ML19289C409
| ML19289C409 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 01/03/1979 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Jens W DETROIT EDISON CO. |
| References | |
| NUDOCS 7901120148 | |
| Download: ML19289C409 (22) | |
Text
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JAfi 3 1979 Docket No.
50-341 Dr. Wayne H. Jens Assistant Vice President Engineering & Construction The Detroit Edison Company 2000 Second Avenue Detroit, Michigan 48226
Dear Dr. Jens:
SUBJECT:
REQUESTS FOR ADDITIONAL INFORMATION IN FERMI 2 FSAR As a result of our continuing review of the Final Safety Analysis (FSAR) for the Enrico Fermi Atomic Power Plant Unit 2, we have developed the enclosed requests for additional information.
Please amend your FSAR to comply with the requirements listed in the enclosure.
Our review schedule is based on the assumption that the additional information will be available for our review sy February 26, 1979.
If you cannot meet this date, please inform us within 7 days after receipt of this letter so that we may revise our scheduling.
Sincerely, 7
s ohn F. Stolz, Chief ht Water Reactors Branch No. 1 Division of Project Management
Enclosure:
Requests for Additional Information cc w/ enclosure:
See next page 79011201
Dr. Wayne H. Jens JAN 3 1979 cc:
Eugune B. Thomas, Jr., Esq.
Mrs. Martha Drake LeBoeuf, Lanb, Leiby & MacRae 230 Fairview 1757 N. Street, N.W.
Petoskey, Michigan 49770 Washincton, D. C.
20036 Peter 8 Marquardt, Esq.
Co-Counsel The Detroit Edison Company 2000 Second Avenua Detroit, Michigaa 48226 Mr. William J. Fahrner Project Manager - Fermi 2 The Detroit Edison Company 2000 Second Avenue Detroit, Michigan 48226 Larry E. Schuerman Licensing Engineer - Fermi 2 Detroit Edison Company 2000 Second Avenue Detroit Michigan 48226 Charles Bechhoefer, Esq., Chairman Atomic Safety & Licensing Board Panel U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Dr. David R. Schink Department of Oceanography Texas,. & M University College Station, Texas 77840 Mr. Frederick J. Shon Atomic Safety.& Licensing Board Panel U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Mr. David Hiller University of Michigan Law School Hutchins Hall Ann Arbor, Michigan 48109 Mr. Jeffrey A. Alson 772 Green Street, Building 4 Ypsilanti, Michigan 48197
- ~
ENCLOSURE REQUESTS FOR ADDITIONAL INFORMATION ENRICO FERMI ATOMIC POWER PLANT UNIT 2 DOCKET NO. 50-341 Requests by the following branches in NRC are included in this enclosure.
Requests and pages are numbered sequentially with respect to previously transmitted requests.
Branch Page No.
Core Performance Branch - Reactor Fuels Section 241-3 Reactor Systems Branch 212-10 through 212-27 O
241.0 CORE PERFORMANCE BRANCH - REACTOR FUELS SECTION 241.4 Our requirement for routine fuel surveillance is discussed (4.2) in paragraphs I.D and II.D of Section 4.2 (Revision 1) of the Standard Review Plan-Provide a description of the on-line fuel rod failure detection methods and a description of the post-irradiation fuel surveillance program planned for Ferai 2.
241.5 Asymmetric loads from seismic and LOCA events are being (4.2) reviewed by NRC as a generic task (A-2, NUREG-0371) for both PWRs and BWRs. The results of an analysis that show that the fuel assemblies and channel boxes can withstand this phenom-enon and that coolable geometry is maintained should be provided.
This evaluation should be, performed using state-of-the-art methodology and criteria for the Fermi 2 design.
Fermi 2 should further agree to perform a reevaluation should the criteria developed from generic task A-2 warrant such evalua-tion.
o 241-3
~
212-10 21 2.0 REACTOR SYSTEMS BRANCH 212.15 In section 5.5.1.4 it is stated that impeller missiles will (3.5) not penetrate the pump case.
Provide or reference the supporting analysis to justify this statement.
212.16 Provide or reference an analysis showing that the probability (3.5) that impeller missiles ejected from the broken pipe will cause significant damage within containment is acceptably low.
If a study for another plant is referenced, justify that the conclusions are appropriate for Fermi 2.
212.17 Note 7 of Figure 4A.2-21 specifies that lines and equipment (4.6) fromthestoragetagktofR4-2009beheattracedtomaintain a temperature of 80 F + 5 F to prevent precipitation.
How is this accomplished? What instrumentation,is provided to permit the operator to verify that these temperature limits are satisfied? What surveillance testing will be required in the Technical Specifications to Easure that the limitations are met?
Describe how testing can be performed to demonstrate the required flow rate from the SLC pumps.
If testing is performed by recirculating water to the test tank, how is flow rate deter-mined? Figure 4A.2-21 does not show a flow sensor in the test line.
212.18 Fermi 2 will have the new two-stage target rock S/R valves.
(5.2.2)
Describe the tests used to qualify these valves for operation under normal and post-LOCA environmental conditions.
If these tests have not been completed, provide the scheduled completion date.
