ML20073R497
| ML20073R497 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 05/31/1991 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20073R494 | List: |
| References | |
| NUDOCS 9106100100 | |
| Download: ML20073R497 (4) | |
Text
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o WAsHINoToN, D.C. 20666 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMEN 0 MENT NO. 57 TO FACILITY OPERATING LICENSE NO. NPF-47 GULF STATES UTILITIES COMPANY RIVER BEND STATION, UN!T 1 DOCKET NO. 50-458
1.0 INTRODUCTION
By letter dated June 4, 1990, and suppleme a by letters dated February 13, 1991, d
April 11, 1991, and May 2, 1991, Gulf States Utilities Company (GSU) (the licensee) requested an amendment to Facility Operating License No. NPF-47 for the River Bend Station (RBS), Unit 1.
The propos:d amendment would revise Technical Specifications (TSs) 4.0.5, 3.4.3.1, and 3.4.3.2 in accordance with the guidance provided in Generic Letter (GL) 88-01, "NRC Position on IGSCC in BWP Austenitic Stainless Steel Piping." GL 88-01 required licensees to:
(1) provide plans regarding pipe replacement or other measures taken to mitigate IGSCC and provide long-term reliability, (2) implement an inservice inspection (ISI) program for austenitic stainless steel piping which conforms to the positions in GL 88-01, (3) implement a TS change to include the ISI of austenitic stainless steel piping in accordance with GL 88-01, (4) implement TS cn.anges related to leakage detection in conformance with GL 88-01, and (5) notify the NRC when flaws are identified that do not meet IWB-3500 criteria of Section XI of the American Society of Mechanical Engineers (ASME) code or when a change in the condition of a cracktid weld occurs.
GSU responded by letters dated July 21, 1988 and May 12, 1989.
By letter dated January 25, 1990, the NRC forwarded its Safety Evaluation of the GSU response and requested TS changes be submitted that meet the staf f's ;30sitions in GL 88-01.
The June 4, 1990, February 13, 1991, April 11, 1991, and May 2, 1991, letters provided those TS changes.
The April 11, 1991, and May 2, 1991, letters provided information that did not change the initial proposed no significant hazards consideration determination.
Specifically, the proposed changes are:
(1) TS 4.0.5 would be modified to state that ISI for piping susceptible to IGSCC would be performed in accordance with GL 88-01; (2) TS 3.4.3.1 would be modified to reflect actions to be taken when leak detection systems are inoperable; (3) TS 3.4.3.2 would be modified to reflect a new unidentified leakage rate limit and actions to be taken when that limit is exceeded; and (4) the BASES of TS 3/4.4.3.1 would be revised to reflect the TS changes.
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Two editorial changes were also prcposed which would delete references to the first refueling outage, which has alreadt een completed.
2.0 EVALUATION The proposed change to TS 4.0.5, " Applicability, Surveillance Requirements,"
is based on the staff pesition in GL 88-01 on inspection methods and personnel and would add a requirement that ISI be performed in accordance with the NRC staff positions in GL 88-01.
GSU also committed to revise the Updated Safety Analysis Report (USAR) and the RBS procedures which involve inservice inspection of piping welds.
The proposed change is considered an enhancement to the current ISI program.
The proposed change to the ACTION statement of TS 3.4.3.1, " Leakage Detection Systems," is based on the staff position in GL 88-01 on leakage detection and would separate the action to be taken when leakage detection systems are inoperable into two parts.
The first part of the ACTION statement would allow operation up to 30 days with the drywell atmosphere particulate radioactivity monitoring system, or either the drywell air coolers condensate flow rate monitoring system or the drywell atmosphere gaseous radioactivity monitoring system inoperable, provided a " grab" sample is taken and analyzed once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Otherwise, the plant would be required to be in HOT SHUTOOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The second part of the proposed ACTION statement would allow continued operation for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with the drywell floor and/or pedestal sump drain flow monitoring systems inoperable.
If the systems could not be restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the plant would be required to be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The drywell leak detection system is descriu9d in Sectio., 5.2.5.1.1, " Detection of Leakage Within the Drywell," of che USAR aad consists of airborne and particulate radioactive monitorinq systems, thi drywell cooler condensate flow monitoring system, and the sump drain flow monitoring system.
The variables are continuously monitored and inaicated in the control room and comply with Positions C.3 and C.7 of Regulatory Guide (RG) 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems." No changes to the TSs regarding the radioactive monitoring systems or drywell cooler condensate flow system are proposed.
The sump drain flow monitoring system is divided into two subsystems, each having a 600 gallon capacity sump.
These are: the floor sump, located in the general drywell space at the 81 foot elevation and the pedestal sump, located under the reactor vessel at the 73 fcot elevation.
