ML19338G236
| ML19338G236 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 10/09/1980 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Jens W DETROIT EDISON CO. |
| References | |
| NUDOCS 8010280663 | |
| Download: ML19338G236 (9) | |
Text
[&$N skR *IEGy g
I UNITED STATES
!; e # %
NUCLEAR REGULATORY COMMISSION J
3
~J/,
- WASHINGTON, D. C. 20555 3
.I f
OCT 3 1330 Docket No. 50-341 Dr. Wayne H. Jens-Assi: tant 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 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 by November 15, 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, QL Robert L. Tedesco, Assistant Director for Licensing Division of Licensing
Enclosure:
Request for Additional Information cc w/ enclosure:
See next page l
l 8 c2onw663
DCT 3 '390 Dr. Wayne H. Jens Assistant Vice President Engineering & Construction Detroit Edison Company
.2000 Second Avenue Detroit, Michigan 48226 cc: Eugene B. Thomas, Jr., Esq.
David E. Howell, Esq.
LeBoeuf, Lamb, Leiby & MacRce 21916 John R 1333 New Hampshire Avenue, N. W.
Hazel Park, Michigan 48030 Washington, D. C.
20036 Peter A. Marquardt, Esq.
Co-Counsel The Detroit Edison Company 2000 Second Avenue Detroit, Michigan 48226 Mr. William J. Fahrner Project Manager - Fermi 2 The Detroit Edison Company 2000 Second Avenue 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 l'. S. Nuclear Regulatory Commission l-Washington, D. C.
20555 Dr. David R. Schink i
Department 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.
20555
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, g
Requests and pages are numbered sequentially with respect to previously transmitted requests.
Branch Page No.
Reactor Systems Branch 212-48 through 212-52 4
r l
r k
s
. ~ -
212-48 C212.152 Provide a des:ription or reference to the recirculation flow (4A.2.3) control system in Section 4A.2.3 and address the requirements of Standard Review Plan 4.6 with' regard to the standby liquid control system. and the recirculation flow control system.
Q212.153 Do the reactivity control systems [ control rod drive system (CRDS)
(4A.2.3) and standby liquid control system (SLCS)] share any instrunentation or components? Address the vulnerability of the CRDS and SLCS to common mode failures and specify the
. common mc>1e failtre probability.valte for both systems.
Q212.154 Describe the provisions incorporated to protect water in the (4A.2.3 2) control rod drive hydraulic system and the standby liquid control systs frcm freezirg.
Q212.155 Collet fingers of the control rod drive (CRD) me:hanism have (4A.2.3 2.2.2) failed in some EWRs.
In order to resolve this problem, some EWR facilities inder construction have installed a revised collet retainer design. Will the revised collet retainer design be incorporated into the CRD me:hanisms of the Fermi 2 facility? If not, justify not -using the revised design.
Revise Appendix E.4 of the FSAR if required.
0212.156 Provide a failure mode and effects analysis for evaltating the (4A.2.3) control rod drive system as required by Regulatory Guide 1.70.
Revise Appendix F.6 as required.
Q212.157 Cescribe the normal filtration of condensate water on the suction (4A.2.3.2.2.3)and disaharge sides of the control rod drive (CRD) water ptmp.
In the description, provide the micron rating of the filters.
In addition, describe provisions in the design and operating procedtres to protect CRD hydraulic system components and instriments frcm pluggage due to failure of either the ptmp suction filters or the drive water filters, or provide justification that failure of either type filter will not cause pitggage and result in failure of the system to perform its function.
C212.158 Reference the layout studies done to assure that no interference (4A.2.3 1.3.2) exists that.will restrict the passage of control rods.
Also reference the pre-oper eional tests that are used to show acceptable performance.
0212.159 Section 5.4.6.2 of Regulatory Guide 1."O requires that significant (5.5.6) design parameters and conditions fc all components of the RCIC system be identified and that all components be shown on appropriate P&I diagrams.