212.19 The overpressure analysis for Fermi 2 is incomplete. The (5.2.2) analysis should include the effects of the ATWS reactor recirculation pump trip on high reactor pressure and be
+
based on an initial opercting pressure (up to the Technical Specificationlimit) that will result in the most limiting peak pressure.
For the pilot-operated valves used in Fermi 2, credit for 3/4 of the rated capacity is allowed for the safety mode of operation in accordance with ASME B&PV Code Section III, NB-7000. Credit can be taken for the second safety grade signal for reactor scram (typically the neutron flux scram for the case of MSIV closures). Provide a reanalysis of the most limiting overpressure transient taking account of the restrictions noted above.
212.20 On page 5.2-12 it is stated that the safety relief valves open
,(5.2.2) by application of external power to limit a pressur _. rise in addition to the safety mode and ADS mode.
Confirm that the
s 212-11 (5.2.2) sa'fety analyses presented in Chapter 15 and the overpressure analysis presented in section 5.2.2 do not take credit for the power-operated relief mode of these valves.
It is the staff's position that no credit should be allowed for this function because the actuation system does not satisfy safety grade criteria.
212.21 There have been at least eight damaging water hamer occurrences (5.4.6) in the turbine supply or exhaust lines of HPCI or RCIC systems that were attributed to steam driven slugs of water.
Contri-buting causes included a) water drawn into the exhaust line from the supp:ession pool, b) inadequate draining of the steam supply line, and c) trapping of water slugs upstream of the supply line isolation valves during maintenance. The outboard isolation valve typically has a seal-in feature such that the valve opens fully following an open signal. What imorove-ments have been provided in the design of Fermi 2 to minimize the occurrence of damaging water hammers in the turbine supply and exhaust lines of the HPCI and RCIC systems.
212.22 On at least three occasions, leakage of steam past valves in (5.4.7) the steam supply lines to the RHR heat exchangers has resulted in steam bubble formation in the heat exchangers and the occur-rence of damaging water hamer following RHR pump start up.
For Hatch 2, a near-zero leakage valve was added to the existing valves to alleviate this problem. What design features are used in Fermi 2 to prevent this cause of water hammer?
212.23 Operation of the RHR system in the steam condensing mode (5.4.7) involves isolation of the heat exchanger, dropping of the shell water level, venting of-non-condensables and introduction of steam.
Inadvertent opening of one of the valves used to isolate the heat exchanger would result in injection of relatively cold water into steam and the possible occurrence of damaging water hamer.
Is the system design such that water hammer could not occur following valve opening?
If not, what design features are provided to prevent inadvertent valve opening?
212.24 Response to previous request 212.14 demonstrates that shutdown (5.4.7) cooling can meet single failure criteria if the saf ety/ relief valves are utilized. Since the Fermi 2 S/R valves require air pressure to operate in this mode, provide assurance that a qualified air supply is available to operate these valves for the time period when shutdown cooling is required. What testing do you intend to perform to demonstrate that the integrity of this air supply is maintained?
212-12 212.25 Check valves in the discharge side of the HPCI, LPCI/RHF.,
(6.3)
RCIC systems perform an isolation function in that they protect low pressure systems from full reactor pressure.
The staff will require that these check valves be classified ASME IWV-2000 Category AC, with the ieck testing for this class of valve being performed to code specifications.
It should be noted that a testing program vhich simply draws a suction on the low pressure side of the outermost check valves will not be acceptable.
This only verifies that one of the series check valves is fulfilling an isolation function. The necessary testing frequency will be that specified in the ASME Code, except in cases where only one or two check valves separate high to low pressure systems.
In these cases, leak testing will be performed at each refueling after the valves have been exercised.
Identify all ECCS check valves which should be classified Category AC as per the position discussed above. Verify t,at you will meet the required leak testing schedule, and that you have the necessary test lines to leak test each valve. Provide the leak detection criteria that will be proposed for the Technical Specifications.
212.26 Downstream portions of the containment spray lines and the (6.3) spray header are empty. Hence, there may be significant dynamic loads during the initial period after spray initiation as'tiie water colurn traverses bends or tees and encounters changes in flow area. Since the containment spray is net tested with water, there is no experimental verificatio i tnat the-installed system can undergo these dynamic loads wichout sustaining damage to supports and restraints.
Provide che basis -Tor determining t-hat the system can withstand the dynamic loads and reference any supporting test results.
212.27 What provisions are made to protect level instrumentation for (6.3) the condensate storage tank and the lines from this tank leading to the HPCI and RCIC systems from the effects of cold weather?
212.28 The listing of missiles from pressurized equipment does not (3.5) include high pressure gas storage bottles, tanks or acccumulators (e.g., for S/R valves and CRD's). Provide the basis for elimin-ation of these missile sources from consideration.
212.29 How are the hydraulic lines from the CRD hydraulic control units (3.5) to the control rods and other essential portions of the control rod drive system protected from the effects (pipe whip and/or jet impingement) of moderate or high energy line breaks?
-212.30 On numerous occasions at the Hatch and Brunswick facilities, the (6.3)
HPCI and RCIC systems have been rendered inoperable because of isolations caused by ventilation inlet-outlet high differential temperatures.
