ML20212H238
| ML20212H238 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 09/22/1999 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20212H221 | List: |
| References | |
| NUDOCS 9909300248 | |
| Download: ML20212H238 (7) | |
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UNITED STATES g
g NUCLEAR REGULATORY COMMISSION c
f WASHINGTON, D.C. 2066Hooi I.
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.228 TO FACILITY OPERATING LICENSE NO. DPR-49 IES UTILITIES INC.
CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE DUANE ARNOLD ENERGY CENTER DOCKET NO. 50-331
1.0 INTRODUCTION
By letter dated April 30,1999, IES Utilities Inc., the licensee for the Duane Arnold Energy Center (DAEC), submitted proposed changes to Technical Specification (TS) Surveillance Requirement (SR) section 3.4.3.1, "SafetyRelief Valves (SRVs) and Safety Valves (SVs)." The proposed change allows an increase in the as-found SV and SRVs. safety mode setpoint tolerance, determined by test after the valves have been removed from service, from +1%/-3%
to
- 3%.
2.0 BACKGROUND
Appendix A, General Design Criterion 15, " Reactor coolant system design," of 10 CFR Part 50 states that "The reactor coolant system and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the
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reactor coolant pressure boundary are not exceeded during any condition of normal operation, j
including anticipated operational occurrences."
i The proposed change does not alter the SRV safety lift setpoints, relief setpoints, the SRV lift setpoint test frequency, or the number of SRVs required to be operable. Also, the proposed.
change requires the as-left safety valve function settings to be within *1% of the specified nominallift setpoints prior to installation before testing. The NRC staff has previously granted approval to individual Boiling Water Reactor (BWR) licensees to increase the as-found SRV tolerance to 3%. The basis for the approval was an NRC staff safety evaluation report (SER) for a licensing topical report NEDC-31753P, "BWROG In-Service Pressure Relief Technical Specification Revision Topical Report," evaluating the setpoint tolerance increase. The NRC staff SER included six conditions which must be addressed on a plant-specific basis for licensees applying for the increased SRV setpoint tolerance:
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(a) Transient analysis of all abnormal operational occurrences as described in NEDC-31753P, should be performed utilizing a i3% tolerance for the safet/ mode of spring safety valves (SSVs) and SRVs. In addition, the standard reload methodology (or other method approved by the NRC staff) should be used for this analysis.
(b). Analysis of the design basis over-pressurization event using the 3% tolerance limit is required to confirm that the vessel pressure does not exceed the ASME pressure vessel code upset limit.
(c) The plant-specific analysis' described in items (a) and (b) should assure that the number of SSVs, SRVs, and relief valves (RVs) included in the analyses correspond to the number of valves required to be operable in the TS.-
(d) Reevaluation of the performance of high pressure systems (pump capacity, discharge pressure, etc.), motor-operated valves, and vessel instrumentation and associated piping must be completed, considering the 3% tolerance limit.
-(e) Evaluation of the i3% tolerance on any plant-specific operating modes (e.g., increased core flow, extended operating domain, etc.) should be completed.
(f) Evaluation of the effect of the 3% tolerance limit on the containment response during loss of coolant accidents and the hydrodynamic loads on the SRV discharge lines and containment should be completed.
3.0 EVALUATION The saft@ objective of the SRVs is to prevent over-pressurization of the nuclear system. This protects the nuclear system process barrier from fanure which could result in the uncontrolled release of fission products. The pressure relief system at DAEC includes six SRVs and two SVs. One SRV is set at 1110 psig, a second SRV is set at 1120 psig, two SRVs are set at 1130 psig, and two more SRVs are set at 1140 psig. The two SVs are set at 1240 psig,10 psig less than the vessel design pressure of 1250 psig. Existing TS provides a *1% as-found tolerance and *1% as-left setpoint tolerance. The proposed modifications would provide a 13 L % as-found tolerance and *1% as-left setpoint tolerance. The licensee's submittal was evaluated against the generic SER described above.
