ML20049H429
| ML20049H429 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 01/31/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0798, NUREG-0798-S02, NUREG-798, NUREG-798-S2, NUDOCS 8203030004 | |
| Download: ML20049H429 (67) | |
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NUREG-0798 Supplement No. 2 Safety Evaluation Report related to the operation of Enrico Fermi Atomic Power Plant, Unit l\\ o. 2 Docket No. 50-341 l
Detroit Edison Company, et al.
U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation January 1982 p
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NOTICE Availabihty of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from onc of the following sources:
1.
The NRC Public Document Room,1717 H Street., N.W.
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NUREG-0798 Supplement No. 2 Safety Evaluation Report related to the operation of Enrico Fermi Atomic Power Plant, Unit No. 2 3
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Docket No. 50-341 Detroit Edibon C5mpany, et al.
U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation January 1982 j..a % y,_
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ABSTRACT Supplement 2 to the Safety Evaluation Report related to the operation of the Enrico Fermi Atomic Power Plant, Unit 2 provides the staff's evaluation of additional information submitted by the applicant regarding outstanding review issues identified in Supplement 1 to the Safety Evaluation Report, dated September 1981 9
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Contents Page Abstract...............
iii 1
Introduction and General Discussion..
1-1 1-1 1.1 Introduction.........
- 1. 8 Summary of Outstanding Issues.............
1-2 1.8.1 Prelicensing Issues.........................
1-2 1-3 1.8.2 License Conditions.......
3 Design Criteria for Structures, Systems, and Components........
3-1 3-1 3.7 Seismic Design 3.7.7 Seismic Input........
3-1 3.10 Seismic and Dynamic Qualification of Mechanical and Electrical Equipment Important to Safety........
3-1 3.11 Environmental Qualification of Safety-Related Electrical Equipment...
3-3 4
Reactor......
4-1 4.2 Fuel System Design........
4-1 4.2.3 Design Evaluation...................................
4-1 4.4 Thermal and Hydraulic Design..........
4-2 4.4.1 Design Evaluation...............................
4-2 6
Engineered Safety Features................
6-1 6.2 Containment Functional Design.....
6-1 6.2.7 Containment Leak Testing........................
6-1 l
6.3 Emergency Core Cooling System.....................
6-2 l
- 6. 3.4 Evalution Findings.........
6-2 1
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Auxiliary Systems..............................................
9-1 9.5 Fire Protection, Communication, Lighting, and Emergency Diesel Engine Systems..............................
9-1 9.5.1 Fire Protection....................................
9-1 9.5.7 Emergency Diesel Engine Lubricating Oil System.....
9-1 v
Fermi SSER2
CONTENTS (Continued)
Page 13 Conduct of Operations.
13-1 13.5 Industrial Security 13-1 16 Technical Specifications.
16-1 22 TMI-2 Requirements...
22-1 22.2 TMI Action Plan Requirements for Applicants for Operating Licenses 22-1 II Siting and Design........
22-1 II.B.3 Postaccident Sampling Capability 22-1 II.K.3.22 Automatic Switchover of Reactor Core Isolation Cooling System Suction - Verify Procedures and Modify Design 22-3 II.K.3.44 Evaluation of Anticipated Transients with Single Failure To Verify No Fuel Failure...............
22-4 III Emergency Preparedness and Radiation Protection 22-4 III.A.2.2 Improving Licensee Emergency Preparedness...................
22-7 Appendix A Chronology Appendix B Bibliography Appendix C NRC Staff Contributors and Consultants Appendix E Fire Protection Review Appendix F Environmental Qualification Safety Evaluation Report vi Fermi SSER2 1
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1 INTRODUCTION AND GENERAL DISCUSSION I
1.1 Introduction The " Safety Evaluation Report related to the-operation of Enrico Fermi Atomic i.
Power Plant, Unit No. 2" (NUREG-0798) (SER', prepared by the staff of the Nuclear Regulatory Commission (staf f), wa:. isst.d on -July 10, 1981.
The SER provided a summary and results of the staf f's radiological safety review of the application by the Detroit Ed sion Company (applicant) for an operating license j
for Fermi 2.
The SER concluotJ that upon favorable resolution of outstanding
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matters described therein, the plant could be operated without endangering the health and safety of the public.
