ML19316B180
| ML19316B180 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 05/27/1980 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | Jens W DETROIT EDISON CO. |
| References | |
| NUDOCS 8006120016 | |
| Download: ML19316B180 (14) | |
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- p nnery UNITED STATES
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j NUCLEAR REGULATORY COMMISSION E
WASHINGTON, D. C. 20555
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N 27 g 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 Report (FSAR) for the Enrico Femi 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 by August 1,1980.
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,
l'%ht'sw6t.-
u A. Schwencer, Chief Licensing Branch No. 2 Division of Licensing
Enclosure:
Request for Additional Infomation cc w/ enclosure:
See next page 8006120oI(,
o Dr. Wayne H. Jens Assistant Vice President Engineering & Construction Detroit Edison Company 2000 Second Avenue Detroit, Michigan 48226 cc:
Eugene B. Thomas, Jr., Esq.
Mr. Jeffrey A. Alson LeBoeuf, Lamb, Leiby & MacRae 772 Creen Street, Building 4 1333 New Hampshire Avenue, N. W.
Ypsilanti, Michigan 48197 Washington, D. C.
20036 David E. Howell, Esq.
Peter A. Marquardt, Esq.
21916 John R Co-Counsel Hazel Park, Michigan 48030 The Detroit Edison Company 2000 Second Avenue Mrs. Martha Drake Detroit, Michigan 48226 230 Fairview Petoskey, Michigan 49770 Mr. William J. Fahrner Project Manager - Fermi 2 William J. Scanlon, Esq.
The Detroit Edison Company 2034 Pauline Boulevard 2000 Second Avenue Ann Arbor, Michigan 48103 Detroit, Michigan 48226 Mr. 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 Departnent of Oceanography Texas A & M University College Station, Texas 77840 Mr. Frederick J. Shon Atomic Safety & Licensing Board Panel U. S. Nuclear Regulatory Commission Washington, D. C.
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ENCLOSURE REQUESTS FOR ADDITIONAL INFORMATION ENRICO FERMI ATOMIC POWER PLANT UNIT 2 DOCKET NO. 50-341 Requests by'the following branch in NRC is included in this enclosure.
Requests and pages are numbered sequentially with respect to pmviously transmitted requests.
i Branch Page No.
Reactor Systems Branch 212-37 through 212-47 s
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212.37 21 2.0 REACTOR SYSTEMS ~ BRANCH 212.41A The response to Q212.41 is not acceptable. Provide a quantitative (158.1.1.3.2) analysis of the " loss of feedwater heating" transient assuming a 1500 drop in feedwater temperature.
212.45A The response to Q212.45 specifies appropriate transients have been (158.3.3.3.3) re-analyzed without recirculation pump trip (RPT). However this trip is discussed in the text for the " recirculation pump seizure" transient. Has credit been taken for the RPT for this transient?
212.47A In the response to Question 212.47, it is indicated that the 0
(158.1.4.2.1) suppression-pool-temperature Technical Limit of 110 F is reached after 6 minutes.
Include this item in Table 15B.1.4-1.
In addition, add the following safety actions and their proper time of occurrence to Table 158.1.4-1 to be consistent with Figure F.6-13:
a) Scram b) Reactor vessel isolation c) Pressure relief d)
Initial core cooling 212.62A Your response to previous request 212.62 is inadequate.
(5.2.7)
Please clarify the following specific items:
(1) What are the sensitivities and response times of the containment airborne particulate and airborne gaseous radioactivity monitors? Pr6 vide this infonnation in counts per unit time, as well as in its leakage equivalent of gallons per unit time. Provide the sources of the above information.
(2) The pump monitoring system contains level switches to detect leakage into the sumps. The sensitivity and response time guidance of Regulatory Guide 1.45 cannot be separated. The difference in time between level switch activations will be equivalent to a specific sump level change and an average leak rate in that time.
The sump level monitors can either sense this level change, or not. The requirements cannot' be met in a longer time than specified. Therefore, the staff's position continues to be that unless the containment atmosohere radiation monitoring system can meet the one gallon per minute in one hour sensitivity and response time requirements, the sump monitoring system shculd be modified to meet the guidance of Regulatory Guide 1.45.
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212-38 212.73A The response to Question 212.73 is not acceptable.
Insufficient detail is provided to evaluate the themal power monitor (TPM).