The events occurred when there was a relatively f
212-13 (6.3) sharp drop in autside temperature. As noted in section
- 5. 2. 7.1. 3.1 and Tables 5.2-7 and 7.6-1, Fermi 2 incorporates this type of HPCI and RCIC isolation. Provide a discussion of the modifications that have been or will be made to prevent inadvertent isolations of this type which affect the availability of the HPCI and RCIC systems.
The trip settings for isolation of the HPCI and RCIC systems due 212.31 (6.3) to high area temperature are given in Table 7.6-1 in terms of degrees above ambient temperature.
It is assumed that this ambient temperature as used in Table 7.6-1 is some constant temperature representative of a typical ambient temperature in a given area and that the trip setting might be say, 90 + 125 = 215 for HPCI. Please confirm or correct this There are two concerns. The first concern is assumption.
that specified setpoint is so low that irradvertent isolation The of the HPCI and RCIC systems might occur when needed.
second concern is the method of specification that would be applied for Femi 2 If the method of specification is that the setting is <215 F for the above example, the setting could 3
be set too low and cause inadvertent isolation when the system Spgcificagionofanallowablerangeofthesetpoint is needed.
Please provide the basis temperature (x F<SP<y I) is preferred.
for determination of the setpoints used for high area temper-ature and provisions to prevent inadvertent isolation of the HPCI and RCIC due to high area temperature signals.
212.32 Some relief valve discharge lines (e.g., for RHR system)
(5.4.7) penetrate primary containment and have outlets below the surface of the suppression pool. Since these lines form part of the primary containment, the concern is that excessive dynamic loads during relief valve actuation may cause line cracking or rupture.
Identify these lines penetrating containment and provide information concerning neasures taken to prevent line damage. Of particular concern in this regard are water slugs in lines discharging ste.am (e.g., RHR heat exchangers). Such water slugs would be drawn up from the suppression pool as the result of low pressures with steam condensation or result from inadequate draining of low points.
212.33 At some BWR installations, the check valves in the turbine (6.3) exhaust lines of the RCIC and HPCI systems which serve a containment isolation function have been damaged as the result of intermittent closure. The intennittent closures arise from flow oscillations in the exhaust lir.e associated with formation and collapse of steam bubbles in the suppression pool. One type of corrective action involved use of a sparger to reduce the oscillatior;s. What design features are used at Fermi 2 to prevent this type of damage?
212-14 212.34 It is stated that if torus water emptied into the basement, (6.3) tnc ECCS suction connection would still remain submerged (s2 3/4 f t. below water level). The inference is that ECCS operation would not be impaired. Justify by reference to tests that vortex formation at the suction connection with this small amount of submergence will not impair operation of the ECCS. Discuss preoperational tests at Fermi 2 to demonstrate that there is no impairment of ECCS operation.
212.35 On page 6.3-12b it is stated that "even if the ECCS were (6.3) initiated with empty lines the systems would still be capable of cooling the core within the guidelines as stated in sub-section 6.3.1."
The pertinent statement in section 6.3.1 referenced here is not obvious.
It is assumed that this means that the ECCS can withstand the dynamic loads resulting from startup with empty lines, since an inadequate fill system design or surveillance procedu"es leading to water hammer following ECCS initiation would not be considered a singic failure in the performance evaluation of section 6.3.3.
For example, in the case of a large break with an injection valve failure leading to loss of all LPCI, the core spray system is available. Water hammer in the discharge lines should not result in loss of this system. Discuss the capability of the ECCS (LPCI, CS and HPCI) to withstand dynamic load; resulting from system initiation with empty discharge lines such that operation is ensured in the event of single failures esulting in the loss of performance of other ECC systems.
212.36 The discussion in section 6.3.2.2.5 of the fill system used to (6.3) prevent water hamer due to empty discharge lines in the RHR and ECC systems is inadequate. Since there have been about fifteen damaging water hamer events resulting from empty discharge lines of core spray and RHR systems, the adequacy of fill systems, including instrumentation and alarms is a matter of concern.
Please respond to the following:
1.
Provide a detailed description of the fill system including instrumentation and alarms with appropriate references to a P&ID.
2.
Level transmitters apparently are not used to detect trapped air bubbles upstream of injection valves. Pressure readings downstream of the pump discharge check valves which are greater than the gravity head corresponding to the highest point in the system do not necessarily indicate the absence of trapped air pockets. What provisions for Fermi 2 are made to avoid inadvertent trapping of air pockets? In the discussion include consideration of leaking valves in bypass test lines.
212-15 (6.3) 3.
If maintenance is required on a particular loop (e.g., in RHRs) requires draining, how does the fill system protect the other loop and systems (e.g., CS)?
4.
What surveillance testing will be required to demonstrate that the fill system instrumentation is capable of perform-ing the desired function?
5.
How are surveillance tests made to determine if the discharge lines for the RHR and CS systems are full as required in the Standard of Technical Specifications.
6.
Assuming the jockey pump system does mot maint:in full lines, water hamer could occur during surveillance tests of the RHR and CS pumps.
