ML20206J301
| ML20206J301 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 04/30/1999 |
| From: | UNION ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20206J299 | List: |
| References | |
| NUDOCS 9905120184 | |
| Download: ML20206J301 (9) | |
Text
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3.4.3 l
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SURVEILLANCE REQUIREMENTS l
SURVEILLANCE FREQUENCY L
SR 3.4.3.1 Verify the safety function lift set In accordance of the SRVs and SVs are as follows: points l
with the Inservice Number of Setpoint Testing Program SRVs (osia) i 1120'bl.0.T 1
1110 33.0
- 11.0. t 33.0 1
t 2
1130
- 11.0. t 33.0 l
2 1140 y 1.0.
33.Q Number of Setpoint SVs (osia) 1240h2.0.t 2
36 Following testing, lift settings shall be within 1 1%.
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NOTE--------------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.
1 Verify each SRV opens when manually 24 months actuated.
9905120184 990430 7
PDR ADOCK 05000331 P
pm DAEC 3.4-7 Amendment 223
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B 3.4.3
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BASES (continued)
SURVEILLANCE SR 3.4.3.1 REQUIREMENTS This Surveillcnce requires that the SRVs and SVs will open at the pressures assumed in the safety analysis of Reference 1.
The demonstration of the SRV and SV lift settings must be performed during shutdown,, since this is a bench test, to be done in accordance with the Inservice Testing Program. The lift setting pressure shall correspond to ambient conditions of the valves at no.ninal operating t
ures and pressures. The SRV and SV setpoints are
(~ i:.
-3 for OPERABILITY: however the valves are reset to
-i :.% during'the Surveillance to allow for drift.
The Surveillance Frequency is an accordance with the Inservice Testing Program requirements contained in the ASME Code.Section XI. This Surveillance must be performed during shutdown conditions.
SR 3.4.3.2 A manual actuation of each SRV is performed to verify that, mechanically, the valve is functioning properly and no blockage exists in the valve discharge line. This can be demonstrated by the response of the turbine control valves or bypass valves, by a change in the measured steam flow by pressure switches and thermocouple readings downstream of the SRV indicating steam flow, or by any other method suitable to verify steam flow. Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve. Also, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to centrol reactor pressure when the SRVs divert steam flow upon opening. Sufficient time is therefore allowed after the required pressure and flow are achieved to perform this test. Adequate pressure at which this test is to be performed is approximately 150 psig which is the lowest pressure EHC can maintain.
Adequate steam flow is represented by approximately 1.15 turbine bypass valves open.
Plant startup is allowed prior to performing this test because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME Code requirements, prior to valve installation. Therefore, this SR is modified by a Note that states the Surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam (continued)
OAEC B 3.4-19 Amendment 223 t
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- y-to NG-99-0598 Page1of6 Safety Assessment Description of Change The proposed change establishes more appropriate Technical Specification (TS) setpoint tolerance limits for the main steam safety and relief valves (SVs and SRVs) for the Duane Arnold Energy Center (DAEC). The DAEC TS surveillance requirement (SR) 3.4.3.1 -
currently allows a tolerance band of-3% to +1%. The setpoint tolerance limits will be -
changed to i3R The existing TS requirement that the valves be reset to il% following testing will remain.
Background
There are six SRVs and two SVs associated with the DAEC main steam system. These valves are provided for overpressure protection of the nuclear vessel in accordance with the ASME Code Section Ill to limit the peak allowable pressure to 110% of vessel design pressure or 1375 psig at the vessel bottom. The spring safety valve setpoint is 1240 psig, which is 10 psig less than the 1250 psig vessel design pressure.
The use of the i 1% allowable as-found SRV safety lift setpoint tolerance in plant TS was generic in the industry. As a result, the BWR Owners' Group (BWROG) developed NEDC-31753P,"BWROG In-Service Pressure Relief Technical Specification Revision Licensing Topical Report," to support the use of the i 3%'
SRV safety lift setpoint tolerance consistent with that specified in ASME Section XI requirements. ' On March 8,1993, the NRC Staffissued their Safety Evaluation (SE) of GE Licensing Topical Report (LTR) NEDC-31753P. In the SE, the NRC found the generic change of setpoint tolerance to i 3% to be acceptable. The valve setpoint is to be restored to within i 1% following testing.
The NRC indicated in the SE that licensees planning to implement TS changes to increase the SRV setpoint tolerance:; should provide the following plant specific analyses:
- 1. Transient analysis of all abnormal operational occurrences as described in NEDC-31753P, should be performed utilizing a i 3% setpoint tolerance for the safety mode of SSVs and SRVs. In addition, the standard reload methodology (or other method approved by the staff) l should be used for this analysis.