It is the staff's position that credit be taken only for the fixed APRM (high flux) scram and accordingly that those abnormal operational transients for which the TPM is assumed to initiate reactor scram be reanalyzed taking credit only for the fixed APRM scram.
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212-39 REACTOR SYSTEMS BRANCH 212.124 Provide assurance that adequate NPSH exists for the remaining ECCS (6.3) systems after an ECCS passive failure such as pump seals or valve seals in a water-tight pump room from which the water could not drain back to the suppression pool.
212.125 Severe water hanner occurrence in the ECCS discharge piping during (6.3) startup of the ECCS pumps is avoided by ensuring that the discharge pipes are maintained full of water. The domineralized water system described in the response to Question 212.36 is used to achieve this function for the RHR and Core Spray discharge piping. Since the demineralized water also supplies water to numerous other systems, the following areas need to be clarified:
a) Justify the use of a common filling system for the RHR and the Core Spray piping versus independent jockey pumps.
b)
Identify the expected demands on the demineralized water system.
What is the effect on the availability of water in keeping the discharge pipes full because of these other demands on the demineralized water system?
c) The discharge piping " fill system" is apparently considered to be an auxiliary system. Are any priority interlocks provided to ensure that the " filling system" will be given priority over the other uses of the demineralized water system water?
d) The individual fill lines apparently do not have instrumentation to monitor low pressure. What assurance is there that when the demineralized water pumps are operating that the individual RNR or Core Spray discharge lines are full of water?
The HPCI discharge piping is kept full of water by feedwater flow because the HPCI pipe intercepts the feedwater pipe from the bottom side of the pipe. Provide assurance that the HPCI pipe will be filled and that air pockets can not form in the piping. Describe any instrumentation that would be used to monitor the HPCI discharge piping.
212.126 Figures 6.3-25 and 6.3-33 indicate that the 60% and 80% DBA have a 0
(6.3) maximum peak cl. ' ling temperature of about 2100 F.
Explain why the PCT for the 60% and 80% DBA analysis yielded the same PCT as the DBA.
Provide a plot to show the PCT of the lowest powered node to experience a CPR <l.0 prior to jet pump uncovery for the following figures:
6.3-25 6.3-33 6.3-44
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(6.3)
Explain why the onset of boiling transition shown for the DBA and the 60% DBA occur at about the same time. Also, why are the DBA and 60%
DBA boiling transition times both greater than the 80% DBA valw?
f 212.127 Diversion of a LPCI loop to containment cooling is supposedly (6.3) prevented unless:
a) the accident initiation signal for containment cooling is
- present, b) the reactor vessel has been reflooded te at least 2/3 core level.
l However, a key locked switch in the control room pennits manual over-ride of these pennissives. What instructions are provided to the operator for use of this override switch?
212.128 Section 6.3.2 states that no operator action is required for at least (6.3) 10 minutes following the initiation of ECCS operation. Section 7.3.1.2.1 indicates that operator may be required to sustain core cooling with the HPCI system after 10 minutes.
Provide an analysis to show the consequences of delaying operator actions for at least 20 minutes.
Confirm that adequate NPSH will exist 11! operator action is not initiated prior to 20 minutes after a LOCA (see Question 212.76).
212.129 Identify the specific valves included in Table 3.2-1 under CRD (3.2.2)
Hydaulic System, item 2.
212.130 Discuss the possibility of the CR0 mechanism becoming a missile (3.5.1) inside containment.
212.131 Discuss the potential for missiles inside the containment due to (3.5.1) gravitational effects (of such components as electrical hoists or any unrestrained equipment) during maintenance, reactor operation, and following a LOCA.
212.132 Subsection 3.5.1.1.2 of the FSAR refers to an investigation of (3.5.1) potential missiles due to pressurized component failure (valve bonnets, valve items, thennowells, piping, etc.). Reference a report on the results of this investigation or supply sample calculations showing how you arrived at the conclusion that the velocity of missiles from the above sources would be too low to cause penetration of important equipment.
212.133 The performance of essentially all types of safety / relief valves has been (5.2.2) less than expected for a safety component. Because of reportable events involving malfunctions of these valves on operating BWRs, the staff is of the opinion that significantly better safety / relief 4
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valve perfonnance should be required of newer plants. Provide a detailed description of improvements between your plant and presently operating plants in the areas listed below.
In additicn, explain why the noted differences will provide the required perfonnance 1
improvement.
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(1) Specifications. What new provisions have been employed to ensure that valve and valve actuator specifications include 1
design requirements for operation under expected environmental j
conditions (esp. temperature, humidity, and vibration)?
