ML20058F595
| ML20058F595 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 11/22/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20058F587 | List: |
| References | |
| NUDOCS 9312080193 | |
| Download: ML20058F595 (10) | |
Text
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NUCLEAR REGULATORY COMMISSION wasniwovow, o.c. roswoooi SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 157 TO FACILITY OPERATING LICENSE NO. DPR-40 OMAHA PUBLIC POWFR DISTRICT I
FORT CALHOUN STATION. UNIT 1 i
DOCKET NO. 50-285
1.0 INTRODUCTION
By letter dated June 17, 1993, as supplemented on October 8,1993, Omaha Public Power District (OPPD) submitted a request for changes to the fort Calhoun Station, Unit 1 Technical Specifications (TS). The requested changes would implement administrative changes. The June 17, 1993, letter was noticed in the i
Federal Reaister on August 4, 1993. The supplemental information of October 8, 1993, was noticed in the Federal Reoister on November 8, 1993.
The licensee also submitted by letter dated November 11, 1993, a request that the amendment be issued on November 17, 1993, prior to the end of the 30-day notice period, which ends December 8, 1993. The letter stated the approval of Topical Report OPPD-NA-8302 was not approved until November 2, 1993, and since the supplemental submittal dated October 8,1993, was noticed for comment on November 8,1993, this would result in an excessive delay in startup after a refueling outage. Due to these circumstances, the staff has determined that the amendment can be issued prior to the end of the 30-day notice period.
2.0 EVALUATION The licensee's proposed changes to the fort Calhoun Station's TSs are administrative corrections. These changes are as follows:
Paae 3-0a. 3-1. 3-40. 3-54. 3-56. 3-60. 3-62. 3-77. 3-79. and 3-84 In order to provide consistency throughout the TSs with a defined term, the J
licensee proposed to revise surveillance requirements which state an "18 month," or "18 month during shutdown" interval to state that these surveillances are conducted on a " refueling frequency." Refueling frequency is defined in Specification 3.0.2 as at least once per plant operating cycle. To ensure that the 18 month interval is adequately stated in the TSs, the licensee proposed that Specification 3.0.2 be revised to reflect the Combustion Engineering Standard Technical Specification (CESTS) definition for refueling frequency which is at least once per 18 months.
Pace 1-4. 1-8. and 2-15a j
The Bases to Specification 1.2 and 1.3 are being revised to be consistent with Table 1-1, Item No. 5, and Specification 2.1.6 is being revised to add discussion concerning the presence of water-filled loop seals upstream of the pressurizer safety valves.
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. TS 2.1.6(1) requires that the pressurizer safety valves be operable with their lift settings adjusted to ensure valve opening at 2500 psia i 1% and 2545 psia i 1%. The setpoints are established in accordance with the American Society of Mechanical Engineers (ASME) code as required by Specification 3.3(1). The ASME code requires that safety valves whose design basis is to relieve steam pressure be set using steam inlet conditions. Upstream of the pressurizer safety valves are water-fliled loop seals designed to reduce leakage of non-condensible gases through the valves.
Since the ASME code requires that the setpoints be established using steam, the presence of the water-filled loop seal could potentially influence the pressure at which these valves open.
To address this issue OPPD' completed Engineering Analysis EA-FC-92-066 which verified that the reactor coolant system could withstand an overpressure transient with a safety valve setpoint deviation of +6% and be within the results in the Updated Safety Analysis Report (USAR).
In addition, it was determined that a safety valve setpoint deviation as low as -4% would not cause unnecessary challenges to safety systems. The Basis of Specification 2.1.6 is being revised to incorporate this additional information concerning the presence of the water filled loop seals and potential setpoint deviations.
In addition, the Basis of Specification 1.2 and 1.3 are being revised to indicate that the reactor high pressure trip is set at less than or equal to 2400 psia which is consistent with the requirements of Specification 1.2, Table 1-1, Item No. 5, and that the power-operated relief valve (PORVs) setpoint is consistent with the reactor high pressure trip.
The wording " steam system safety valves," is also being revised in the basis of Specification 1.2 to " main steam safety valves" to be consistent with wording implemented in Amendment No.146.
The word "or" is being corrected to the word "of" in the Basis of Specification 2.1.6.
Pace 4 The definition of Channel Check contained on page 4 is being revised to correct a typographical error. The word "behaviour" is misspelled and is being corrected to read " behavior."
