ML20235H187
| ML20235H187 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 02/15/1989 |
| From: | Cockfield D PORTLAND GENERAL ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 8902230507 | |
| Download: ML20235H187 (6) | |
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Portland General ElechicConvisy David W. Cockfield Vice President, Nuclear February 15, 1989 Trojan Nuclear Plent Docket 50-344 License NPF-1 U. S. Nuclear Regulatory Comission ATTN: Document Control Desk Washington DC 20555
Dear Sirs:
Response to Request for Additional Information Updated FSAR. Amendment 6. Dated July 1. 1987 Amendment 6 to the Trojan Final Safety Analysis Report (FSAR) was submit.ted in accordance with Title 10 Code of Federal Regulations, Part 50.71(e)
[10 CFR 50.71(c)) on July 1, 1987. On August 31, 1988 the. Nuclear Regulatory Comission (NRC) requested additional information concerning Amendment 6 changes to FSAR Sections 3.2, 7.2, and 9.1.
Attached is Portland General Electric Company's (PCE) response to the NRC requOpt for this additional information.
We would be pleased to discuss any questions or coments you may have regarding this information.
Sincerely, me CW 440 log Attachment c40 c:
Mr. John B. Martin C00 Regional Administrator, Region V Ng U.S. Nuclear Regulatory Comission 00 l
LO O
. g Mr. William T. Dixon i
c4 State of Oregon
. k Department of Energy
- - ac Mr. R. C. Barr
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NRC Resident Inspector Trojan Nuclear Plant 121 S.W Samon Street Putkirv1 Oregon 97204 i
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Trojan Nuclear Plant Document Control Desk Docket 50-544 February 15, 1989 License NPP-1 Attachment Page 1 of 3 PORTLAND GENERAL ELECTRIC COMPANY'S (PCE).
RESPONSE TO THE NUCLEAR REGULATORY COMMISSION (NRC) REQUEST FOR ADDITIONAL INFORMATION CONCEREING AMENDMENT 6, TO THE TROJAN FINAL' SAFETY ANALYSIS REPORT (FSAR) 1.
Page 3.2-11 was revised to state that repair or replacement of rad-waste components is " accomplished in accordance with PGE's in-house position on Regulatory Guide (RG) 1.143".
The Staff does not endorse nor recognize, a priori, any licensee in-house position.
Furthermore, it is not the staff's practice to approve internal licenseo positions since they are subject to change over time.
Please revise the referenced paragraph to explicitly identify differ-ences and/or exceptions to the Regulatory Guide.
Should your revised position with respect to RG 1.143 differ from that upon which the Plant was licensed, then an appropriate review should be conducted to determine whether or not prior NRC review is required.
Response
Portland General Electric Company takes no exceptions to RG 1.143.
FSAR Section 3.2 will be revised in Amendment 9 to the FSAR to remove the reference to the in-house position.
2.
The reactor trip signal (RTS) and Engineered Safety Features (ESP) system accuracies as shown on Tables 7.2-3 and 7.3-4 respectively were revised.
For each parameter whose instruments or system accuracy was revised, discuss how the current Technical Specification actuation setpoints and tolerance for those parameters remain valid for the revised values, and provide basis for the determination that an unreviewed safety question does not exist (e.g., a change to the Technical Speci-fications is not required). The safety evaluation that was performed in accordance with PCE Nuclear Division Procedure NDP 100-5 may also be provided if it contains the requested information.
Response
I Table 1 and Table 2 illustrate the changes that wore made to Tables 7.2-3 and 7.3-4 in Revision 6 to the FSAR.
For each case, Tables 1 and 2 list the maximum pre-Amendment 6 instrument loop error values and the Amendment 6 values.
These maximum errors were added to or subtracted from the safety analysis limits, as appropriate, and the result was then compared to the Technical Specification trip setpoints.
The available margin is also noted in the tables.
w_---__-___
L TEojan Nuclear plant Document Control Desk Docket 50-344 February 15, 1989 License NpF-1 Attachment page 2 of 3 In May 1987, Calculation TE-127 was performed to determine the accur-acies for the reactor protection and engineered safety features actua-tion system (ESFAS) instruments.
