ML20215M927
| ML20215M927 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 10/17/1986 |
| From: | Agosti F DETROIT EDISON CO. |
| To: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| VP-86-0141, VP-86-141, NUDOCS 8611030478 | |
| Download: ML20215M927 (21) | |
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October 17, 1986 VP-86-0141 Mr. James G. Keppler FRIORITY ROUTING Regional Administrator tfrtt_ guuL--'
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Dear Mr. Keppler:
Reference:
1)
Fermi-2 NRC Docket No. 50-341 NRC Licence No. IPF-43 2)
Detroit Edison to liRC Letter, " Meeting on Fermi-2's Startup Test Program",
VP-86-0098, dated August 8, 1986.
Subject:
Startun Tent Procram Chancen Reference 2) informed you of our. plans [o make changes to the Fermi 2 Startup Test Program and suggected an informational meeting on the subject.
At a meeting on September 24, we diccussed the startup test modifications we were planning and the process we were intending to follow to accomplish these changes within the regulatory framework.
Consistent with our earlier correspondence, attached please find summaries of the changes and safety evaluations for each ctartup test program change in accordance with 10CFR50.59(b).
The firct impact on the Startup Test Program will be in Test Condition 3, currently expected to ctart around December 1.
He are submitting these summarion'well in advance of the date required by the Fermi 2 License (Condition 2.C (14) to allow the NRC the earlient possible participation in the process and avoid impacting the progrecs of the $tartup Test Program.
Summary descriptions of changes to the following startup tests are attached.
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VP-86-0141 Page 2 1.
Startup Test 1 -(FSAR Section 14.1.4.8.1.) " Chemical and Radiochemical" 2.
Startup Test 16 -(FSAR Section 14.1.4.8.16) " Core Power Distribution-TIP Une'ar tainty" 3.
Startup fest 19 - (FSAR Section 14.1.4.8.19) " Core Power-Void Mode Response" 4.
Startup Test 20 -(FSAR Section 14.1.4.8.20) " Pressure Regulator" 5.
Startup Test 21 -(FSAR Section 14.1.4.8.21) "Feeduater System" 6.
Startup Test 22 -(FSAR Section 14.1.4.8.22) " Turbine Valve Surveillance" 7.
Startup Test 28 -(FSAR Section 14.1.4.8.28) " Recirculation System-One Pump Trip"
- 8. Startup Test 28 -(FSAR Section 14.1.4.8.28) " Recirculation System-Recirculation Pump Runback" An approved safety evaluation for each of these changes is on file at Fermi 2.
The Onsite Review organisation has reviewed and concurred with the conclusions of each safety evaluation.
Also, General Electric concurs with making these changes to the Startup Test Program (see ).
In addition to the changes listed above, we hope to take credit for any unplanned MSIV full isolation or turbine / generator trip, or recirculation runback to satisfy the intent of one or all of the following startup tests:
1)
Startup Test 23-(FSAR Section 14.1.4.8.23) "!!SIV-Full Isolation" 2)
Startup Test 25-(FSAR Section 14.1.4.8.25) " Turbine Stop Valve and Control Valve Fast Closure Trips"
Mr. James G. Keppler October 17, 1986 VP-86-0141 Page 3 3)
Startup Test 28-(FSAR Section 14.1.4.8.28)
" Recirculation System-Recirculation Pump Runback" As we discussed at the September 24 meeting, should we have an unplanned trip from the appropriate reactor and system conditions for which we acquire adequate data, we plan to take credit for this situation in accordance with 10CFR50.59 and advise you as required by Fermi 2 License Condition 2.C (14).
If you have any questions, please contact Mr. Rob Woolley at (313) 58G-4211.
Sincerely, hC
[
i Attachments cc:
Mr. J. J. Stefano Mr.
W.
G.
Rogers Mr. G. C. Wright Mr. J. M. Taylor, Director Office of Inspection and Enforcement USNRC Document Control Desk Uashington, D.C.
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Startup Test 1-Chemical and RMiochemical Performed in Test Conditions 3 and 6 2
Sumrary Deecrintien of Chance Fermi'2 FSAR Section 14.1.4.8.1 doccribes the Chemical and Padiochemical Startup Test (STUT.000.001) which will be used to secure information pertaining to the reactor coolant and determine that the sanpling equipment, procedures and analytic techniques are Mequate to supply the data required to demonstrate the chemistry of all parts of the entire reactor system ncet specification and process requirements.