(2) Quality Assurance. What new programs have been initiated to assure tnat valves are manufactured to specifications and i
will operate to specifications? For example, what tests are performed by the applicant to assure that the blowdown capacity is correct?
(3) Valve Operability. Provide your surveillance program to monitor the perfonnance of the safety / relief valves.
Identify the information that will be obtained and how these data will l
be utilized to improve the operability of the valves.
For example, how will this program reduce the malfunctions that have occurred in operating reactors?
(4) Valve Inspection and Overhaul. The FSAR states that one half of the safety / relief valves will be bench checked and visually inspected every refueling outage. However, depending on operating cycle length, this may result in several years htween inspections. Operating experience has shown that
- afety/ relief valve failure may be caused by exceeding the i
manufacturer's recomended service life for the internals of i
the safety / relief valve or air actuation. At what frequency do you intend to visually inspect and overhaul the ADS portion of the safety / relief valve?' For both safety / relief and ADS l
modes, what provisions exist to ensure that valve inspection and overhaul are in accordance with the manufacturer's recomendations and that the design service life is not exceeded for any component of the safety / relief valve?
212.134 Explain the increase in flux at time T=2-3 seconds in Figure 158.4.4-1.
j (158.4.4) 212.135 Explain why the transient resulting from recirculating flow (15B.4.5) control failure with increasing flow is most severe at 65% NBR power and 50% core flow. Correct the discrepancy of the stated starting point for the transient on page 15.1.11-5(58% power and 48% flow) vs that stated on page 158.4-29(65% power and 50% flow).
j 212.136 a) Modify Table 158.0-2 as follows:
(15B.0)
- 1) Provide actual values of MCPR instead of the entries >l.06 and >l.10.
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- 2) Provide all data for both generator load rejection events and the turbine trip event with steam bypass.
- 3) Replace the dashes (-) under the " Duration of Blowdown" column with appropriate values.
b) Correct inconsistencies between parameter values in Table 15B.0-2 and appropriate values in the text description for each transient and accident.
For instance, the MCPR for the " turbine trip with partial bypass failure" transient is indicated as 1.14 in the text description and as 1.18 in Table 158.0-2.
c) For each event category in Table 15B.0-2, identify the most limiting transient or accident for MCPR and maximum vessel pressure.
d) Include the " inadvertent safaty/ relief valve opening" transient in Table 158.0-1.
e) Explain why transient 158.2.1 does not appear in Table 15B.0-1 o-in Section 158.2 of the FSAR.
212.137 The depressurization rate has a proportional effect upon the voiding (15B.1.3.3.4) action of the core. For the " pressure regulation failure-open to 130"." transient, the assumed depressurization rate results in a L8 trip.
If a smaller depressurization rate is assumed, a trip from low turbine inlet pressure would occur. Provide justification that the assumed depressurization rate, provides the most restrictive margins on MCPR and peak vessel pressure.
212.138 Provide the following information regarding generator load rejection (158.2) transients:
a)
It is indicated that the "genarator load rejection" transients with bypass and partial bypass failure are less severe than the corresponding " turbine trip" transients. Normally the opposite is true; specific examples of this are the analytical results.
frtm the Susquehanna, WPPSS, and Grand Gulf FSARs, and the 238 Nuclear Steam Supply System GESSAR. Explain this difference.
b)
Explain why the MCPR for the " generator load rejection with partial bypass failure" transient has a MCPR of < 1.14 it if is, in fact, less restricitve than the coresponding " turbine trip" transient which has an indicated MCPR of 1.14 (Section 15B.2.3.3.3.2).
c) Provide analytical results (figures) for both " generator load rejection" transients and provide more detail in Table 15B.2.2-1 a9 158.2.2-2 for the time period indicated in the figure to be pro, dad.
d) NSOA Figure F.6-6 for the " generator load rejection" and
" turbine trip" transients is too general.
It does not account for bypass failure and it does not include the initial care
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212-43 1
(15B.2) cooling and containment isolation safety actions. To conform with standard practice, provide a separate NSOA -figure for each
" generator load rejection" and " turbine trip" transient, 212.139 In a letter to the applicant on December 27, 1978, titled "Sunnary (15B.2.2) of November 28, 1978 Meeting Regarding Turbine Trip Transient (158.2.3)
Analysis," the staff accepted the pmposed revision to the analyses method for the " generator load rejection" and " turbine trip" transients provided the four comments on pages 2 and 3 are addressed in the revised analysis. All connents have been addressed or response requested by other staff questions with the exception of connent (2). Address connent (2) in the analysis of this transient.
