety margins,-as it recognizes the reduction in capability of the upgraded NIS to detect specific asymmetric radial power distributions and adjusts downward the nuclear overpower trip in a manner consistent with the reduced capability.
The change to the LCO requirements of Technical Specification 3.5.1 and the surveillance requirements of Technical Specification 4.1 regarding the dropped rod protective function are consistent with the STS for systems of this type and with the design basis of this function, as presented in the SRR. The LCO and associated action statement will assure that the four required channels are operable or the plant is placed in a safe configuration for continued operation. The action specifies that with one less than the minimum channels operable, the inoperable channel will be placed in trip or power reduced to a level at which the turbine runback design feature is not required to mitigate a design basis two dropped rod event and the rod control system is placed in the manual mode. When the rod control system is in manual, control rod movement for the purposes of plant operation evolution is possible, but upon receipt of negative reactivity indications due to the presence of a single or double dropped rod scenario, automatic rod withdrawal will not occur and, accordingly, the conclusion of the design basis event, as described in the attached SRR, are assured. This action is not precisely consistent with any STS action statement, due to the plant-specific nature of the SONGS 1 design documented in Section 3.2.5.3 of the Attachment 3 SRR, but is consistent with the technical specification philosophy of maintaining safety systems operable or placing the plant in an acceptable alternate operating configuration.
The change to the requirements of Technical Specification 3.11 are necessary to specifically delineate the operability requirements for IAO measurement capability and the actions to be taken in the event of loss of..IAO measurement capability. The mode applicability of Mode 1 above 90% power is consistent with the existing requirements of part D of 3.11, and is not considered to be a change. The revision of parts C and D is necessary due to the lack of auxiliary channels for IAO determination and, accordingly, the inappropriateness of these existing statements. The proposed action statements are consistent with the STS philosophy, they assure that the considerations regarding safety analyses as stated in the bases are maintained and, as such, thes6 action statements are appropriate.
The change to the test requirement of Technical Specification 4.4 is a revision to the test acceptance criterion for 125 volt battery bank no. 1 to require 85% of the manufacturers rating. This acceptance criterion represents a 5% increase over the existing specification and is necessary due to the installation of the upgraded NIS. The current acceptance criterion of 80% of the manufacturer's rating is based on IEEE 485 recommendations. The acceptance criterion establishes a standard for battery replacement once the battery capacity has decreased to a specified value below the manufacturer's
-6 rating. The increased acceptance criterion value represents a more restrictive replacement standard. The reason for this increased conservatism is due to addition of battery loads, associated with the NIS upgrade, beyond the normal battery design margin. Consequently, to assure sufficient battery capacity throughout the battery service life, a more restrictive replacement criterion will be implemented. As battery capacity decreases due to normal agingfactors, replacement criterion assures that at end of service life the battery capacity is still sufficient to supply all necessary loads.
Analysis Conformance of the proposed amendments to the standards for a determination of no significant hazard as defined in 10 CFR 50.92 (three factor test) is shown in the following:
- 1. Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
RESPONSE: NO The revisions of this proposed change ensure that the new NIS will be operable and surveilled in an appropriate manner. The changes are consistent with the NIS system design and the STS, and, therefore, are considered to have been previously reviewed for use at PWRs with an NIS of Westinghouse design. The revision to the modifier of the nuclear overpower trip setpoint and the dropped rod rod stop are for consistency with the revised NIS detector orientation and the LCOs and surveillances will assure that the probability and consequences of previously analyzed are not changed. The dropped rod accident now postulates a two dropped rod scenario prior to the automatic safety action to preclude exceedance of the DNBR acceptance criterion, but the previously established criterion is met. The revisions to the IAO monitoring capability for the most part delete the information not applicable to the new NIS. The remainder of the IAO changes are consistent with previously analyzed accidents. Additionally, the revisions described for the battery test requirements will provide assurance of battery capacity commensurate with its duty requirements.
This revised acceptance criterion is required in order to be consistent with a recalculation of the duty requirements of this battery, giving consideration to the NIS upgrade. Accordingly, it is determined that operation of the facility in accordance with this proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
-7 RESPONSE: NO The revisions of this proposed change assure that the operation of the plant is restricted to within the bounds of previously analyzed accidents. Specifically, the NIS operability and surveillances are consistent with the STS and existing accident analysis. The revisions to the modifier of the nuclear overpower trip and the dropped rod rod stop ensure that the plant is maintained within the bounds of previously evaluated accidents. The dropped rod accident is now a double dropped rod event prior to the required protective action, but the DNBR acceptance criterion is met. The limits on IAO and IAO measurement capability are unchanged. The battery specification revision provides assurance that the system is tested in a manner consistent with its recalculated duty cycle.
Accordingly, it is determined that operation of the facility in accordance with this proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?
RESPONSE: NO The revisions of this proposed change provide assurance that the plant is maintained within previously evaluated margins of safety.
This conclusion is supported for the NIS upgrade by the safety review section of Attachment 3 to this proposed change. The revisions to the modifier of the nuclear overpower trip and the dropped rod rod stop are necessary to preserve an existing margin of safety, consistent with existing analyses. In the case of the dropped rod accident, the DNBR acceptance criterion of 1.30 is the safety margin that continues to be met. The IAO limits and component operability remain unchanged. The revision to the battery test acceptance criterion will serve to ensure that an existing safety margin is maintained. Accordingly, it is determined that operation of the facility in accordance with this proposed change does not involve a significant reduction in a margin of safety SAFETY AND SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Based on the Safety Evaluation, it is concluded that:
(1) Proposed Change No. 180 does not involve a significant hazards consideration as defined by 10 CFR 50.92; and (2) there is reasonable assurance that the health and safety of the public will not be endangered by the proposed change. - Existing Specifications - Proposed Specifications - NIS Safety Review Report - NIS Installation Design LAB:9321F