3.1 Transient Analysis / Reload Methodoloav The licensee must consider the impact of the tolerance increase on abnor. mal operational transients (AOTs). For DAEC, analysis (cycle 16 reload analysis) of AOTs has been conducted utilizing the 3% tolerance and with all SRVs in service. All future reload analyses are expected to assume the 3% tolerance. The limiting events the licensee evaluated are all considered pressurization events involving a rapid vessel pressurization. The resulting sharp increase in neutron flux results in a rapid decrease in the critical power ratio (CPR). The flux increase is terminated by a scram which is initiated by the reactor protection system. The vessel pressurization is' subsequently relieved by SRV actuation. The evaluations determine the operating limit minimum critical power ratio (MCPR). The operating limit MCPR is that value for which CPR remains above the safety limit MCPR if one of these limiting events occurs. The O
w 3-analysis showed that the thermal limits of the limiting transient would not be affected by the relaxation of SRV setpoint tolerance. Further, other transient events remain non-limiting and bounded by the above event. The NRC staff finds the licensee's analysis acceptable because it was performed using a methodology previously approved by the NRC.
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3.2. Analysis of the Desian Basis Overoressurization Event The licensee is required to reevaluate the limiting design-basis pressurization transient using the 3% tolerance limit to confirm that the vessel pressure does not exceed the American Society of Mechanical Engineers (ASME) pressure vessel code upset limit. The ASME Boiler and Pressure Vessel Code Section lli permits pressure transients up to 10% over design pressure (110% x 1250 psig = 1375 psig). The limiting pressurization ACT analyzed is a main steam isolation valve (MSIV) closure event occurring at the end of full power life without credit for a reactor trip on MSIV position sensing. As discussed in Section 4.2 of NEDC-31753P, evaluations have determined that the most severe transient is the closure of all main steam isolation valves (MSIVs), followed by a reactor scram on high neutron flux, i.e., a failure of the direct scram signal on MSIV closure. The licensee analyzed the MSIV closure event using the NRC staff-approved model ODYN with the 3% tolerance. Analyses performed in support of the DAEC core operating limits report (COLR) for Cycle 16 show that the maximum vessel pressure (bottom head) is 1282 psig for MSIV closure with flux scram. This value is 93 psig less than the
- vessel design pressure of 1375 psig. _ Thus, increasing the setpoint tolerance to +3% will not exceed the vessel design ASME limit of 1375 psig, and is thereby acceptable to the NRC staff.
3.3. TS Ooerability Statement f'or SRVs..
The licensee has stated that the overpressure protection system must accommodate the most severe pressurization transient. Evaluations have determined that the most severe transient is the closure of all MSIVs, followed by a reactor scram on high neutron flux. The reload analysis for Cycle 16 assumes no SRVs or SVs out of service. In accordance with TS Limiting Condition for Operation (LCO) 3.4.3;the safety function of eight SRVs and SVs shall be OPERABLE.
This is acceptable to the NRC staff.
The surveillance frequency of the SRVs is specified in the plant TS to be in accordance with the plant in-service testing (IST) program. The IST program is required to meet the ASME Code
. which specifies that the SRVs must be tested at least every 5 years. However, the licensee stated that the current licensing basis for DAEC is that at least half of the SRV population is j
removed and tested each refueling outage. This test frequency is sufficient to meet the test frequency specified in the NRC SER for NEDC-31753P, and is acceptable.
1 3.4. Reevaluation of the Performance of Hioh Pressure Systems The licensee reevaluated performance of high pressure systems (pump capacity, discharge i
pressure, etc.), considering the 3% tolerance limit. The DAEC utilizes a Low-Low Set (LLS) system which controls reactor pressure between 900 and 1025 psig (nominally). The LLS logic is armed when one SRV actuation occurs coincident with a high reactor pressure scram signal.