If damage occurred, the event would be reported in a LER. However, if special fill and vent procedures were used prior to these tests, water hammer would not occur, but the inadequacies of the jockey pump system might not be evident. Discuss the procedures to be used in surveillance tests involving startup of RHR and CS pumps and the reporting procedures to be used if special filling and venting procedures are used and indicate partially empty lines.
212.37 In describing the LPCI system, the text refers to several (6.3) valves in Figure 6.3-6 that are not identified in the figure.
Valve F010 is shown to be normally closed instead of being key-locked open.
Please amend the FSAR accordingly.
212.38 The ECCS should be designed to provide sufficient capability (6.3) to cool the reactor in the event of any single active or passive failure in the ECCS during the long-term cooling phase following an accident.
Insufficient information is presented in the FSAR to demonstrate that this requirement will be satis-fied with regard to passive failures.
The staff position is that leakage detection and alarms be provided to alert the operator to passive ECCS failures during long-term cooling which allow sufficient time to identify and isolate the faulted ECCS line. The following considerations should be addressed:
1.
Identification and justification of the maximum leak rate should be provided.
2.
Maximum allowable time for corrective operatoi cction should be provided and justified.
s 212-16 Demonstration should be provided that the leak detection (6.3) 3.
system will be sensitive enough to initiate (by alarm) operator action, permit identification of the faulted line, and permit isolation of the line prior to the leak creating undesirable consequences such as flooding of redundant equipment. The minimum time to be considered is 30 minutes.
It should be shown that the leak detection system can 4.
identify the faulted ECCS train and that the leak is isolable.
The leak detection system must meet the follnwing standards:
5.
(a) Control Room Alarm (b)
IEEE-279, except single failure requirements Fermi 2 should determine the effects of ECC5 passive failures such as pump seals, valve seals, and measurement devices. This analysis should address the potential for ECCS flooding and ECCS inoperability that could result from a depletion of suppression pool water inventory. The analysis should include consideration of (1) the flow paths of the radioactive fluid through floor drains, sump dump discharge piping, and the auxiliary building; (2) the operation of the auxiliary systems that would receive this radioactive fluid; (3) the ability of the leakage detection system to detect the passive failure; and (4) the ability of the operator to isolate the ECCS passive failure, including the case of an ECCS suction valve seal failure. Also, examine the auxiliary system piping in the location of ECCS equipment and address the potential of nonsafety-grade pipe tg cause flooding.
The ECCS for Fermi 2 contains manual as well as motor-operated 212.39 (6.3) valves. Consideration must be given to the possibility that manual valves might be left.in the wrong position and remain undetected when an accident occurs. Provide a list of location and type of all manually operated valves in the safety systems and discussion of the methods used for each valve to minimize The staff will require the possibility of such an occurrence.
remote indication in the control room for all ECCS valves (manual or motor-operated).
Recently, at another facility similar to Fermi 2, a potential 212.40 comon mode flooding of ECCS equipment rooms was identified.
(6.3)
The problem involved the equipment drain lines (see IE Circular No. 78-06, May 25,1978). Verify that the specific design for floor and equipment drains for Fermi 2 are such that flooding in any one room or location will not result in flooding If isolation valve 3 of redundant ECCS equipment in other rooms.
or limit switches are used to prevent common flooding, identify these valves and switches and discuss provisions to be included in the Technical Specifications to assure adequate surveillance.
212-17 212.41 It is not evident that the assumed drop of 100 F in feedwater (15A) temperature gives a conservative result of this transient with manualrecirculationflowconfrol. For example, a feedwater temperaturedropofabout150FoccurregatonedomesticBWR.
Also temperature drops of less than 100 F can occur and involve more realistic slow changes with time.
Assuming all combinations result in slow transients with the surface heat flux in equilibrium with the neutron flux at fhe occurrence of scram, a smaller temperature drop than 100 F that still causes scram should result in a smaller MCPR.
Please evaluate this transient and justify that the assumed values of the magnitude and time rate of change in the feedwater temperature are conservative.
212.42 In the evaluation of the generator load rejection transients (ISB) in Appendix 15B, it is stated that a value of 0.15 second full-stroke closure time (from fully open to folly closed) is assumed although a typical closure time is approximately 0.2 second.
In Chapter 10.0 closure times of 0.18 to 0.22 second are noted.
Closure times, from the partially open to fully closed position, are not provided in the Fermi 2 FSAR.
In addition, there are no minimum closure time limits in the Technical Specifications.
For full-stroke closure, the assumed closure time would appear to be conservative in terms of the supplied information. However, for operation in the full arc (full throttling) mode, the closure times may be significantly less than 0.150 second for typical cases where the control valves are only partially open.
With respect to this transient, T.ere are two concerns. The first concern is that minimum closure times for either full or part-stroke in Fermi 2 may be less than those assumed in the analysis. The second concern is that the analysis, which u based on 105% NBR steam flow and valves wide open initial conditions, gives a less conservative analytical result than an initial condition at the same as a somewhat lower power with i
control valves partially open as expected. Demonstrate that control valve closure times smaller than 0.150 second do not result in unacceptable increases in AMCpR and reactor peak pressure or provide either a) justification that smaller closure times cannot occur or b) a minimum closure time to be incorpor-ated in the Technical Specifications.