Part of SIUT.000.001 requires the Rex: tor Water Cleanup System (RG) to be taken out of service to demonstrate the integrated performance of the PHCU and condensate domineralizer (CD) Systems. The no-nG test was to be performed at Test Conditions 3 and 6.
InsteM of testing these two systems at Test Conditions 3 and 6, we will perform the no-clean up portion of STUT.000.001 at Test Condition 6 only.
Performance of this test at Test Condition 3 provides an opportunity to verify procedures and to obtain preliminary data of the MG and CD systers, however, this is not a prerequisite test for Test Condition 6.
Testing at Test Condition 6 will demonstrate the ability of these systems to adequately manage coolant chemistry at the most demanding plant operating condition.
Changing STUT.000.001 to perform the no-cleanup test only in Test Condition 6, constitutas a procolure change for which we performed a 10CFR50.59 Safety Evaluation. This procedure does not change any of the wording in FSMt Section 14.1.4.8.1.
Summarv of Safety Pvaluation In cocpliance with 10CFR50.59, we have evaluated the procedure change of eliminating the nG and CD systems (no-clean test) portions of SIVP.000.001 in Test Condition 3 aM will perform it in Test Condition 6.
The results of this evaluation are as follows:
a) This change does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment, previously evaluated in the FSAR becauce no chmges are rtade e
to the evaluated design and because the no-cleanup test at Test Condition 6 provides the required data to asscos the chemical control performance of the RG aM condensate domineralizer systems. Testing at Test Condition 6 will demonstrate the ability of theco systems to adequately manage coolant chemistry at the most demanding plant operating
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Startup Test 1 Page 2 b) This change does not create the possibility of an tecident or malfunction of a type different from any evaluated previously in the FSAR because testing at Test Condition 3 is not required in order to operate at higher power levels or to be able to effectively perform the test at Test Condition 6.
c) This change does not reduce the margin of safety as defined in the basis for any Technical Specifications because no chmges are made to the evaluated design and conplete testing of the I5CU and condensate domineralizer systems is performed at Test Condition 6.
Conclusion Based on the above justification and also the fact that Limerick, IIope Crock and Clinton plants provided similar justification in which the hTC accepted and consequently deleted the no-cleanup test at Test Condition 3.
Therefore we conclude that this will not adversely affect any safety systems or the safe coeration of the plant since conplete testing in performed in Test Condition 6 and does not involve an unreviewa3 safety question.
Startup Tect 16 Core Power Distribution-Tip Uncertainty Performed in Test Conditions 3 and 6 4
Sumrary Descrintion of Chance Fermi 2 FSAR Section 14.1.4.8.16 concerns the " Core Power Distritxition" and deceribes the Transversing Incore Probo (TIP) Startup Test (S'IUT.000.018). This test is to confirm the reproclucibility of the TIP System readings at approximately 50% and 75% power levels. The test criteria states that Total TIP uncertainty chall be less than 6%.
S'IUT.000.018 in being deleted in its entirety. Pcsults from previous plant startups have shown that the measured Total TIP asymnetry has always been well below the acceptance criteria of 6%. Detailed analysis of 45 TIP ccts from eight plants for power levels ranging from 18% to 100% and core flow from 33% to 105% showed that the average Total TIP uncertainty wac 3.8%.
Data from more rccent plcnt startups shows that the average value of the Total TIP uncertainty to bc lena than or equal to 2.17%. Furthermore, a statistical analycic of the more recent data shows that there is a 99.7% confidence that 99.7% of the Total TIP uncertaintics will be loss than the acceptance critoria of 6%.
Sunrarv of Safety Fvaluation In corpliance with 10CFR50.59 we have evaluated the deletion of S'IUT.000.018 " Core Power Distribution". The results of this evaluation are as follows:
a)
This chango doca not increace the probability of occurrence or the conacquences of an accident of malfunction of equipment, previoucly evaluated in the FSAR becauco no changen are made to the evaluated decign and TIP system i
operability in demonstrated during power ascension tenting when the procccc conputer undergoes the dynamic system tect care at Tent Condition 2.
Previous plant startupa chow that the meacured Total TIP asyntotry has always bcon well below the acceptance criterion, Gt.