As discussed in NEDC-31753P and the accompanying technical evaluation report (TER), the increased setpoint tolerance will have no noticeable impact on the pressures and temperatures l
4 pressure, etc.), considering the 3% tolerance limit. The DAEC utilizes a Low-Low Set (LLS) system which controls reactor pressure between 900 and 1025 psig (nominally). The LLS logic is armed when one SRV actuation occurs coincident with a high reactor pressure scram signal.
As discussed in NEDC-31753P and the accompanying technical evaluation report (TER), the increased setpoint tolerance will have no noticeable impact on the pressures and temperatures at which the high pressure systems would be required to operate. No significant impact on the performance of the high pressure systems is expected, including the generic letter 89-10 program motor-operated valves.
3.5. Evaluation of Plant-soecific Ooeratina Modes Evaluation of the i3% tolerance on any plant-specific alternate operating moden (e.g., increased core flow, extended operating domain, etc.) was completed by the licensee, in NEDC-31753P, plants are categorized based on the combination of valves used for pressure relief (either safety valves, relief valves, or' safety relief valves). The DAEC has a combination of Target Rock safety relief valves and Dresser safety valves and is categorized as a Group 2 plant.
The BWROG reviewed the abnormal operating occurrences for each plant category and identified the most severe (limiting) events in NEDC-31753P, The limiting events for a Group 2 plant are typically:
(1) Load rejection without bypass (2) Feedwater controller events (3) Turbine trip without bypass As discussed in NEDC-31753P, the evaluations of these events were performed using GE standard reload licensing methodology as described in NEDE-2401 1 -P-A-9 (General Electric Standard Application for Reactor Fuel) and NEDE-2401 1 -P-A-9-US (General Electric Standard Application for Reactor Fuel (Supplement for United States)). These evaluations show that the MCPR occurred prior to the first SRV actuation.
DAEC plant-specific alternate operating modes are analyzed as part of the reload analysis. The DAEC has used +3% of 1110 psig (1143.3 psig) as the lowest safety relief valve setpoint and
+3% of 1240 psig (1277.2 psig) as the lowest spring safety valve setpoint for the Cycle 16 reload analysis. Alternate operating modes analyzed as part of the DAEC Cycle 16 reload analysis include:
(1) Single-loop operation (2) - Extended load-line limit
. (3) ARTS (Average Power Range Monitor, Rod Block Monitor and Technical Specification j
Improvement) Program The analyses included evaluations for the currently approved operating domains: Maximum Extended Operating Domain (MEOD), increased Core Flow and Single Loop Operation. The
. NRC staff has found the analyses to be acceptable, i
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> 3.6. Containment Resoonse/ Hydrodynamic Loads Evaluation of the effect of the 3% tolerance limit on the containment response during loss of coolant accidents and the hydrodynamic loads on the SRV discharge lines and containment were completed.
The effect of SRV actuation at potentially higher pressures may affect the containment t
temperature and pressure response, as well as the dynamic loading due to SRV actuation.
Plant unique design features make it necessary to perform these evaluations on a plant specific l
basis. The licensee performed these evaluations by demonstrating that the current containment I
pressure and temperature responses bound those which would be obtained with the SV or SRV
- actuation at +3%.
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3.6.1 Containment Response during Loss of Coolant Accidents The BWROG identified three postulated pipe break scenarios and discussed the effect of SRV actuation at a higher pressure on each: limiting break LOCA, small break LOCA, steamline break outside of containment.
For the limiting break LOCA, the reactor vessel depressurizes very rapidly through the break.
Because the vessel immediately depressurizes, no SRV actuation will occur. Therefore, an increase in SRV opening pressure does not impact the limiting (large) break LOCA analysis.
For a small break LOCA, the vessel depressurizes much more slowly, As the break size becomes smaller, the vessel may remain near the normal operating pressure, and upon vessel isolation (MSIV closure) the vessel may pressurize and open SRVs.