212.43 Operation of Fermi 2 with partial feedwater heating is not (15A) limited by the Technical Specifications. Such operation might occur during maintenance or as a result of a decision to operate with lower feedwater temperature near end of cycle. Justify that this made of operation will not result in a) greater maximum reactor vessel pressurer than those obtained with the assumptions used in Appendix 5A, or b) a more limiting AMCPR than would be obtained with the assumptions used in section 15.0 and Appendix 15A. The basis for the maximum reduction in feeds ter heating considered in the response should be provided (e.g., specific turbine operational limitations).
212-18 212.44 The flow-rated APRM scram (TPM) which initiates scram for (15.0) some abnormal operational transients is not discussed in the Fermi 2 FSAR.
Provide the following information:
1.
The flow relation for this trip setpoint used in Chapter 15.0 and Appendix 15B analyses.
2.
The flow-related trip setpoint and al1%able values to be given in the Technical Specifications.
3.
The time constants for this trip used in the Chapter 15 and Appendix 15B analyses and to be used in Fermi 2 and in the Technical Specifications for Fermi 2.
4.
The specific transients for which this scram signal initiated scram.
5.
The time constant representative of the fuel rods for the transients listed under (4).
212.45 The recirculation trip (RPT) is used in Appendix 15A analysas.
(15A)
However, no information concerning this trip was found in the FSAR.
Provide or reference sufficient information to permit evaluation of the adequacy of this trip if it is to be used in Fermi 2.
If the RPT will not be used, revise tne affected transients to account for the elimination of the RPT.
212.46 Turbine trip and generator load rejection witnout bypass are (15A) treated as infrequent in Appendix 15B which result in an MCPR less than the safety limit of 1.06, for the specified initial MCPR of 1.24.
This approach is not acceptable. These transients should be analyzed for an initial MCPR such that the MCPR during the transient does not go below the safety limit of 1.06.
In the analysis of inadvertent opening of a safety / relief 212.47 (15.0) valve, it is stated that a plant shutdown should be initiated if the valve can not be clo. sed. How much time does the operator have to initiate plant shutdu n before exceeding Technical Specification limits for suppression pool temperature? What would be the sequence of events, expected results, and safety considerations if plant shutdown is delayed or not manually initiated? What is the effect of having the recirculation flow control in the automatic mode?
212.48 Recent operating experience has shown that less of instrument (15.0) air could result in situations which may have the potential for causing or compounding more serious events. Therefore, an analysis for loss of instrument air should be provided.
This analysis should consider all equipment which is dependent either directly or indirectly on instrument air and evaluate in detail the effects of loss of instrument air.
In addition
212-19 (15.0) to a complete loss of instrument air, consider loss of air to individual components. Discussion should include:
1.
the potential for transients to be more severe than that predicted by the analyses FSAR 15.1.34.
2.
operator action necessary for continued operation or safe shutdown and the potential for more serious events to develop or be caused by such operation or shutdown.
3.
the adequacy of alarms to inform the operator.
Include a discussion of loss of air to individual components such as temperature control valves for reactor building closed loop cooling water.
212.49 The discussion of the Fermi 2 operating m'ap does not clearly (15.0) state the limits you intend to place on reactor operation with respect to reactor recirculation system availability. Does Fermi 2 intend to operate with less than full availability of both recirculation loops? If so, show that such operation has been analyzed for transient and accident crnditions. Discuss startup testing which would verify results of transient analyses for operation with one recirculation loop. What operating restrictions would be placed on the recirculation system and reactor thermal output during single pump operation?
212.50 The transient analysis for loss of all grid connections shows (15.0) main steam line isolation valve (MSIV) closure at 34 seconds, due to loss of condenser vacuum. A concern is that the MSIV's may close at an earlier time in the transient and result in higher system pressures. Apparently, credit is taken for MSIV air accumulator operation since the normal air supply to the MSIV's would trip at the start of this transient.
Discuss design provisions and verification testing which demonstrate that MSIV performance is qualified to the extent assumed in the analysis.
Related to the same potential for faster MSIV closures, is the Fermi 2 design such that a loss of all grid connections may result in an isolation signal which would close the MSIV's?
What sources of electrical power are used for MSIV isolation logic and isolation actuators? Would these sources of power be available following a loss of all grid connections? Do the logic and actuators fail safe to cause a MSIV isolation signal on loss of electrical power?
,, 212.51 For the transient, with turbine trip initiated at 0 second, the (15.B) bypass valves should be open at about h second and be closed at about 5b seconds. However, Figure 15B.2.5-1 indicates closure of the bypass valves at about 12 seconds. The pressure relief valve actuations (Curve 4) also appear inconsistent with the pressure traces.
Is predicted vessel pressure (Curve 1)
212-20 (15.B) or. steam line pressure (Curve 2) used in determining actuation of the Target Rock values which, it assumed, are opening only in the safety mode. The curves and identifying symbols are difficult to decipher in Figure 15B.2.5-1 in the FSAR. Verify that Curve 4 (relief) refers to the flow rate of Target Rock valves in the safety mode of operation and that Curve 5 (safety valve flow rate) is zero.