In addition, if the total TIP uncertainty in lenc than the annured uncertainty of 6%, thu margin of cafety is increased.
b)
Thin change doen not create the pocalbility of an accident or malfunction of a type different from any cycluated previously in the FSAR becauco no changen are made to the evaluated design and TIP cystem crorability in demonctrated during power ancension testing when the proccan conputer undcrgoco the dynamic syntem tect cacc.
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Startup Test 16 Page 2 c)
This change does not reduce the rargin of safety as defined in the basis for any Technical Specifications because previcus plant startups at other plants show that the Total TIP asymmetry has always been well below the acceptance criterion, 6t. Further, a statistical analysis of this data indicates that there is 99.7% confidence that 99.7% of the TIP uncertainties will be less than 6%. In addition, if the Total TIP uncertainty is less than the assumed uncertainty of 6%, the margin of safety is increased.
Conclucion Dased on the test results of other power plants ad other tests at Fermi 2, we expect our TIP System to be ruch less than the 6%
acceptance value listed in the Fermi 2 FSAR. In view of the above information and also the fact that Ilope Crock, Clinton and River Bend plants provided similar information in which the ICC accepted and consequently deleted this test, we conclude that deletion of UTUT.000.010 does not adversely affect any safety related systers or the safe operation of the plant and does not involve an unreviewed safety question.
A Startup Test 19-Core Power-Void l' ode Response Performed in Test Conditions 4 and 5 Furrary Deecription of Chancie Fermi 2 FSAR Section 14.1.4.8.19, " Core Power-Void Mode Ecoponse" Startup Tect (STUT.000.021) measures the stability of the reactor core power-void dynamic responce and demonctrates that the cores behavior is within its specifial limits. To accomplish this, the stability is recasured by attenpting to make rapid reactivity changes por S'IUT.000.021. Since the core power-void modo resulting from a conbination of the neutron kinetics and core-thermal hydraulics is least stable near the natural circulation end of the 100% roi line this tect was to be perforned during Test Conditions 4 and 5.
Daced on the following, SnTr.000.021 is being deleted in its entirety.
Moncurennt of cyctem stability by movement of control rods were developal for small reactor corcs. Uce of thic technicue for largo loosely coupled DWPs, typical of current plants in the ctartup testing phase, will not provido significant information on the stability of the cyctem becauce moving a control rod providea only a very small recponne in a localized arca. For large BWR cores (i.e., Formi 2), control rod worths in the power /ficu range of interect are nuch lecc than for a small tightly coupled core. Instead, core wide disturbanccc provido more meaningful data for largo corca as is done in MUr.050.022, (FSAR i
Section 14.1.4.0.20), " Pressure Fogulator" testing, which measures the cyctem recponse to pressure disturbances caused by actions of the proccure regulator cyctem.
In addition, extensive special testing of ctability characteristics has been performd at coveral EURc, including Vermont Yankee, Caorno, Leibstadt, Peach Dottom-2 and Brown's Ferry. The test data has demonstrated the stability characteristics of BUR 3 over a wide range of conditions and has been reviewed along with c::toncivo supporting analyces, as part of the ITC's Safety Evaluation Ecport on core thermal-hydraulic stability Surrary of Safelv rvaluation In conpliance with 10CFR50.59, we have evaluated the dolction of MUT.000.021. The recultn of thin evaluation are as follows:
a)
Thic chcngo docc not increaco the probability of occurrence or the conccquences of cn accident or rcalfunction of equipment, previously evaluated in the FSAR becauco no changen are rode to the evaluated design and becauno cyctem ntability in adequately macurni dueing mur.050.022, "Prennuro Pogulator". M Ur.050.022 meccurec tbc syntem recponne to coro wide dicturbancen cauced by preccuro I
dinturbancen which provir'c maningful data for largo corea (i.e., Fermi 2). Converacly, SnTP.000.021, " Core Power-Void l'cr!c necponco", measuren the nyctem etability by novoment of control roda which won developed for crall tightly coupled reactor corce m1 when applico to large ]cocely couplul DUna I
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Startup Test 19 Page 2 will not provide significant information on the stability of
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the system because of the low signal-to-noise ratio.