- 2. Analysis of the design basis overpressurization event using the 3% tolerance limit for the l
' SRV setpoint is required to confirm that the vessel pressure does not exceed the ASME L
pressure vessel code upset limit.
- 3. The plant specific analyses described in items 1 and 2 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 technical specification.
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- 4. Re-evaluation 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.
- 5. Evaluation of the i 3% tolerance on any plant specific alternate operating modes (e.g.,
increased core flow, extended operating domain, etc.) should be completed.
- 6. 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.
Assessment Plant specific analyses and evaluations have been performed which fulfill the six requirements of the NRC SE. The fbliowing discussion provides a summary of the results of these analyses.
- 1. Transient Analysis ofAbnormal Operational Occurrences l
An evaluation of the effects of pressui:zation transients on the fuel thermal limits was performed to determine whether the increase in SV and SRV safety function lift setpoint tolerance from +1% to +3% would be acceptable. The limiting events selected for evaluation are all considered pressurization events. In these events, a rapid vessel pressurization occurs, resulting in a sharp increase in neutron flux. The flux increase results in a rapid decrease in j
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 following data was used as input to the transient analysis for the DAEC Core Operating Limits Report (COLR) for the current cycle (Cycle 16).
Valve Manufacturer Nominal Trip Set Point Reload License Analysis Value Type (psig)
(psig)
SRV Target Rock 1110.0 1143.3 SRV Target Rock 1120.0 1153.6 SRV Target Rock 1130.0 1163.9 l
SRV Target Rock 1130.0 1163.9 l
SRV Target Rock i140.0 1174.2 SRV Target Rock 1140.0 1174.2 SV Dresser 1240.0 1277.2 SV Dresser 1240.0 1277.2 l
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to NG-99-0598 Page 3 of 6 A discussed in Section 4.3.4 of the BWROG LTR, in all of the base cases, the MCPR
< urred prior to the first SRV actuation. Since the SRVs open after the time of the MCPR and the peak surface heat flux remains the same as in the +1% base case, there is no impact on the calculated thermal limits for these events. Therefore there is no impact on thermal limits 'as a result ofincreased SRV setpoint tolerance to +3%.
- 2. Analysis ofDesign il asis Overpressurization Event The overpressure protection system must accommodate the most severe pressurization transient. As discussed in Section 4.2 of the BWROG LTR, evaluations have determined that the most severe transient is the closure oi'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 following assumptions and initial conditions were used in analyzing the MSIV closure with flux scram:
- 1. The initial reactor power is consistent with current licensing basis assumptions for vessel overpressure protection.
- 2. The initial core flow is 100 afrated.
- 3. Conservative end-of-cycle nuclear dynamic parameters are used.
- 4. The simultaneous closure of all MSIVs is initiated at the beginning of the event, with a 3-second MSIV stroke time.
- 5. The direct scram signal on MSIV position is assumed to fail, and the scram is initiated from a high neutron flux signal.
- 6. Consistent with the licensing basis for overpressure protection evaluations, no credit is i
taken for externally powered valves. Thus, the following modes of valve actuation are used:
a.). All SVs are operable and actuate in the safety mode.
b.) All SRVs are operable and actuate in the safety mode.
. A~nalyses performed in support of the DAEC 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. Therefore it is concluded that expanding the nominal setpoint tolerance to + 3% will not exceed the vessel design ASME limit of 1375 psig.
- 3. Number ofSVs andSRVs 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.
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Page 4 of 6 c
- 4. Performance ofHigh Pressure Systems i
The Technical Evaluation Report of the BWROG LTR states:
' BNL focussed it.= evaluation on Items 2 and 3 [small break LOCA and steamline break outside containment], since in 3 large break LOCA the reactor depressurizes very rapidly and no SRV actuation is expected to occur. For a small break, either inside or outside containment,' which does not result in reactor depressurization, SRV actuations are possible.
Plants were considered separately depending on whether they had low-low set (LLS) relief j
. system logic.' The only effect on BWR plants with an LLS relief system is that initiation of the logic would be slightly delayed. Once this logic is initiated, however, the opening setpoint is automatically reset to a lower value by the LLS logic. Subsequent SRV actuations would, therefore, not be affected by a higher SRV opening pressure."
The DAEC utilize.s a LLS system which controls reactor pressure between 900 and 1025 psig i
(nominally). The LLS logic L anned when one SRV actuation occurs coincident with a high '
reactor pressure scram signt..is discussed in the BWROG LTR and the 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. There will be no significant impact on the performance of the high pressure systems, including the generic letter 89-10 program motor-operated valves.