Pace 2-18 Specification 2.2(2)dl. is being clarified by adding valve LCV-018-3 to the equipment required to be operable when the required volume of boric acid may be combined between Boric Acid Storage Tanks (BAST) CH-IIA and CH-IIB.
Specifications 2.2(2)d2. through 2.2(2)d4 state the requirements when LCV-218-3, CH-IIB, and CH-IIA are inoperable.
Pace 2-20 Specification 2.3(1)c. is being revise to clarify that it is required that all four safety injection tanks have a tank " level" of at least 116 inches.
The current specification requirement to have a tank " liquid" of at least
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. 116 inches is not gramatically correct and is being revised.
Specification 2.3(1)e. and 2.3(1)f. are being clarified to indicate that it is required to maintain at least one low pressure safety injection pump and one high pressure safety injection pump on each " associated 4160 V engineered safety feature" bus. These s there is electrical independence for the pumps.pecifications ensure that There are two low pressure safety injection pumps, one powered from each 4160 Y bus. There are three high pressure safety injection pumps powered from three 480 V buses. Two of the 480 V buses are independently powered, one from each 4160 V bus, and one is a " swing bus" which could be powered from either of the 4160 V buses.
Specification 2.3(1)f. ens ~ res the minimum requirements of maintaining two u
high pressure safety injection pumps with electrical independence.
Specification 2.3(1)j. ensures maintaining two high pressure safety injection pumps with independent suction sources. This change only clarifies that the buses stated in Specifications 2.3(1)e. and 2.3(1)f. are the 4160 Y buses and not the 480 V buses, thereby requiring that a minimum of two independent high and low pressure safety injection pumps are operable.
Paoe 2-22 It is proposed to revise the Basis of Specification 2.3, " Emergency Core Cooling System (ECCS)," to delete the reference to low temperature / low power physics testing. This low temperature testing refers to the one-time testing conducted at 200*F which was performed during the initial startup (Cycle 1). Specification 2.3 requires minimum ECCS equipment be operable before the reactor can be made critical. There are no special t e exceptions stated in Specification 2.3.
Therefore, these requireme.nts also apply when the reactor is made critical during low power physics testing and the statement in the Basis which indicates that ECCS is not required is incorrect and is being deleted.
Pace 2-57e The Basis to Specification 2.10.4 is being revised to delete the specific steps of how to measure reactor coolant system (RCS) flow by using reactor coolant pump differential pressure. The specific steps on conducting this test are appropriately included in procedures, and it is inappropriate to include the specific steps in the Basis of a specification. The phrase "that will" is being revised to "may" to indicate that pump differential pressure is not the only available method for determining RCS flow.
Paoe 2-58 Reference (1) on page 2-58 is being revised from Final Safety Analysis Report (FSAR), Section 14.18 to the current nomenclature for this document which is Updated Safety Analyses Report (USAR) Section 14.18.
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Paoes 2-62 and 2-64 It is proposed to revise Specification 2.14(3) " Engineered Safety Features System Initiation Instrumentation Settings - Containment High Radiation (Air Monitoring)" and Table 2-1, " Engineered Safety Features System Initiation Instrument Setting Limits" to correct inconsistences between TS 2.14 and the Offsite Dose Calculation Manual (ODCM). Radiation Monitors RM-050 and RM-051 are process monitors; however, they may be considered to be effluent monitors w1en monitoring the Auxiliary Building Exhaust Stack.
The ODCM is utilized to control effluent radiation monitor setpoints but not process radiation monitor setpoints. Therefore, it is proposed that isolation function setpoints for effluent monitors be calculated in accordance with the ODCM and isolation function setpoints for process radiation monitors be calculated in accordance with the applicable Chemistry Manual calibration procedure.
Paoe 2-62 It is proposed to revise the Basis of TS 2.14(5) to delete the specific value for the valve stroke time. The Basis for this specification is that the valves close in sufficient time to ensure adequate net positive suction head is available to tiie safety injection pumps. Deleting the specific stroke time will make the Basis of Specification 2.14(5) consistent with the remainder of the TSs which do not state specific valve stroke times.
The reference to the FSAR loss of coolant accident analysis described in the Basis of 2.14(5) is being revised to the correct nomenclature for this document which is the USAR.
Pace 2-66 Specification 2.15(2) is being revised to correct a typographical error.