Furthermore, as a result of extensive testing and analysis of core thermal hydraulic stability over a wide range of conditions, it has been demonstrated that General Electric BWR fuel and core designa meet the stability criteria set forth in General Design Criteria 10 and 12 of 10CFR50.
b)
This change does not create the possibility of an accident or malfunction of a type different from rny evaluated previously in the PSAR because no changes are m:de to the evaluated decign and the system stability is adequately measured during SWT.050.022, "Precrure Pogulator". SWT.050.022 neasures the system recponse to core wide disturbanceo caused by pressure disturbances which provide meaningful data for large coren (i.e., Fermi 2). Conversely, SIUP.000.021, " Core Power-Void riodo Responce", neasures the system stability by movement of control rods which was developed for cmall tightly coupled reactor corec and when applied to large loosely coupled BWRc will not provide nignificant inforration on the stability of the system becauce of the low signal-to-noise ratio. Furthermore, as a result of extencive tocting and analysis of core thermal hydraulic stability over a wide range of conditions, it has been demonstrated that General Electric BWR fuel and core decigns meet the stability critoria cet forth in General Design Criteria 10 and 12 of 10CFR50.
c)
This change does not reduce the margin of safety an defincd in the basis for any Technical Specificationn becauco no changen are made to the evaluated design. In addition, information on the cyctem'n etability is provided by surveillance proceduren and Technical Fpecification 3.4.1.1 recomrnndations for the nonitoring of neutron. flux. Adequate deconstration of acceptable renponce of the core power-void modo is provided during STr.050.022 and by surveillance proccdurec.
Conclusion An a recult of the extensive tecting and analysin of core therral hydraulic ctability, it han been dononntrated that General Electric DUR fuel and core designs rcet the stability criteria set forth in General Decign Criteria 10 and 12 of 10CFR50, Appendix A.
Daned on the above discunnion, the Staff'n Safety Evaluation Report (letter, C. O. Thoran to !!. C. Pfefferlen (GE) acceptance for referencing of Licenning Topical Report imE-240ll, Rev. G, AroMront 8, Thermal !!ydraulic Stability Anendront to CFsrAR II; April 24, 1985), and the fact that Itope Creek, Clinton, !!ine !!ile Point and River Dend planta have provided nimilar juctification in which the imC accepted and concequently deleted thin tent.
Startup Test 19 Page 3 Ue conclude that the deletion of swr.000.021 will not adversely affect any safety related systems or safe operation of the plant arri does not involve an unreviewed safety question. System stability is adequately measured during SMr.050.022, " Pressure Regulator", and has been extensively tested at several Bims covering a wide range of designs.
Therefore, swr.000.021, " Core Power-Void tiode Response" can be deleted from the Power Ascension Test Program.
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Startup Test 20-Pressure Regulator Performed in Test Conditions 4 md 5 Surrorv Dencrintion of Chance Fermi 2 FSAR Section 14.1.4.8.20, "Prescure Regulator" Startup Test (S'IUT.000.022) determines the optinum cettings for the pressure control loop by analysis of the transients induced in the reactor pressure controls system, demonstrates the takeover capability of the backup pressure regulator on failure of the controlling pressure regulator and to set spacing between the cetpoints and the appropriate value, demonstrates smooth pressure control transition between the control valves and bypass valvec when the reactor generates more steam than used by the turbine.
We are deleting the pressure regulator test (STUT.040.022) at Test Condition 4 and the backup preocure regulator takeover testing (CIUT.050.022) at Test Condition 5.
The justification for deleting these steps are as follows:
The precoure regulator cyctem is prirarily censitive to vessel steam flow (and hence power level) since the reactor is basically operated at a constant proccure cetpoint for varying steam flous. Therefore, testing of the pressure regulator recponce should cover the range of expected core power levels and is not significantly dependent on core flow since the steam flow at a fixed power level is incensitive to the core flow rate. Testing of the pressure regulator cystem during Tect Condition 2, 3, 5 and 6 adequately covers the range of expected power levels during plant operation. Therefore, tocting of the preocure regulator system at Test Condition 4 is not required for verification of the controller performance and testing at Tect Conditions 2, 3, 5 and 6 will provide adequate confirmation of the system performance over the entire operating range.
Precoure regulator tocting (specifically the pressure cetpoint chmges) at Tect Condition 4 also providen inforration on the stability of the cystem. Houover, information on the stability of the reactor at Test Condition 4 can also be obtained by ronitoring the neutron flux (both local and core average). Thece curveillance recommendations provide for monitoring of the APm and lpm detectors when operating at natural circulation conditions and provide cufficient information on the stability of the reactor at Tent Condition 4 in addition to providing operator training for the monitoring procedures. Preocure regulator testing (preccuro cetpoint changec) will be performed at Test Condition 5, which bounds the least stable portion of the normal operating region, and will provide 03ditional information on the stability of the reactor. Thercfore, preccure regulator testing at Tent Condition 4 can be deleted.