Supplement No. I to the SER provided (1) the staff's evaluation of additional q
information provided by the applicant regarding outstanding review issues identified in the SER, (2) the staff's evaluation of additional information provided by the applicant regarding revised designs, and (3) the staff's i
response to the comments in the report by the Advisory Committee on Reactor Safeguards (ACRS).
i By Amendment 40 to the Final Safety Analysis Report (FSAR) and by letters j
identified in Appendix A to this supplement, the applicant has provided addi-tional information regarding several of the outstanding issues in Supplement No. I to.the SER.
I This supplement (Supplement No. 2 to the SER) provides the staff's evaluation i
of additional information provided by the applicant regarding some of the out-standing issues identified in Supplement No. 1.
j Except for the appendices, each section of this supplement is numbered and j
titled the same as the corresponding section of the Safety Evaluation Report that has been affected by the additional evaluation.
Except as noted, each 4
section is supplementary to the corresponding section in the SER.
Appendix A i
to this supplement is a continuation of.the chronology of principal actions i
related to the staff's safety review of the application.
References are listed j
in Appendix B.
The NRC licensing project manager for the review of the Fermi 2 l
operating license' application is Mr. Lester L..Kintner.
Mr. Kintner may be contacted by calling (301) 492-7070 or writing:
Mr. L. L. Kintner Division of Licensing i
U.S. Nuclear Regulatory Commission i
Washington, DC 20555 This SER is a product of the NRC staff.
NRC' staff members who were principal contributors to this report are identified in Appendix C.
A number'of consultants assisted the staff in the review, fhe organizations which provided consultants to the staff are listed below.
The individual con-sultants are listed in Appendix C.
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EG&G Idaho, Inc.
Gage-Babcock & Associates Brookhaven Nation ~1 Laboratory Appendix D contains a list of errata for Supplement No. 1 to the SER.
Appendices E and F are the staff's evaluation of fire protection and environ-mental qualification of electrical equipment, respectively.
- 1. 8 Summary of Outstanding Issues 1.8.1 Prelicensing Issues The partial or complete resolution of some of the outstanding issues identified in Supplement No. 1 to the SER is described in appropriate sections of this supplement.
The outstanding issues remaining in the staff operating license review are listed below, with the number of the appropriate section in the SER, Supplement No. 1 to the SER, or this supplement that describe the issue, the status, and plans for resolution. The staff will complete its review of these items before the operating license is issued.
The resolution of these items will be discussed in a future supplement to the SER.
(1) Mark I containment analyses (SER Sections 3.8.1, 3.8.2, and 3.9.3, and SER Supplement No. 1 Sections 3.d.4, 3.9.3, 3.10, and 18)
(2) Seismic and dynamic qualification of equipment (SER supplement No. 1 Section 3.10 and this supplement Section 3.10)
(3) Environmental qualification of safety-related electrical equipment (this supplement Section 3.11)
(4) Seismic and loss-of-coolant accident (LOCA) loading on fuel (SER Section 4.2.3)
(5) Break in control rod scram discharge. volume (SER Supplement No. 1 Section 6.3.4.1)
(6) Procedures for testing interlocks on 'he residual heat removal system (SER Section 7.4.2)
(7) Radwaste system modifications (SER Supplement No. 1 Section 11)
(8) TMI Issues (Section 22)
(a)
I.C.7 NSSS-vendor review of procedures (SER)
(b)
I.G.1 Training during low power testing (SER)
(c)
II.B.3 Interim procedure for postaccident sampling (this supplement)
(d)
II.D.1 Testing of safety relief valves (SER)
(e)
II.E.4.2 Containment isolation dependability (SER)
(f)
III.A.1.'1 Upgraded emergency preparedness (SER)
III.A.1.2 Emergency response facilities (SER)
III.A.2 Improved emergency preparedness (this supplement) f I
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Fermi SSER2 1-2' L.
1.8.2 License Conditions In its review of outstanding issues identified in Supplement No. 1 to the SER, the staff has resolved some of the issues that were identified in Supplement No. 1 to the SER as license conditions and has identified additional license conditions.
The license conditions remaining in the staff licensing review are listed below, with the number of the appropriat' section in the SER, Supplement No. 1 to the SER, or this supplement that discusses the license condition.
Technical Specifications resulting from the review are listed in Section 16 of this supplement.