Both have a sensitivity for detection of leakage increase of one gallon per minute (gpe) in one hour.
This complies with Position C.3 of RG 1.45.
A programmable controller (PC) monitors the sumps and alerts the operators in the control room of increased unidentified leakage.
This alert occurs before the TS limit of five gpa unidentified leakage is reached and an alarm sounds in the control room.
When the PC is inoperable or during c surveillance procedure, the leakage flow l
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rate, which is the sum of both sump leakage flows, is calculated manually by the operators using Standard Operating Procedure (50P)-0104.
GSU is confident that the manual method of calculation yields complete and accurate information that demonstrates compliance with the TSs and does not inhibit or prevent the operators from identifying the leakage rates.
The proposed TS changes would allow 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for repair if either of the sump subsystems become inoperable.
During that time, the radiation monitoring systems would be capable of detecting a one gpm increase in leakage.
GSU has determined the radiation monitors provide adequate information regarding leakage rates for the 24-hour period the sumps are allowed out of-service.
The radiation monitoring systems, discussed in Section 11.5.2.1.3.4, " Containment and Diywell Atmosphere," of the USAR, are seismically qualified, powered from a safety-related Class 1E source, and have an acceptable sensitivity and range.
The systems comply with Position C.6 of RG 1.45.
The proposed change to TS 3.4.3.2, " Operational Leakage," is based on the NRC staff position in GL 88-01 on leak detection and would place the plant in a limiting condition for operation when a two gpm increase in unidentified leakage occurs in OPERATIONAL CONDITION 1.
GSU determined that during startup, leakage rates could increase as part of the normal filling, flow, and pressurization of the reactor coolant system, and would not be indicative of leakage resulting from IGSCC.
Additionally, in any other operational condition the probability of the occurrence of IGSCC would be low due to the short period of time the plant is in other operational modes and the conservative condition of the plant.
When the LCO is exceeded, the proposed ACTION statement would require that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the source of the increased leakage be identified as not being IGSCC or the plant would be required to be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The drain sump moni toring system will be used to identify leakage rates.
If the PC is inoperable, operators would use the same procedure, SOP-0104, to calculate the unidentified leakage rate, and the total leakage (averaged over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) and maintain the 24-hour total leakage in order to confirm compliance with TS 3.4.3.2.
GSU plans on revising this procedure to reflect the additional ACTION statement.
GSU is confident that the manual method of calculation yields complete and accurate information that demonstrates compliance with the TSs end does not inhibit or prevent the operators from identifying the leakage rate.
During the time when the drywell and/or pedestal sump drain flow monitoring systems are inoperable and the reactor coolant system leakage limits need to be verified, GSU proposes to use the radiation monitoring systems to identify one gpm leakage rate increases.
As discussed above, GSU has determined the radiation monitors provide adequate information regarding increases in leakage rates when other leakage detection monitoring systems are inoperable and comply with the staff positions in RG 1.45.
The staff position in GL 88-01 on monitoring frequency suggests a frequency of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, however, GSU proposes to continue to monitor reactor system leakage once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> as currently identified in TS Surveillance Requirement 4.4.3.2.1.
GSU bases this proposal on the fact that RBS implements a 12-hour operations shift and any increase in leakage will be alerted in the control room at one gpm based on the radiation monitors.
If the sump flow leakage increased to the five gpm setpoint, an alarm would sound and the plant would
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,9 be required to shutdown.
Therefore, there are no proposed changes regarding monitoring frequency and TS 4.4.3.2.1.
The two editorial changes to TS Surveillance Requirement 4.4.3.1.d. and 4.4.3.2.2.a. delete the references and footnotes regarding tests that were not required to be performed or tests that were to be performed during the first refueling outage.
The first refueling outage has been completed, therefore, the references and footnotes may be deleted.
Because of the changes to TS 3.4.3.1, GSU proposed changes to the BASES section 3/4.4.3.1.
The BASES provide the operators background, additional information and guidance on the TSs.
Based on the review of the GSU amendment request, the guidance provided in GL 88-01 and RG 1.45, and the applicable secticns of the USAR, Safety Evaluation Report, and Standard Review Plan, the NRC staff finds the proposed changes acceptable.
3.0 STATE CONSULTATION
In accordance with the Commissi7n's regulations, the Louisiana State of fichl 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 facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance reqJirements.
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 (56 FR 13664).
Accordingly, the amendment meets the eligibility criteria for 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 prepared in connection with tM issuance of the amendment.
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 j
defense and security or to the health and safety of the public.
Prtncipal Contributor:
Claudia M. Abbate, PDIV-2 Date:
May 31, 199)
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