Tecting of the backup preccure regulator is performed by cinulating the failure of a colected preocure regulator. Thic tect is currently planned to be performed at Tect conditionc 2, 3, 5 and 6.
Tecting at Tect Condition 12, 3 and 6 provides adequate demonstration of the capability of the backup pressure regulator to control precoure in the
Startup Test 20 Page 2 event of a failure of the controlling pressure regulator since these test conditions bound the power level of Test Condition 5.
Sumary of Safety Evaluation In accordance with 10CFR50.59, we have evaluated the deletion of the two steps of S'IUP.000.022. The results of this evaluation are as follows:
a)
This change does not increase the probability of occurrence or the consequences of an accident or malfunction of equipnent, previously evaluated in the FSAR because no changes are made to the evaluated design. Testing of the Pressure Regulator system at Test Condition 4 is not required since the pressure regulator will be tested at other test conditions that bound the power level of Test Condition 4.
Stability data during Test Condition 4 will be obtained by monitoring of the APHi and LPHI detectors. Technical Specification 3.4.1.1 and Surveillance Procedure 54.000.16 and testing at Test Conditions 2, 3, and 6 provides adequate demonstration of the capability of the backup pressure regulator to control pressure in the event of a failure of the controlling preocure regulator since these test conditions bound the power level of Test Condition 5.
b)
This change does not create the poss!bility of an accident or malfunction of a type different from any evaluated previcusly in the FSAR because no changes are made to the evaluated design and adequate demonstration of tLo Trformance of the Pressure Regulator system and backup prescure regulator is provided at other test conditions.
c)
This change does not reduce the margin of safety as defined in the basis for any Technical Specifications because no changes are made to the evaluated design and testing at the other test condition will bound the power levels of the deleteA teet conditions.
Conclusion Testing of the Pressure Regulator system at Test Condition 4 is not required since the pressure regulator will be tested at other Test Conditions that bound the power level of Test Condition 4 and stability data during Test Condition 4 can be obtained by nonitoring of the APRI and LPRI detectors as recorxnended by GE SIL-380, Rev 1.
In addition, testing of the backup pressure regulator by simulating a failure of the controlling pressure regulation at Test Conditions 2, 3 and 6 adequately demonstrates the performance of the backup system.
Based on the above justifications and the fact that Hope Creek Clinton, Nine Mile Point, and River Ecnd plants have provided similar justifications in which the NPC accepted and deleted tests, we
Startup Test 20 Page 3 conclude that deleting the. failure to the bcckup pressure regulator test portion of the pressure regulator testing at Test Condition 5 and deletion of all pressure regulator testing in Test Condition 4 does not adversely affect any safety systems or the safe operation of the plant and therefore, does not involve an unreviewed safety question.
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Startup Test 21-Feedwater System Performed in Test Condition 4 Sumary Descrintion of Chance Fermi 2 FSAR Section 14.1.4.8.21, "Feedwater System" Startup Test (S'IUT.000.023) verifies that the feedwater system has been adjusted to provide acceptable reactor water level control. Nominal water level setpoint changes are used to evaluate the feedwater control system settings for all power and feedwater purp modes. Performance of this test is currently planned for Test Conditions 2-6.
Based on the following discussion,nthe water level setpoint tests (SIUT.04B.023) are being deleted. for Test Conditions 4.
Acceptance criteria define acceptable performance of the feedwater control system to testing pertubations. Criteria require that the transient response of any level control system-related variable to any test input must be non-divergent and also specify response characteristics for given disturbances. Testing during the power accension program is performed at Test Corditions 2-6 to demonstrate co:tpliance to these criteria.
The feodwater control system raintains the rass balance of the reactor vessel by supplying water to the vessel to match the steam flow exiting the vessel, thereby raintaining a constant water level during normal operation. Therefore, the feedwater control system is prirarily dependent on the vessel steam flow and hence the reactor pouer.
Testing of the feedwer control system at Te.st Conditions 2, 3, 5 and 6 adequately bounds ta cxpected power levels for system operation.
Since the power level of Test Condition 4 is similar to that of Test Condition 5, the feo3 water control system performance at Test Condition 4 is not expected to be significantly different than at Test Condition 5.