The statement "... channel has not been stored to operable status," is being corrected to read "... channel has not been restored to operable status."
The word " store" is being replaced by the correct word " restored."
Specification 2.15(3) is being revised to correct a typographical error.
The statement '... falls below the limits given in the columns entitled
" Minimum Operable Channels" of " Minimum Degree of Redundancy",' is being corrected to read '... falls below the limits given in the columns entitled
" Minimum Operable Channels" or " Minimum Degree of Redundancy".' The word "of" is being replaced by the correct word "or."
Pace 3-15 Table 3-3, Item 12, Surveillance Method, is being clarified to indicate that the known pressure is applied to two separate pressure transmitters.
The known pressure cannot be applied to the pressure switch. The
" redundant interlock" discussed in this item is a separate pressure transmitter in a separate instrument loop. The proposed wording only clarifies where the known pressure is applied.
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. Pace 3-18 Table 3-4, Item 1.(a)(2)(ii), is being revised to make the "(1)" a superscript as this item applies to Note (1) contained on page 3-19.
Table 3-4, Item 1.(a)(2)(111), is being revised to delete redundant wording. The phrase " change exceeding 15% of the rated thermal power" is included twice in the specification.
Table 3-4, Item 1.(b)(2)(1), is being revised to state the sample frequency of I per "8" hours and to make the "(1)" a superscript as this requirement applies to Note (1) contained on page 3-19.
These items were inadvertently incorporated in Amendment No. 133.
Paae 3-21 It is proposed to revise Specification 3.3(2)a, to add an asterisk that was inadvertently deleted in an earlier amendment. The deletion o: curred in Amendment No. 104 when Specification 3.3 was reorganized to incorpo-rate a separate specification for steam generator tube inspections.
Pace 3-60 Specification 3.7(3) is being revised to clarify that the emergency lighting system required to be surveillance tested in accordance with this specification is the emergency lighting required for plant safe shutdown.
f_aae 3-61 Specification 3.8 is being revised to update the reference. Reference (1) is being revised from "FSAR" to the current nomenclature for this reference s
which is "USAR."
E_aoe 3-62 Specification 3.9(2) and Specification 3.9(4) are being combined. As presently written, Specification 3.9(4) is not a surveillance, but is the acceptance criteria for the surveillance required by 3.9(2), therefore, it is appropriate for these two specifications to be combined.
It is also proposed that the statement concerning the location of where readings will be deleted as it is unnecessary. Specification 3.9(5) and 3.9(6) are renumbered to reflect the deletion of 3.9(4). Specificaticn 3.9(6)a. and 3.9(6)b. are being revised to change the word " verifying" to " verify." The phrase "at least" is being added to Specification 3.9(4) to clarify that the acceptance criteria for pump pressure is at least 40 psig above the steam generator pressure at rated steam flow.
Pace 3-76 It is proposed to revise the numbering of Specification 3.12, " Radioactive Material Sources Surveillance," to correct a typographical error. The correct number is " Specification 3.13," as " Specification 3.12" is,
" Radioactive Waste Disposal System."
. Paae 3-84 It is proposed to revise the word " valves" contained in Specification 3.16(1)d. to the correct word " valve."
Pace 4-4 Specification 4.4.1 is being clarified. The current discussion implies that the floor below the new fuel rack is made entirely of open grating, i
which is not true. The proposed change would clarify that the design basis of the storage area for new fuel is to preclude flooding.
Paae 5-1 Specification 5.2.2.a. is being revised to add a line which was inadvertently deleted in Amendment No.132. The requirement states that "The minimum number and type of licensed and unlicensed operating Table 5.2-1."
The requirement should state, " The minimum number and type of licensed and unlicensed operating personnel required onsite for each shift shall be as shown in Table 5.2-1."
The line " personnel required onsite for each shift shall be as shown in," is being added to correct this specification.
Paaes 5-4. 5-5. 5-6. 5-7. 5-8. and 5-9 It is proposed to revise Specification 5.5 to reflect organizational changes, title changes and to revise the submittal of Safety Audit and Review Committee (SARC) reports to the Senior Vice President from 14 days to 30 days. The title changes involve: revising the title " Chairman" to
" Chairperson," clarifying the SARC membership by allowing additional technical experts to be members at the discretion of the SARC Chairperson, deleting the Manager - Radiological Services, and adding the Vice President as a member. The timeframe for submittal of SARC reports to the Senior Vice President is being revised because the Senior Vice President-is a member of the SARC as specified in Specification 5.5.2.2.