Provide the significant design
. parameters and conditions for all ECIC components in Section 5.5.6 and verify.that each component can be identified on Figures 5.5-6
'(Sheet 1) and 5.5-6 (Sheet 2).
212-49 Q212.160 Provide the following information on the RCIC system:
(5.5.6) a) A description of the functional testing associated with infividual RCIC ccmponents during normal operation.
b) A description of RCIC isolation valve arrangements associated with the lines penetrating the reactor coolant pressure botndary, turbine exhaust line vactam breaker system, pump sucticn line, minimtm flow pump discharge line, and turbine exhaust line.
c) A description of the electrical interlo:ks associated with the RCIC system.
d) A des:ription of the most limiting single failtre in the combined function of the RCIC and HPCI systems.
C212.161 Several failtres of safety valve headers have resulted in valves (3.5.1) becoming missiles (NUREG-0307).
Justify why the safety valve header and valve is not considered as a credible missile.
Q212.162 For the majority of events analyzed in Section 15B.0, the (15B.0) recirculation flow control mode (automatic or manual) asstmed in the analysis is not specified. Our concern is that the mode selected ma. not result in the most severe margins on MCPR and ceak vessel pressure.
a) Specify the recirculation flow control mode assumed for each event analyzed in Section 15B.O.
b) Specify the change in MCPR and peak vessel pressure for each event if the opposite recirculation flow control mode had been asstmed in the analysis.
Q212.163 Provide a detailed discussi:n of activity above the suppression (15B.0) pool, activity releases to the environs, and offsite radiological doses for the bounding transient or accident for releases.
In addition, provide justification for the selection of the botnding transient or accident.
Q212.164 Provide an analysis of the " inadvertent EHR shutdown cooling (15B.0) operation" transient in Section 158.1 to complete the spectrtm of positive rea:tivity insertion transients. Also provide an analysis of the " failure of RHR shutdown cooling" transient in Section 15B.2 to comply with Branch Technical Position RSB 5-1 (Standard Review Plan 5.4.7).
0212.165 No evaluation of the " pressure controller failtre - closed" (15B.2.1) transient was provided in the FSAR.
Provide a quantitative evaluation of this transient assuning failure of the backup pressure regulator.
t c
l O-212-50 Q212.166 For the recircul.ation flow control failtre with increasing flow (15B.4.5.3 2) transient (Section -15B.4.5), provide the initial ocerating MCPR determined at 65% NB rated power and 50% core flow.
In addition, provide the K factors as a function of core flow for the p
automatic and manual flow control modes of operation.
Provide recirculation ptnc M-G trip setpoints for the manual flow control made asstmed in the analysis. Also, you reference the GE topical report NED0-10802 as the dynamic model to simulate this event. Because UEDO-10802 does not dederibe the complete event, discuss in greater detail the overall T.ethod used to calculate the CPR.
0212. 52A The response to 0212.52. indicates that studies show use of a 40 F (15B.5.1)
HPCI temperature is conservative.
Provide a reference to these studies.
Q212.167 In connection with parameters and asstaptions used for LOCA (158.6.5 3.2) calculations inside containment, provide the following items:
a) The basis for use of an MSIV closure time of 3.5 seconds in Table 6.2-1.
b) An explanation as to why the CBA break sizes in Tables 6.3-12 and 6.2-1 are different.
In connection with this, correct the CBA break size specified in Table 6.3-10.
c) A tabulation of all permitted axial power shapes addressed by LOCA calculations inside containment.
Identify the least favorable axial shape associated with each break size and provide justification of its conservatism.
Q212.168 In the description of event sequences for LOCA inside containment, (15B.6.5.2.1) confirm that the zero reference time for Tables 6.3-13 and 6.2-7 is the same.
Q212. 89A
' Die intent of Question 212.89 is to have the applicant provide a (15B.6.5) list of all plant-specific break sizes and locations analyzed.
In addition to this request, provide the peak cladding temperature and peak local oxidation associated with each plant-specific break size.