Therefore, testing at Test Conditions 2, 3, 5 and 6 will a3cquately confirm the system perforrance over the entire operating range.
Sumary of Safety Evaluation In conpliance with 10CFR50.59, we have evaluated the deletion of Test Condition 4 for STUT.04B.023. The results of this evaluation are as follows:
a)
This change does not increase the probability of cccurrence or the consequences of an accident or ralfunction of equipment, previously evaluated in the FSAR because no changes are rade to the evaluated design. Testing of the feedwater control system at Test Condition 4 is not required since the feedwater system will be adequately tested at other test conditions that bound the power level of Test Con 3ition 4.
Startup Test 21 Page 2 b)
This change does not create the possibility of an tecident or ralfunction of a type different from any evaluated previously in the FSAR because no changes are maSe to the evaluated design and adequate demonstration of the perforrance of the feedwater control system is provided at other test conditions.
c)
This change does not reduce the rargin of safety as defined in the basis for any Technical Specifications because no changes are rade to the evaluated design and testing at the other test conditions will bound the power levels of the deleted test conditions.
Conclusion Testing of the feedwater control system at Test Conditions 2, 3, 5 and 6 provides demonstration of systera perforrr.nce over the entire i
operating range. Based on the above justification and the fact that the Clinton and Nine Mile Point plants have submitted similar justification in which the UFC has accepted and consequently deleted Test Condition 4 testings, we conclude that deletion of Test Condition 4 testing does not adversely affect any safety systems or the safe operation of the plant and as such doce not involve an unreviewed safety questicn.
Startup Test 22 'Ibrbine Valve Surveillance Performed in Test Condition 3 Surrarary of Description of Chances Fermi 2 PSAR Section 14.1.8.22 " Turbine Valve Surveillance" Startup Test (SWP.000.024) requires the demonstration of acceptable proccdures and maxinum power levels for surveillance testing of the rain turbine control and stop valve without producing a reactor scram.
Based on the following justification, we are deleting the perforrance of STUT.033.024 for Test Condition 3 only. Individual main turbine control, stop and bypass valves are tested routinely during plant operation as required for turbine surveillance testing. At several test points, the response of the reactor is determined by analyzing test data and conparing the results to acceptance criteria which define the limiting system performance. The criteria require that peak values for system variables (neutron flux, heat flux, vessel pressure and steam flow) do not exceed prescribed limits.
The purpose of the S mr.000.024 tests are to determine the raxirum power level at which the curveillance can be performed with ample margin in order not to scram the plant by the actuation of turbine valves at various test points at selected power levels which are then used to extrapolate to the maxinum power test condition. The operability of the turbine control valves and turbine bypass valves is demonstrated during periodic surveillance testing of the turbine valves required by current Technical Specification curveillance requirements 3/4.3.8. The necessary S mr.000.024 testing to determine the maxirum power level at which the turbine valve surveillance test can be performed will be started at Test Condition 5 and extended up through Test Condition 6, which bounds the testing which would be performed at Test Condition 3.
Summarv of Safety Evaluation In corpliance with 10CFR50.59, we have evaluated the deletion of Test Condition 3 for SWP.03B.024. The results of this evaluation are as follows:
a)
This change does not increase the probability of occurrence or the consequences of an cccident or malfunction of equipment, previously evaluated in the FSAR because no changes are made to the evaluated design. Functional testing of the turbine valves is performed during routine surveillance testing required by Technical Specification surveillance requirements 3/4.3.8. The necessary testing to determine the maxirum power level at which the turbine valve surveillance test can be performed will be startcd at Test Condition 5 and extended up through Test Condition 6, which bounds the testing which would be performed at Test Condition 3.
b)
This change does not create the possibility of an accident or malfunction of a type different from any evaluated previously
Startup Test 22 Page 2 in the FSAR because the testing done during the search for the maxirum surveillance power level is performo3 between Test Condition 5 and Test Condition 6 which bounds the testing which wouM be performed at Test Condition 3.
c)
This change does not reduce the margin of safety as defined in the basis for any Technical Specifications because no changes are made to the evaluated design, and turbine valves operability is adequately demonstratcd by the performance of the Technical Specification surveillance tests, which are routinely performed at the intervals and operating conditions specified within the Technical Specifications.