At the DAEC, a Low-Low Set (LLS) relief system is installed that controls the reactor pressure between 900 and 1025 psig. The LLS system is initiated upon receipt of a high pressure scram signal along with an SRV actuation. With a setpc;nt tolerance increase to +3%, the first SRV actuation could occur at a higher pressure. Hpwever, the increased SRV opening pressure will only affect the timing of the first SRV actuatba. Once the logic is initiated, the opening and closing setpoints of pre-selected SRVs are automatically reset to lower values by the LLS logic.
This logic is unaffected by the setpoint tolerance change since the logic acts on the relief mode of SRV actuation and not on the safety mode of operation. The acceptance criteria given in 10 CFR 50.46 are still satisfied for all break sizes and locations and hence the setpoint tolerance change for LOCA considerations is acceptable.
The steamline break outside containment also results in vesselisolation and pressurization. As previously discussed, the first SRV actuation could occur at a higher pressure due to LLS operation. Therefore, the effect on the steamline break outside of containment is negligible.
Thus, the NRC staff finds this acceptable.
3.6.2 Hydrodynamic Loads on the SRV Discharge Lines and Containment
. The large break LOCA is the limiting event for peak containment pressure and LOCA pool swell
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4 appropriate model-data comparisons for the SRV Line Clearing Model are documented in General Electric Company, Report No. NEDO-23 749-1, September 1978, " Comparison of Analytical Model for Computing Safety / Relief Valve Discharge Line Transient Pressures and Forces to Monticello TOuencher Data." To ensure that a conservative load definition is obtained, when applying the SRV Line Clearing Model, the licensee assumed an SRV flow rate of 1.225 times the ASME rated SRV flow as described in NEDO-21888, May 1984.
On March 27,1985, the NRC issued Amendment No.115, permitting an increase in the DAEC rated power from 1593 megawatt thermal (Mwt) to a maximum steady state core power level of 1658 Mwt. The power uprate raised the SRV opening pressure -setpoints by 30 psig. The opening setpoint of the SVs was not modified. The modified SRV opening pressure setpoints are 1110 psig,'1120 psig,1130 psig, and 1140 psig. The licensee stated that the analysis was applied to the DAEC SRV Discharge Lines at 103% of nominal setpoint pressure as described in the Mark l Containment Program Load Definition Report, GE Report, NEDO-21888 Class 1, Revision 2, November 1981. The SRV nominal setpoints used were 1080 psig,1090 psig, 1100 psig and 1110 psig. A study was performed to evaluate the effects of the power uprate on the loads associated with the discharge of SRVs. The results of this study concluded the increased SRV setpoints would not cause a significant increase in SRV quencher discharge related loads.
The NRC staff has found the analyses to be acceptable.
3.7 Technical Soecification Chanoes The setpoint tolerance in TS SR section 3.4.3.1, "SafetyRelief Valves (SRVs) and Safety Valves (SVs)," is changed from +1%/-3% to *3%. This is acceptable as described in this SE.
For consistency, TS Bases page B 3.419 will be changed to indicate that the "SRV and SV setpnints are 3% for OPERABILITY; however the valves are reset to *1% during Surveillance to allow for drift." The NRC staff has no objection to the proposed basis section change.
4.0 STATE CONSULTATION
in accordance with the Commission's regulations, the Iowa State official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
This amendment changes a requirement with respect to installation or use of a facility
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component located within the restricted area as defined in 10 CFR Part 20 and changes a i
surveillance requirement. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluent 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 this
. amendment involves no significant hazards consideration and there has been no public comment on such finding (64 FR 38028). Accordingly, this amendment meets the eligibility i
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7-criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b),
no environmentallmpact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
6.0 CONCLUSION
j The NRC staff 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 j
Commission's regulations, and (3) the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
- Principal Contributor: Brenda Mozafari Date: September 22, 1999