Q212.169 It is not clear D1 the FSAR if provisions have been made to (5.5.7)
-acconnodate thermal expansion of the water between the RHR isolation valves F008 and F009.
If a relief valve is not provided, show that piping integrity would be maintained asstming a LOCA'or steam line break in the vicinity of this piping.
Q212.170 The FSAR does not mention " flush valve" operation at Fermi-2.
Are (5.5.7) the RHR lines f1tched prior to initiation of the shutdown cooling mode? If so, are the valves operated from the control room and what is the source of the fltch water? Disctes the consequences of anitting the flush operation and/or leaving the flush valves open while attempting to initiate shutdown cooling.
~~
.~
212-51 Q212.171 Discuss the procedures for minimizing the ootential for 0
(5.5.7) exceeding the allowable cooldown rate (100 F/hr) of the RHR and the reactor coolant system when placing the plant in a shutdown cooling mcde following planned normal conditions or an emergency.
Q212.172 Deferred question till later date.
(5.5.7 )
Q212.173 Prc, vide a more detailed description and the location of the RHR (5.5.7) ptmp suction strainer inside the suppression pool.
Incitde pipe bends and the minimtm height of the suppression pool water level above the suction strainer.
Show that the NPSH at the centerline of the RHR ptmp will be met at the ptmp's design ecndition as well as at the most limiting operating condition.
Also, disc xss the size of particles that could pass throtgh the strainer ar.d continue to the RHR ptmp passages.
How much material blockage would it take to significantly affect RHR ptmp suction flow frcm the suppression pool following a LOCA?
0211. 174-Provide assurance that your relief valve design is qualified (5.2.2)
(including testing after being subjected to an environment (6.3) representative of an extended time period at normal operating conditions) to support your assumption that four of the five ADS valves will operate. A quantitative history of safety / relief valve operation, including similar valves in other plants should be included in this evaluation.
Subsection 5.2.2.5 of the FSAR states that it is not feasible
~
to test the safety / relief valve setpoints while the valves are installed.
It would appear that improper setpoints (e.g., an erroneous setpoint calculation) would be a credible conmon mode failure which could result in degradation of the pressure relier systems. Show that adequate safety margin has been incitded in the overprenrization analysis to protect against a ccemon mode failure of the safety / relief valves to open at'the prescribed values.
/
Q212.175 It is unclear diether comparative " grab" samples of the (5.2.7) contintously monitored containment atmosphere can and will be taken on a periodic basis.
Resolve this ambiguity.
If
" grab" samples are not to be taxen, justify the emission of these comparative data.
0212. 176 Identified leakage is determined dtring pre-operational (5.2.7) testing or is measurable during reactor operation.
Provide the frequency that these data will be recorded and indicate what procedural guidelines are to be used to identify trends.
i
- 212-52 Q212. 38A The' response to Q212.38 requires additional clarification.
(5. 2.7 )
The maximtm anticipated leaktge rate and the maximtm.
allowable time. for operator action have not been. identified.
' (6.3)
. Provide a scenario for the response of the leak detection system i
and the operator response for ttle-maximtm anticipated leak rate.
-Included in this response should be quantitative values for the j
leak rate and - the response times.
+
Is the air supply line to the safety / relief valve acetmulators 0212. 177'
- (5.2.2) safety grade. (Figtre 5.1-3 ~ and Drawing 6M721-2089)? If the 1
- air ' supply line were to break upstream of the ball check valve, would there be an indication of this break and an indication of the.acetmulator status in the control room? If indications are given, -what operator action muld be' required?
0212.178 State -the ntaber of safetp/ relief valve a:tuations permitted j
(5.2.2)-
between maintenance periods and describe how the ntmber of actuations are recorded.
4 4_
Q212.179 Does your design incorporate-a fast scram system? If so, has (5.2.2) this _been-accotuted for in yotr overpressurization analysis?
t-1 s
)
?
t 4
i e
d f'
.g'
.~. --
y o
.,.,,.,,-,-..._m.
., _,,.....