212.52 Figure 15B.5.1-1 for the inadvertent HPCI startup is (ISB) inconsistent with the text.
Please correct this inconsistency.
In addition, the surface heat flux, average fuel temperature and vessel steam flow are displaced. For example, surface heat flux and neutron flux should be equal at the start of the transient and have about the same value with time as the neutron flux levels out as indicated. The assumption that the HPCI temperature is 40"F does not appear to be* conservative if the text description of the course of this transient is correct.
A higher HPCI temperature could result in a level 8 trip of the turbine at neutron flut j W below scram setpoint, with a resultant lower MCPR thaa that obtWed using the 40 F value.
Provide a reanalysis usin; more conservative temperatures or justify present result;.
212.53 The following questions rafer to the feedwater controller failure (ISB) events:
1.
Figure 15B.1.2-1, which shows a maximum feedwater flow cf about 150 percent, is not consistent with the text which states that the trarsient was simulated by programming a maxi:,um of 135 perc(nt.
Clarify this descrepancy.
2.
This event is analy::ed for a low initial power (s70%).
Demonstrate why this initial powe> gives a lower value of MCPR than lower or higher (up to that corresponding to 105% NBR) initi41 powers.
3.
With sudden increate in feedwater flow, there will be a drop in the feedwater temperature which contributes to the reactivity increase during the first part of the transient. For example, the combination of feedwater temperature drop and a smaller maximum flow rate could lead to a level 8 trip with the surface heat flux close to the flux scram setpoint.
If the feedwater temperature at the reactor vessel has been assumed constant, the transient should be analyzed to include the effect of this temperature variation on MCPR. The basis for determining the time variation in FW temperature at the reactor vessel should be provided. Justify why a smaller increase in feed-water flow rate in conjunction with the change in feedwater temperature does not give a lower MCPR.
212-21 (ISB) 4.
If this transient gives the limiting change in MCPR, the assumed maximum feedwater flow rate should be verified in the preoperational test program or the analysis snould demonstrate that such a test is not needed to assure the safety limit will not be exceeded.
212.54 The turbine building for Fermi 2 does not conform to seismic (15.0)
Category 1 requirements.
Evaluate the consequences of an SSE which initiates a turbine trip or generator load rejection (whichever is limiting) with loss of trip signals from the turbine building, turbine bypass and other non-safety grade equipment that could mitigate the consequences of the accident.
Assume the most limiting single failure for safety grade equipment that mitigates consequences of the accident and provide the percent of fuel rods undergoing boiling transition. The fuel failure should be based on all rods that are calculated to violate the safety limit, not a statistical combination.
212.55 Figure 158.0-1 is not consistent with the text of section (15.0) 15B.0.3.3.0.
Provide the correct fioure and modifications to the text as required.
212.56 Provide results of an analysis to demonstrate that no sinale (6.3) failure will result in overpressurization of the RHR system.
Provide the design basis used to determine the capacity of the relief valves of the RHR system.
212.57 In section 5.5 of the FSAR, it is stated that the RCIC system (5.5) is designed to meet seismic Category 1 requirements. However, the suction of the RCIC pump is normally aligned to the condensate storage tank which does not meet seismic Category 1 requirements.
The HPCI system suction, which is also normally aligned to the condensate storage tank, is automatically transferred to the suppression pool if a low level in the condensate storage tank is reached. However, realignment of the RCIC suction requires manual action. In the event of an SSE with concurrent loss of offsite power, sufficient time may not be available to pennit credit for such manual actions to prevent unacceptable consequences.
Provide an analysis of the consequences of an SSE with concurrent loss of offsite power to demonstrate that the use of manual actions after 10 minutes to accomplish switchover of the RCIC suction line is acceptable or else modify the system by providing an acceptable water source (seismic Category 1) or automatic transfer to the suppression pool. The analysis should be based
212-22 (5.5) on the assumption that the condensate storage tank and other components and system which are not seismic Category 1 are not available to mitigate the consequences of the postulated event and should include the effects of the worst single failure.
212.58 During long-term cocaing following a small LOCA, the operator (6.3) must control primary system pressure to preclude over-pressurizing the pressure vessel after it has been cooled off.
1.
Describe the instructions given the operator to perform long-term cooling.
2.
Indicate and justify the time frame for performing the required action.
~
3.
List the instrumentation and components needed to perform this action and confirm that these components meet safety grade standards.
4.
Discuss the safety concerns during this period and the design margins available.
5.
Provide temperature, pressure, and RCS inventory graphs that would show the important features during this period.
The above discussion should account for the following:
1.
2.
Operator error or single failure.
212.59 Describe the consequences following the loss of component (5.5) cooling water to both of the reactor recirculation pumps and demonstrate that no consequences important to safety result.
Include a discussion of the effect that the loss of component cooling water has on the recirculation pumps seals.
If additional sour ces of water are available to the seals show that these sources are independent of the componer.t cooling water supply.