- 5. Evaluation ofPlant Specific Alternate Operating MOs
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- In the General Electric "BWROG In-Service Pressure Relief Technical Specification Revision Licensing Topical Report" NEDC-31753P (LTR), plants are categorized based on the combination of valves used for pressure relief (either safety valves, relief valves, or safety reliet tJ es). 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 LTR reviewed the abnormal operating occurrences for each plant category and identified the most severe (limiting) events. 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 the BRWOG LTR, the evaluations of these events were performed using GE standard reload licensing methodology 'as described in NEDE-24011-P-A-9 (General Electric Standard Application for Reactor Fuel) and NEDE-24011-P-A-9-US (General Electric Standard Application for Reactor Fuel (Supplement for United States)). These evaluations l
show that the MCPR occurred prior to the first SRV actuation.
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Page 5 of 6 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 I
Specification Improvement) Program
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. 6. Evaluation ofContainment Response during Loss ofCoolant Accidents and Hydrodynamic Loads on the SR V Discharge Lines and C<mtainment The effect of SRV actuation at potentially higher pressures may affect the containment 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 basis. These evaluations can be performed by demonstrating that the current containment pressure and temperature responses bound those which would be obtained with
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the SV or SRV actuation at + 3%.
l Containment Response during Loss ofCoolant 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.
1 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 can have no impact on 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 which controls the reactor pressure between 900 and 1025 psig. The LLS system initiates once a high pressure scram signal is exceeded and an SRV actuation has occurred. With a setpoint tolerance increase to
+3%, the first SRV actuation could occur at a higher pressure. The increased SRV opening pressure will only affect the timing of the first SRV actuation. 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. Therefore, the effect on the small break LOCA is negligible because only the initial SRV actuation
- could occur at a higher value.
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The steamline break outside containment also results in vessel isolation and pressurization.
l Again, only the first SRV actuation could occur at a higher pressure due to LLS operation.
l Therefore, the effect on the steamlir.c break outside of containment is negligible.
Hydrodynamic Loads on the SR V Discharge Lines and Containment The limiting event for peak containment pressure and LOCA pool swell loads is the large break LOCA. As discussed above, the reactor vessel will depressurize very rapidly through the break and no SRV actuation will occur. The increase in setpoint tolerance will therefore have no effect.
The SRV discharge line clearing loads are evaluated using a first principle analytical model which has been shown to conservativeiy predict full scale data recorded during extensive in-plant testing (" Mark I Containment Program Final Report - Monticello T-Quencher Test, General Electric Company," Report No. NEDO-21864, June 1979).
The analytical bases and appropriate model-data comparisons for the SRV Line Clearing Model are documented in " Comparison of Analytical Model for Computing Safety / Relief Valve Discharge Line Transient Pressures and Forces to Monticello T-Quencher Test Data," General Electric Company, Report No. NEDO-23749-1, September 1978. To ensure that a conservative load definition is obtained, when applying the SRV Line Clearing Model, the SP.V flow rate is assumed to be 1.225 times the ASME rated SRV flow. (NEDO-21888, May 1984).
The aforementioned analysis was applied to the DAEC SRV Discharge Lines at 103% of nominal setpoint pressure as described in the Mark I Containment Program Load Definition Report, GE Report, NEDO-21888 Class 1, Revision 2, November 1981. At the time of analysis the SRV nominal setpoints were 1080 psig,1090 psig.1100 psig and 1110 psig.
The DAEC subsequendy implemented License Amendment 115, Power Uprate, which 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 i140 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 that no i
significant increase in SRV quencher discharge related loads is expected to result from the increased SRV setpoints.
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to NG-99-0598 ENVIRONMENTAL CONSIDERATION 10 CFR Section 51.22(c)(9) identifies certain licensing and regulatory actions which are
- cligible for categorical exclusion from the requirement to perform an environmental j
assessment. A proposed amendment :o an operating license for a facility requires no environmental assessment if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant hazards consideration; (2) result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite; and (3) result in a significant increase in individual or cumulative occupational radiation exposure. IES Utilities Inc. has reviewed this request and detennined that the proposed amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR Section 51.22(c)(9). Pursuant to 10 CFR Section 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of the amendment. The basis for this determination follows:
Basis The change meets the eligibility criteria for categorical exclusion set forth in 10 CFR Section S t.22(c)(9) for the following reasons:
1.
As demonstrated in Attachment i to this letter, the proposed amendment does not I
involve a significant hazards consideration.
2.
There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite. This change affects the as-found safety valve and safety relief valve setpoint tolerance only and will not affect either the amount or type of effluents normally released from the plant.
3.
There is no significant increase in individual or cumulative occupational radiation exposure. This change affects the as-found safety valve and safety relief valve setpoint tolerance only ar.d will not significantly affect either individual or cumulative occupational radiation exposure.
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