Therefore, this position is knowledgeable of actions taken by this committee, and 30 days is more than adequate for submittal of the written report. Thirty days is consistent with NUREG-1432 Specification 5.5.2.c.
Pace 5-6 Specification 5.5.2.4 is being revised to correct a typographical error.
The word " advise" is being replaced with the word " advice" to correct the error.
Paaes 5-7. 5-8. 5-Ba. and Pace iii of the Table of Contents It is proposed to revise Specification 5.5.2.8 and 5.5.3 to be more consistent with CESTS 6.5.2.8 (NUREG-0212, Revision 2). Specification 5.5.3 will be deleted and incorporated into 5.5.2.8.
Since Specification 5.5.3.a is being moved to 5.5.2.8, the sentence stating that this audit is under the cognizance of the SARC is no longer required and is being deleted. The audit schedule for review of the Emergency Plan and
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. Safeguards Contingency Plan are being maintained at a 12 month interval consistent with 10 CFR 50.54. The requirement to audit the Radiological Effluent Program, Specification 5.5.2.8.h, is being maintained as it is included in NUREG-1432. Page iii of the Table of Contents is being revised to reflect the deletion of Specification 5.5.3.
Pace 5-12 It is proposed to revise Specification 5.9.1.c to indicate that the Monthly Operating Report will be submitted "no later than the fifteenth of each month" instead of "to arrive no later than the fifteenth of each month."
This requirement is consistent with CESTS 6.9.1.6 (NUREG-0212, Revision 2).
Pace 5-17a It is proposed to revise Specification 5.9.5 to remove the revision numbers and dates of the NRC approved reload analysis topical reports and indicate reference to the latest NRC approved revision as stated in the Core Operating Limits Report (COLR). Also, Item No. 4, which references an NRR Safety Evaluation related to Amendment No. 143, is being deleted. The analytical method approved by the NRC in the Safety Evaluation Report was incorporated into the latest revision of OPPD-NA-8302-P-A "Neutronics Design Methods and Verification," therefore, Item No. 4 no longer applies.
These changes are administrative in nature; and therefore, the staff finds them acceptable.
3.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION
DETERMINATION The licensee submitted an application for amendment on October 8, 1993, that was required in order to support plant startup from the 1993 refueling outage. The changes requested were administrative in nature, in that only the revision numbers of NRC approved core reloed analysis topical reports were being changed.
Failure to receive approval of this application for amendment before November 17, 1993, would delay the startup.
As accurately detailed by the licensee in the attachment to its November 11, 1993, letter OPPD submitted revisions to the fort Calhoun Station Unit No. I core reload topical reports on February 2,1993 (LIC-93-0015), in order to support Cycle 15 operation. These revisions were editorial in nature, and incorporated methodology changes previously approved by the NRC for use at other Combustion Engineering plants.
Technical Specification 5.9.5 requires that the analytical methods used to determine the core operating limits be those previously reviewed and approved by the NRC. The specific revision numbers of the topical reports are listed in TS 5.9.5.
The application for amendment of TS 5.9.5 is purely an administrative action to reflect the latest revision of the topical reports that have been approved by the NRC as referenced in the core operating limits report.
An application for amendment was processed within OPPD to reflect the anticipated approval and the new revision numbers of the topical reports. This
1 was ready for submittal to the NRC in August 1993, but could not be sent until after OPPD received NRC approval of the topical reports.
Topical report OPPD-NA-8301-P Rev. 5 was approved by the NRC on July 29, 1993 and OPPD-NA-8303-P Rev. 4 was approved August 18, 1993. Since topical report OPPD-NA-8302-P Rev. 3 had not received NRC approval consistent with OPPD's planned schedule, it was discussed with NRR staff whether the specific revision numbers of the topical reports needed to be stated in the TSs. Following NRR staff guidance received on August 11, 1993, OPPD initiated a new application for amendment to delete the specific revision numbers, thus eliminating the need to process future TS changes each time a new topical report revision is approved.
On August 17, 1993, OPPD asked NRR staff if TS 5.9.5(b)(4), which references reload methodology that was approved in the Amendment No.143 Safety Evaluation, should be deleted from the TSs. The methodology from Amendment No. 143 was incorporated into topical report OPPD-NA-8302-P Rev. 3.