212.60 Provide information demonstrating that loss of the operating (Appendix CRD pump at low reactor pressure will not result in accumulator 4A) deoressurization and loss of scram capability. What operator actions are iequired if a CRD pump fails or a drwe water filter becomes plugged? How much time does the operator have before loss of scram capability is expected. The staff requires that you provide a plan for periodic testing of the scram accumulators in the CRD hydraulic system to assure that there is at least a 20 minute period available, follcwing the loss of both drive water pumps in the control rod drive hydraulic system, before scram becomes marginal.
212-23 212.61 The ASME Boiler and Pressure Vessel Codes,Section III, (5.2)
Article NB-7000 requires that individual pressure relief devices be installed to protect lines and components that can be isolated from normal system overpressure protection.
The information in Fermi 2 FSAR dealing with this reonirement is inadequate. With reference to the appropriate P&ld, identify those portions of the HPCI, RCIC, core spray and RHR systems that can be isolated from normal system overpressure protection.
Discuss the relief devices provided or provide the basis for deciding that relief devices are not needed.
212.62 Regulatory Guide 1.45 specifies that the sensitivity and response (5.2.7) time of the three leak detection systems required in the guide should be adequate to detect a leakage rate, or its equivalent, of one gpm in less than one hour. With respect to these requirements respond to the following:
1.
The applicant has informally notified the staff that the airborne gaseous radioactivity and airborne particulate monitoring systems meet Regulatory Guide 1.45 sensitivity and response time requirements of 1 gpm in less than one hour.
Provide the bases for such a conclusion. The response should include consideration of a) the effects of leaks in the recirculation, steam and feedwater lines, b) the effect of background radiation due to existing leaks, c) plateout, d) reactor operating times, and e) locations of sampling points and leakage detection instru-mentation in demonstrating the ability of the systems to meet the requirement.
2.
The sump monitoring system for Fermi 2 does not meet the sensitivity and response time requirements of Regulatory Guide 1.45.
If the airborne gaseous radioactivity or airborne particulate monitoring systems do not meet the sensitivity and response time requirement, it is the staff's position that the sump monitoring system should be modified to meet the requirement.
Discuss how one of these detection system will satisfy the sensitivity requirement.
In addition, discuss how the operator will determine the leak rate and how long it will take him to make this determination.
3.
Clarify the leakage detection methods employed to determine the level and flow for the drywell equipment drain sump.
What is the quantitative relationship between the drainage,
flow and sump level to the leakage rate from any source?
Provide assurance that all leakage within the drywell and reactor building will flow directly to the sumps and that there are no reservoirs which must be filled before any sump flow occurs.
212-24
~
(5.2.7) 4.
The leak detection system does not appear to be redundant.
It appears that the same instrumentation is used to monitor leakage for a variety of equipment.
Provide assurances that show that equipment failures within the leak detection system will not preclude the identification of leakage within the required sensitivity and response time.
5.
Are all the components in the leakage detection system qualified for post-LOCA environment for the long-term cooling mode of ECCS?
212.63 Three water level traces are given in the Figures in Chapter 15 (15.0 and and Appendix 15A which describe the course of the various Appendix abnormal operational transients.
Define these levels for 15A)
Fermi 2 and relate them to specific equations in NED0-10802.
Include in the definitions the location of zero level.
If these levels cannot be referenced to specific equations in NED0-10802, provide the equations used to obtain the values for these levels and the information needed to relate these equations to those in NED0-10802.
212.64 In analyzing anticipated operational transients, the applicant (15.0) has taken credit for plant operating equipment that has not been shown to be reliable as required by General Design Criterion 29.
The staff has discussed the application of this equipment generically with General Electric. Based on these discussions, it is the staff's understanding that the most limiting transient that takes credit for this equipment is the excess feedwater event.
Further, it is the staff's understanding that the only plant operating equipment that plays a significant role in mitigating this event is the turbine bypass system and the level 8 high water level trip (close turbine stop valves).
It is understood that the applicant will be providing information to demonstrate that the assumption that complete loss of steam bypass occurs during certain abnormal operational transients is not required, However, in order to assure an acceptable level of performance, it is the staff's position that the bypass system and level 8 high water level trip be identified in the plant Technical Specifications with regard to availability, setpoints, and surveillance testing.
The applicant must submit his plan for implementing this requirement along with any system modifications that may be required to fulfill the requirements.
212-25 212.65 In the response to question 212.14, it is stated that if (5.4.7) loss of the shutdown cooling suction line occurs, water can be forced through an S/R valve to the suppression pool to remove decay heat.
Provide or reference test results demonstrating that the new Target Rock S/R valves can operate with sufficient capacity with water under these conditions to achieve this safety function.
212.66 Provide the following infonnation related to pipe breaks (6.3) or leaks in high or moderate energy lines outside containment associated with the RHR system when the plant is in a shutdown cooling mode, l.
Provide the discharge rate from pipe breaks for the systems outside containment used to maintain core cooling. This value should be consistent with the requirements of SRP 3.6.1 and BTP APCSB 3-1.
2.
Determine the time frame available for recovery based on these discharge rates and their effect on core cooling.