212.140 In the analysis of one and two recirculation pump trip events in (15A.3.1.3.2) Section 158.3.1, a minimum design rotating inertia was used to obtain a predicted rate of decrease in core flow greater than expected. Specify the inertia value used for each transient and accident in Section 158.0, and the selection basis for each.
In the selection basis, quantitatively include the effect on MCPR and reactor vessel pressure.
'12.141 Explain why the consequences of the " recirculation pump trip," the (158.3.1.5,
" recirculation flow control failure - decreasing flow," and the 158.3.2.5,
" recirculation pump seizure" events do not seem to result in 158.3.3.5) discharge of nomal coolant activity to the suppression pool.
212.142 The " recirculation flow control failure-decreasing flow" transient i
(15B.3.2.2.1) is similar to the " trip of both recirculation pumps" transient with failure of the master controller. Since the scram safety action occurs, provide a NSOA figure for event 25 in Appendix F.6.
Typically, the pressure relief, reactor vessel isolation, and initial core cooling safety actions occur for this transient.
Include these additional safety actions in the NSOA figure is they are required for the Enrico Fermi 2 design. Explain why the text description for event 25 indicates that protection system action is not required.
212.143 For the " recirculation pump seizure" transient:
(158.3.3.2.1) a) Based on the reactor scram at 2.62 seconds, the core flow and power level should not stabilize at a new equilibrium condition at 20.0 seconds as indicated in Table 15P.3.3-1.
Correct this discrepancy.
1 b) Typically, the pressure relief, containment isolation, and initial core cooling safety actions occur for this transient.
Explain why they do not occur for the Enrico Fermi 2 design.
c) Since the scram safety action occurs for this transient, provide a NSOA figure for event 28 in Appendix F.6.
Include the safety actions in step b) above if they occur and update the text
' description for event 28 accordingly.
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212-44 212.144 SRP 3.2.1,Section III, specifies that "for systems which are (3.2.1) partially seismic Category I, the Category I portion of the system should extend to the first seismic constraint beyond the isolation i
valves which separate the seismic frem the non-seismic portions of f
the system." The figures in the FSAR and the P&I drawings do not I
show the seismic ronstraints. Provide assurances that, where l
appropriate, the seismic Category I criteria extend to the first l
seismic constraint beyond the isolation valves.
212.145 In the analysis for the generator load rejecti6n and turbine trip transients, credit is taken for immediate reactor scram from a valve closure signal (turbine control valve (TCV) for load rejection, turbine stop valve (TSV) for turbine trip) in the non-seismically qualified turbine building. Provide the following 1
analyses:
a) Analyze these transients from a SSE event w/o taking credita for a direct scram from the TCV or TSV closure signal and using only safety grade and seismically qualified equipment components, and structures. Provide the following analytical results:
- 1) Result curve:: similar to those in Figure 15B.2.3-1.
- 2) Peak vessel pressure and MCPR.
- 3) Verification that radiological consequences are within acceptable guidelines.
- 4) Percent of fuel rods that expe-ience boiling transition.
Limited fuel rod failure can be allowed if verification of acceptable radiological consequences is provided in Step 3 above.
- 5) Effect that failure of a single safety grade component has on ACPR, peak vessel pressure, and radiological consequences.
b) Analyze the transient in part (a)..in the same manner with c61ncident loss of offsite power.
212.146 The subject of LPCI loop selection requires additional clarification.
(6.3)
Neither Section 6.3.2 nor 7.3.1.2.4 prpvide enough infomation to evaluate this system.
It is inferredtrom Figure 7.3-8 and from the responses to Q212.70 and Q212.74, that the LPCI loop selection logic selects loop B for LPCI water injection for all situations except when a break is detected in the loop B.
Provide a detailed description of the loop :: election logic and justify the choice of using only one loop (loop B) for all situations other than a break in loop B.