Therefore, it would be redundant to list both the topical report and the Amendment No.143 Safety Evaluation, after OPPD-NA-8302-P Rev. 3 is approved. On August 27 and September 20, 1993, NRR was requested to provide guidance concerning the proposed deletion of TS 5.9.5(b)(4). The NRR staff opted to defer judgement until the review of topical report OPPD-NA-8302-P Rev. 3 was completed.
OPPD expressed to NRR staff a concern regarding inadequate time to complete the i
public comment period for the proposed amendment and obtain approval prior to the anticipated startup following the refueling outage.
Based on direction from NRR, an application for amendment was submitted on October 8, 1993 (LIC-93-0256) as a supplement to an application for amendment dated June 17, 1993 (LIC-93-0159) which contained only editorial and administrative changes.
Topical report OPPD-NA-8302-P Rev. 3 did not receive NRC approval until November 2, 1993. On the same day, OPPD was informed that the amendment supplement would require an additional 30-day public comment period. Therefore, the status of TS 5.9.5(b)(4) could not be ascertained until OPPD-NA-8302-P Rev. 3 was approved, and since the October 8, 1993, application for amendment was not noticed in the Federal reaister until November 8,1993, the need for an emergency amendment to the TSs could not have been foreseen by OPPD.
The Commission's regulations in 10 CFR 50.92 state that the Commission may make a final determination that a license amendment involves no significant hazards considerations if operation of the facility in accordance with the amendment would not (1) involve a significant increase in the probability or consequences of 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.
The Commission has determined that the amendment involves no significant hazards consideration per 10 CFR 50.92, based on the licensee's analysis provided in their October 10, 1993 letter and presented below:
(1) The proposed changes include:
administrative changes to correct typographical errors, references and revision numbers; make the specifications consistent; provide clarifications; and to make changes consistent with organizational changes, or with the CE Standard Technical Specifications.
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The clarification to the basis of Specification 2.1.6 provides a discussion on:
the presence of water filled loop seals, the potential effects the loop seal may have on the setpoint deviation of the safety valves, and that any effect is within the results of the Updated Safety Analysis Report.
The clarification to Specification 2.2(2)dl. provides an additional requirement to maintain valve LCV-218-3 operable which is consistent with the intent of the specification in that the valve must be operable to maintain the required flow path from the Safety Injection and Refueling Water (SIRW) tank.
The clarification to Specification 2.3(1) states which electrical buses the safety injection pumps are powered through and is consistent with the Updated Safety Analysis Report, Section 14.15, which assumes that only one full capacity high pressure pump and one full capacity low pressure pump are available during a Loss of Coolant Accident.
The clarification to the basis of Specification 2.14 only deletes the reference to the specific time for a valve to open.
The clarification to Specification 3.7(3) adds verbiage to state that the emergency lighting system required to be tested by this specification is the emergency lighting system required to achieve a plant safe shutdown.
The proposed revision to Specification 5.9.5 merely incorporates reference to the latest NRC approved revisions of the topical reports as stated in the Core Operating Limits Report (COLR). The change does not modify the methodology or the manner in which they may be implemented.
The proposed changes are administrative in nature and are consistent with the assumptions or results stated in the Updated Safety Analysis Report; therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
(2) The proposed administrative changes are typographical errors, revision numbers and references, and implement changes to make the Specifications consistent. No new or different operation of plant equipment is proposed. No new or different action statements are proposed. Therefore, the proposed changes do not create the possibility of a new or different type of accident.
(3) The proposed administrative changes correct typographical errors and references, and implement changes to make the Specifications consistent. The clarifications being proposed are within the assumptions or results as stated in the Updated Safety Analysis Report; therefore, the proposed changes do not involve a significant reduction in a margin of safety.
-,. The NRC staff has reviewed the licensee's analysis and, based on this review, concludes that the analysis demonstrates that the applicable criteria are met.
Accordingly, the Comission has made a final determination that the amendment involves no significant hazards consideration.
4.0 STATE CONSULTATIQN In accordance with the Commission's regulations, the Nebraska State official was notified of the proposed issuance of the an.endment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR 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 Connission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (58 FR 41509 and 59280). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
In addition, this amendment also relates to changes in recordkeeping, reporting, or administrative procedures or requirements.
Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(10).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 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 t
security or to the health and safety of the public.
Principal Contributor: Steven Bloom Date: November 22, 1993
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