Conclusion Functional testing of the turbine valves is adequately perforred through the S WT.000.024 testing involving the search for the maximum power level at which the turbine valve curveillance test can be perforrced in conjunction with the routine surveillance testing required by Technical Specification surveillance requirements. Based on the above justification and the fact that the Clinton Plant submitted similar justification in which the IEC accepted and consequently deleted this Test Condition, we conclude that this change will not aSversely affect any safety systems or the safe operation of the plant, and does not involve an unreviewed safety question.
Startup Test 28-Recirculation System-One Purp Trip Performcd in Test Condition 3 Summary Descrintion of Chance Fermi 2 FSAR Section 14.1.4.8.28 Item a " Recirculation System-One Purp Trip" Startup Test (S'IUT.000.030) is intended to obtain recirculation system performance during the purp trip and punp restart and to ver ify feedwater control system capability to control water level without a resulting high water level scram. The current program has a single recirculation pung trip at both Test Condition 3 and 6.
The recirculation punp trip for Test Condition 3 of S'IUT.03B.030 is being deleted based on the following:
Response of the system during a single purp trip is determined by analyzing test data and conparing to acceptance criteria which define the required system performance. For the single purp trip test, the reactor unter level, simulated heat flux and APm single must have margin to scram setpoints. These criteria are applied to provide assurance that the system can avoid a scram during a single purp trip to inprove plant availability. Testing of a single pmp trip at Test Condition 6 uill be performed to demonstrate the system perfornance during a single purp trip. The testing is not required to verify FSAR transient analysis since it has been demonstrated by previous DWR/4 startup testing and analytically (FSAR) that the single purp trip has a negligible inpact on safety limits. In addition, the characteristics of recirculation punp trips are well understood and have been demonstrated during power ascension testing at previous plants.
Summary of Safety Evaluation In conpliance with 10CFR50.59, we have evaluated the deletion of Test Condition 3 from S'IUr.03B.030. The results of this evaluation are based on the following:
a)
This change does not increase the probability of cccurrence or the consequences of an tecident or malfunction of equipment, previously evaluated in the FSAR because no changes are made to the evaluated design and testing of a single pumo trip at Test Condition 6 will be performed to demonstrate the system performance during a single purp trip. In addition, testing is not required to verify FSAR transient analysis since it has been demonstrated by previous DWP/4 startup testing and analytically (FSAR) that the single purp trip has a negligible inpact on safety limits.
b)
This change does not create the possibility of an accident or malfunction of a type different from any evaluated previously in the FSAR becauco no changes are node to the evaluated design and testing of a single ptmp trip at Test Condition 6 will be performed to demonstrate the system perforrance during a single purp trip. In cddition, testing is not required to verify FSAR transient analysis since it has been demonstrated by previous BkT/4 startup testing.
Analytically
,', Startup Test 28 Page 2 the FSAR shows that the single pump has a negligible impact on safety limits.
c)
This change does not reduce the margin of safety as defined in the basis for any Technical Specifications because the tenperature data required by SIUT.000.016, Selectcrl Process Terperatures, and the te@erature data that is required by Technical Specification Section 3/4.4.1.4 will be taken during the Test Condition 6 Recirculation purp trip.
S'IUT.03B.033, Drywell Piping Vibration was intended to verify, in conjunction with the Test Condition 3 pu@ trip, that recirculation piping vibration is within acceptable limits. Previous plant startup results iMicate that vibration and deflection measurements of recirculation piping during recirculaticn pump trips and restarts are always well within the expected range and are significantly belcw the allowable range and therefore, this testing can be deferred until it can be obtained during the single purp trip at Test Condition 6.
Conclusion Performance of single purp trips are only required as demonstration of the ability of the feedwater control system to inprove plant availability by avoiding a high water level scram during the single pump trip and is not required to verify FSAR transient analysis. Data for SIUT.000.016, Selected Process Temperatures can be obtained during the single pump trip from Test Condition 6.
In Mdition, previous plant startup results have demonstrated that the vibration and defection measurements during recirculation pu@ trips and restarts for plants similar to Fermi 2 are well below prescribed limits and this data will be adequately verified by making these measurements during the single pimp trip at Test Condition 6.
Based on the above justification and the fact that the Clinton and River Bend plants have submitted similar justifications in which the hTC accepted and consequently deleting this Test Condition, we conclude that deleting the single pump trip at Test Condition 3 will not adversely affect any safety related systems or safe operation of the plant and therefore does not involve an unreviewed safety question.