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212-45 212.147 GE calculations performed for decrease in reactor coolant (15B.0) temperature (Section 158.1) and for reactor pressure increase (Section 15B.2) events using the proposed ODYN Ifcensing basis model (NEDO-24154) have shown that in some cases a more limiting CPR is predicted than by the current REDY licensing bases model (NEDO-10802). Based on a letter to Glen B. Sherwood dated 1/23/80 from Richard P. Denise, the staff's ODYN licensing position is that GE can proceed with ODYN analysis of certain events described in Section 15 of ifcensing application Safety Anals is Reports. Provide the following:
a) An ODYN analysis of the applicable events listed in Table 2-1 and 2-2 of NEDO-24154-P.
i b) A list of all input parameters for each event.
c) Justification that input parameters for each event are conservative.
212.148 Branch Technical Position - RSB No. 3-1, states the Quality Group l
(3.2.2) classification for (a) main steamlines from second isolation valve to turbine stop valve, and (b) main steamline branch lines to first valve, should be Quality Group B.
These lines are classified as Quality Group D+ in Fermi 2.
Confim that the above steamlines will be subject to in-service inspection requirements contained in recent revisions of ASME Code Section XI for Class 2 components.
212.149 A recent experience at Millstone has uncovered a potential problem with (6.3) the LPCI loop selection logic (LER 80-02/1T). GE has found that the LPCI loop selection logic instrumentation may be insensitive to the low pressure differential that would result from a small break in the discharge side of the recirculation loops and, therefore, it would be possible to inject LPCI flow into the broken loop. A postulated break size below the detection capability of the loop selection system could then be more limiting than the large break presently specified in the FSAR.
Show that the' instrumentation for the LPCI loop selection logic at Femi is responsive to the small pressure differentials that result from breaks in the recirculation discharge piping.
If the sensitivity of the instrumentation cannot satisfy these conditions, provide analyses over the range of the small break spectrum to determine if the small break is limiting.
If it is limiting, provide a brief description of the proposed changes that will be implemented to ensure that reactor operation will be within limits.
212.150 In reference to missile protection it is not clear from Figure (3.5.1) 3.5-1 that the HPCI and RCIC lines inside containment cannot both be simultaneously damaged due to pipe whip or fragments of pipes.
Provide the locations of pipe whip restraints inside containment and clarify Figure 3.5-1 to better show the relative positions of the HPCI and RCIC lines.
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212-46 212.151 Amend the FSAR as follows:
1.
Section 6.3.2.2.1 of the FSAR states thac the HPCI is brought to design flow rate within 25 seconds of a RPV low water level signal. Table 6.3-12 states that the HPCI system is at rated flow 30 seconds after a LOCA.
Resolve this inconsistency and verify that all transient and accident analyses incorporate the proper timing.
2.
Section 6.3.2.2.4.1 of the FSAR states that the LPCI injection valves are fully open in 37 seconds after a LOCA. Table 6.3-12 states that the injection valves are fully open in 43 seconds.
Resolve this inconsistency and verify that all transient and accident analysis incorporate the proper timing.
3.
In Section 5.5.1.4 of the FSAR, it states that destructive recirculation pump motor overspeed can be prevented by use of a decoupling device in the shaft between the pump and motor. This conclusion is supported by reference to Report NED0-10677. This report has been superceded by the letter "GE Recirculation Pump Potential Overspeed" by E. A. Hughes (GE) to R. C. DeYoung (NRC), January 18, 1977.
In the letter the conctasion was reached that a decoupling device was not necessary.
In Sub-section 3.5.1.2.2, you reference the Hughes to DeYoung report and state that pump shaft failure will decouple the rotor from 1
the overspeed driving blowdown force. Update Section 5.5.1.4 of your FSAR.
4.
Complete Table 15B.4.4.-1, " Sequence of Events for Startup of Idle Recirculation Loop". The final stabilized condition has not been reached in 10.4 seconds with initiation of a reactor high flux scram, but rather when vessel water level is again stabilized.
5.
Complete Table 158.4.5-1, " Sequence of Events for Recirculation Flow Controller Failure with Increasing Flow." The final stabilized condition has not been reached in 1.5 seconds with initiation of high flux scram.
5.
The thermal power monitor (TPM) is indicated as the primary protection system for mitigating the consequences of the " loss of feedwater heating transient". Section 15B.1.1.2.2 and 158.1.1.3.3 and Table 158.1.1-2 indicate a scram on high APRM thernal power (flow-biased signal) where as Figure F.6-10 indicates a high flux (fixed APRM signal) scram. Resolve this discrepancy.
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Figure F.6-19 does not include initial core cooling and reactor vessel isolation safety actions for the "feedwater controller failure at maximum demand" transient. Explain why these safety actions are not included in Figure F.6-16 and Table 158.?.2-1.
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