The revised accuracies were compared to the safety analysis limits to determine if changes were required to the Technical Specification setpoints.
In two cases, for the pressur-izer low-pressure safety injection setpoint and the steam generator low-low water level reactor trip setpoint, it was determined the Technical Specification setpoint was non-conservative based on the revised instrument accuracies.
In June 1987, the trip setpoints for low-pressurizer pressure safety injection and steam generator low-low water level reactor trip were changed to accommodate the revised accuracies.
These changes were made before the Technical Specifications were revised based on the fact the Technical Specifications call for a setpoint of greater than or equal to a certain value. Since both setpoints were being revised upward, the new setpoints were in compliance with the Technical Speci-fleations. These changes were discussed with the NRC staff.
In July 1987, Amendment 6 to the FSAR was issued. The safety evalu-ation for the change incorrectly concluded the accuracy changes did not affect the Technical Specification setpoints.
In February 1988, it was recognized the revised instrument accuracy for the low-pressurizer pressure reactor trip (120.8 psig) had also caused the Technical Specification setpoint of 1 865 psig to become 1
non-conservative by 0.8 psig. This determination was based on the safety analysis assumption that the trip occurred 11845 psig as I
identified in FSAR Table 15.0-3.
Upon further investigation it was determined that some of the safety analyses were based on the reactor trip occurring at 11845 psig while others were based on 21825 psig.
In July 1988, a safety evaluation was performed to validate all of the safety analyses for 11825 psig. FSAR Table 15.0-3 was subsequently revised in Amendment 8 to reflect the change to 11825 psig.
In May 1988, License Change Application (LCA) 145, Revision 1, was submitted to the NRC to formally revise the Technical Specification setpoints.
j l
pGE has recently implemented corrective action to improve the quality of safety evaluations. The Nuclear Management and Resources Council (NUMARC) guidelines have been incorporated into the safety avaluation i
procedure. Additionally, training has been conducted on safety evaluations and only those who have successfully completed the training are qualified to perform safety evaluations.
1
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' Trojan Nuclear Plant Document Control Desk Dockot 50-344 February 15, 1989 License NPF-1 Attachment Page 3 of 3 p
t 3.
Table 9.1-3 regarding spent vuel Pool Cooling Pump data was revised and the revision, in part, showed an increase in the minimum available Net Positive Suction Head (NPSH) from 20 feet to 57 feet.
Discuss the factors that result in this increase in NPSH.
Response
The NPSH value of 57 feet was erroneously taken from the pump vendors data sheet.
This value does not represent the minimum available NPSH.
Recalculation of the minimum available NPSH based upon water temper-ature, flow rate, system loss in suction lines, elevations, and equipment location yields a value of 28 feet. FSAR Table 9.1-3 will be revised to reflect the correct value in Amendment 9 to the FSAR.
4.
Equation'3.7-22 was revised to reflect a highor stress limit.
(a) Since the seismic load in Equations 3.7-21 and 3.7-22 may not be j
the only " occasional load" in the required load combinations, explain why the other " occasional loads" (such as valves actua-tion loads) are not included in these two equations, and to which equation (s) their contributions are factored.
Response
The elements in FSAR Equations 3.7-21 and 3.7-22 are outlined in FSAR Table 3.9-27, Design Loading Combinations and Stress Limits for American National Standards Institute (ANSI) B31.7, Class 2 and 3 Piping. The other occasional loads, such as valve actua-tions, are f actored into FSAR Table 3.9-27 under the column Design Loading Combinations.
(b) The stress limit of 2.4Sh used in Equation 3.7-22 may not be adequate to ensure piping functionality under faulted Plant conditions. For Class 2/3 piping, the functionality of which is essential for mitigating consequences in a design basis accident, a lower stress limit of 1.8Sh is required. Explain how compli-ance to this requirement is reflected by the change of Equation 3.7-22.
Response
FSAR Equation 3.7-22 was not intended to address piping func-tionality. This equation addresses seismic stresses as combined with other loadings for the safe shutdown earthquake.
FSAR Section 3.9.3.1.2.2, Design Loading Combinations and Stress Limits, describes how operation following faulted conditions is ensured.
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