3.
Describe the alarms available to alert the operator to the event, the recovery procedures to be utilized by the operator, and the time available for operator action.
A single failure criterion consistent with SRP 3.6.1 and BTP APCSB 3-1 should be applied in the evaluation of the recovery procedures utilized.
212.67 Significant dimensional non-conformities were reported for (6.3)
LaSalle County Station RHR, HPCS, and LPCS pumps which were dismantled to ascertain damage as a result of flooding. The staff is concerned with proper operation of all ECCS pumps; those having dimensional deficiencies add to that concern.
Provide assurance that ECCS pumps will operate as required including the RHR pumps in the long-tenn post-LOCA cooling mode. Operating histories of identical pumps, in other facilities, in both testing and/or operating modes are required in tne justification with due note taken of any differences in conditions under which pumps in other plants operated from the conditions expected for operation of the Fenni pumps.
212-26 212.68 In. Regulatory Guide 1.45, the regulatory position is that (5.2.5) leakage detection and collection systems be designed to include provisions to monitor syetems connected to the reactor coolant pressure boundary for si..s of intersystem leakage.
Methods should include radioactivity monitors and indicators to show abnormal water levels or flow in the affected area.
Provide information to demonstrata that the Fermi 2 design meets this position.
Include consideration of leakage past isolation valves and in heat exchangers.
212.69 In order to eliminate cracking of the CRD return line, GE has (4.6) recommended capping of this line.
if you intend to follow this recommendation, provide a P&ID of the CRD hydraulic system containing these modifications. Demonstrate that these modifi-cations will not impair the operability of the CRD's.
Provide a comparison of the reactor vessel makeup capability for both single and multiple CRD pump operation beTore and after the proposed modification.
212.70 The description of the differential pressure sensors used in (6.3) the loop selection logic is inadequate.
Provide information concerning the sensor type, characteristics and locations, the physical significance of the parameter measured and the basis for choice of the setpoint.
212.71 The last paragraph of 6.3.2.2.4 dealing with closure of pump (6.3) discharge valves is incorrect with respect to the previous GE loop selection logic.
In addition, the reason for the closure of only the pump discharge valve in the unbroken loop would appear associated with an injection valve failure rath:r than loop selection failure. Please clarify. The logic relative to closure of the isolation valve in the discharge side of the unbroken loop was not found in Figure 7.3-8.
Modify the FSAR text and FCD to include the functioning of this valve and any instrument specifications (e.g., pressure permissive) affecting its operation.
212.72 In the RHR system functional control diagram of Figure 7.3-8, (6.3)
Sheet 2, pennissive device blocks are given for low reactor pressure in a) the logic for recirculation pump trip and b) actuation of valves F015 and F017. Provide the reasons for these permissives.
In Table 7.3-4 the values for the low pressure permissives are noted as being approximate settings.
If these settings are allowed to vary, what is the effect on the performance evaluation of the ECCS in seccion 6.3?
212-27 212.73 On page 15B.1-2 it is stated that the thermal power monitor (Appendix (TPM) is the primary protection system for mitigating the ISB) consequences of the transient resulting from loss of feed-water heating. A description of this monitor, which typically involves the flow-weighted APRM scram in conjunction with a 6 second time constant circuit, was not found in the Fermi 2 FSAR.
Please provide this description in sufficient detail to permit evaluation of the TPM for Fermi 2.
If the time constant, which affects ss am initiation by the TPM, is less than the effective time constant for the Fermi 2 fuel for this type of type of transient, the TPM should provide a conservative measure of the time variation in surface heat flux. However, if the time constant is appreciably larger than that for the f"., the fixed APRM trip without a time constant would p'avide the scram protection. The resulting MCPR would then be ins than that predicted for the TPM scram which has a lower setpoint.
There is no current provision in the Technical Spec 1Ncations for surveillance of this time constant circuit.
It is ti.-
staff position that credit De taken only for the fixed APRh scram in the evaluation of abnormal operational transients unless the TPM is approved by the staff and appropriate limiting conditions for operation and surveillance requirements are incorporated in the Technical Specifications for Fermi 2.
212.74 A recent GE report, "DC Power Source Failure for BWR 3 and 4,"
(6.3) dated 11/1/78, provides a generic response to staff concerns relative to loss of DC power sources on peak clad temperature.
For smaller break sizes, failure of the DC power sources gives higher peak clad temperatures than failure of HPCI. The LPCI loop selection logic of Figure 7.3-8 is interpreted as indicating that if the differential pressures are not high enough to trip the differential pressure switches, the injection valves will open only in loop B.
For small breaks which might lead to this situation, this means the loop selection logic defaults to the choice of a break in loop A, even if this break actually occurred in loop B.
After 1.PCI injection, this could lead to subcooled water at the break with a higher break flow rate.
Please address this concern for Fermi 2 with respect to peak clad temperature. The ECCS performance evaluation of Section 6.3 to be provided in the Fermi 2 FSAR should include consider-ation of a DC power failure for both small and large breaks and include the effect of the change in logic involving recirculation oump discharge valves from the earlier. loop.
selection logic.
.