Startup Tests 28-Recirculation System-Fecirculation Punp Runback Performed in Test Condition 3 Summary Deccription of Chance Fermi 2 FSAR Section 14.1.4.8.28 Item d.
" Recirculation System-Rocirculation Punp Funback" Startup Test (S7UT.03D.030) verifies that no recirculation system cavitation occurs in the operable region of the power-flow map.
4 One function of the recirculation flow control system is to provide a recirculation flow runback upon the coincident loss of one fedwater punp and low water level alarm (Level 4) indication to avoid scram on low water level (Level 3). S7UT.03D.030, Recirculation Rur.back, sinulates a loss of a feedwater punp at Test Condition 3 near rated recirculation flow to determine the adequacy of the recirculation flow runbeck feature in preventing a scram. It is proposed to denonstrate this feature in conjunction with STUT.06D.023, Feed Pump Trip at Test Condition 6 or during an inadvertent feedwater punp trip event.
Based on the following discussion STUT.03D.030 will be performed in conjunction with S7UT.06D.023 " Feed Punp Trip" at Test Condition 6 or during an inadvertent feedwater trip instead of Test CoMition 3 as stated in the Fermi 2 FSAR Table 14.1-1.
The response of the system during a feedwater pump trip with recirculation runback is determincd by analyzing test data and conparing to acceptance criteria which define the required system performance. For the recirculation flow runback test, the recirculation flow is required to runback upon a trip of the runback circuit. During preoperational testing, the runback logic circuits were functionally tested to verify that they are according to design. Although the purpose of SrUT.03D.030 in to determine the adequacy of the recirculation flow runback feature to prevent scram following a feedwater punp trip, this is not demonstrated since a feedwater punp trip is not initiated. During STUT.06D.023 Feed Punp Trip at Test Condition 6, a recirculation flow runback would occur as the result of the feedwater punp trip coincident with a lov water level (Level 4) indication. Alternatively, an inadvertent fecdwater punp trip event with low water level vould provide deronstration of the recirculation runback feature. These events would result in an actual demonstration of the recirculation runback circuit under real, as opposed, to sinulated conditions.
Surr'arv of Safety Evaluation In conpliance with 10CFR50.59, we have evaluated the combination of STUr.03D.030 and S7UT.06D.023 at Test Condition 6.
The results of this evaluation are as follows:
a) This change does not increase the probability of occurrence or the consequence of an accident or malfunction of equipnent, previously evaluated in the FSAR because no chmges are made to the evaluated design and because the adequacy of the recirculation f1cu runback feature to prevent a scram following a feedwater punp trip will be demonstrated during SIUT.06D.023 " Feed Punp Trip" at Test Condition 6.
The
Startup Test 28 Page 2 performance of the recirculation runback is currently verified during both swr.03D.030 and SWP.0GD.023. During S WP.03D.030, the loss of a feedwater pump in simulated whereas during swr.06D.023 a feedwater pur:p is actually tripped. With the proposed change, the perforrance of the recirculation runback will still be verified under real, as opoosed to sirulated conditions. In addition, S WP.03D.030 and swr.06D.023 both use the same acceptance criteria to evaluate recirculation runback. Furthermore, the runback logic circuits were functionally tested to verify that they are according to design during preoperational testing, b) This change does not create the possibility of m cccident or malfunction of a type different from any evaluated previously in the FSAR because the purpose of SWP.03D.030 is to verify the adequacy of the recirculation runback to avoid reactor rcram upon loss of one feodwater purp which will still be verified during Test Condition 6 by SWP.06D.023.
c) This change does not reduce the margin of safety as defincd in the basis for any Technical Specifications because the performance of the recirculation runback is verified during S'IUT.0GD.023 and the logic circuits were functionally tested to verify that they are according to design during preoperational testing.
Conclusion Testing of the recirculation flow runback feature during preoperational testing and in conjunction with swr.06D.023 at Test Condition 6 or during an inadvertent feedwater trip will demonstrate actuation of the recirculation runbach circuits and demonstrates its adequacy to prevent a reactor scram. As such deletion of the si::ulated recirculation system runback feature in Test Condition 3 and based on the fact that Lircrick, Hope Creek, Clinton and River Eend plants have submitted rimilar justifications in which the NTC accepted and consequently deleted this test, we conclude that this will not adversely affect any safety related systems or safety operation of the plant and does not involve en unreviewed safety question.