(1) Modifications to piping and equipment attached to Mark I containment (SER Section 3.8.1 and SER Supplement No. 1 Sections 3.9.3, 3.10, and 18)
(2)
Inservice testing program for pumps and valves (SER applement No. 1 Section 3.9.6.1)
(3) Environmental qualification of equipment (this supplement Section 3.11)
(4) Hydrodynamic stability analysis (SER Section 4.4.1)
(5) Study of multiple control system failures (SER Section 7.2.2)
(6) Modifications to fire protection equipment (this supplement Section 9.5.1)
(7) Low pressure turbine-disc inspection (SER Section 10.2.2)
(8) Retention of persons with BWR operating experience on shift until 100 per-cent power is achieved (SER Section 13.1 and SER Supplement No. 1 Section 18)
(9)
Implementation of safeguards contingency plan, guard training plan, and physical security plan (this supplement Section 13.5)
(10) Results of turbine trip startup tests (SER supplement No. 1 Section 15.1)
(11) Station blackout simulator exercise (SER Appendix C and SER Supplement No. 1 Section 18)
(12) Final procedure for postaccident sampling (this supplement Section 22, Item II.B.3) i (13) Instrumentation for detection of inadequate core cooling (SER Section 22, Item II.F.2, and SER Supplement No. 1 Sections 18 and 22, Item II.F.2).
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DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, AND COMPONENTS 3.7 Seismic Design 3.7.1 Seismic Input In Supplement No. I to the SER, the staff concluded that input values used for the Fermi 2 seismic design are acceptable subject to documentation of the method of seismic response combination and acceptable site-specific seismic response spectra for 5 percent structural damping.
By letter dated October 27, 1981, the applicant provided the following statement to be included in FSAR Section 3.7.2.1.2.4 in the next amendment:
In the original design performed in 1971, horizontal and vertical seismic effects were not combined in structural design.
In subsequent analyses, the effects of two statistically independent time histories were added algebraically and then combined with the vertical component effect by the SRSS rules.
On the basis of the applicant's October 27, 1981 letter, the staff concludes that the seismic response combination has been satisfactorily documented.
The staff will verify incorporation of this statement into the FSAR.
The applicant met with the staff on October 29, 1981 to discuss the method for generating site specific seismic response spectra and floor response spectra.
The applicant agreed to use the same time history to generate all site-specific 3
i and floor response spectra used in the Fermi 2 seismic design analyses.
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By letters dated December 23, 1981, and January 15, 1982, the applicant j
transmitted response spectra for 5 percent and 7 percent structural damping using the same time history labeled by the applicant as TH-5.
Both of these response spectra are shown to be closely enveloping the design response spectra.
The floor response spectra generated for 5 percent equipment damping and 5 percent structural damping using TH-5 time history were contained in the applicant's letter dated September 4, 1981.
The applicant has agreed to pro-vide floor response spectra for lower equipment damping values required for the equipment qualification reassessment using the same TH-5 time history.
4 On the basis of its review, the staff concludes that acceptable floor response spectra for 5 percent structural damping and 5 percent equipment damping have been satisfactorily documented.
The staff will verify that acceptable floor response spectra required for lower equipment damping values, including i
2 percent, are satisfactorily documented.
3.10 Seismic and Dynamic Qualification of Mechanical and Electrical Equipment Important to Safety In Supplement No. I to the SER, the staff stated that it had identified con-cerns regarding equipment qualification as a result of the plant site audit 1
Fermi SSER 2 3-1
conducted by the seismic qualification review team (SQRT) on July 17-31, 1981.
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By letters dated August 31, 1981 and September 4, 1981, the applicant responded to a portion of these equipment-specific and generic concerns.
The applicant has committed to respond to the remainder of the identified items of concern by September 1, 1982.
The JLaff, with the assistance of the Brookhaven National Laboratory (BNL), has reviewed applicant's additional submittals.
The current status of the applicant's equipment seismic and dynamic qualification program and the extent of his resolution for the seven identified items of concern are summarized as follows:
(1) For Barton Flow Transmitters qualification tests were not conducted at the transmitter's resonant frequency of 30 Hz.
Justification of the device's operability at this frequency, at provided in the applicant's submittal of August 31, 1981, is not acceptable.
In addition, the possible effect of amplification of the input acceleration by the base plate has not been discussed.
By letter dated December 3, 1981, the staff asked the applicant to provide additional clarification of the above two areas in i
discussing the operability of the transmitter at its resonant frequency of 30 Hz.
i (2) For the GE Relay, the following items of concern as listed in the staff letter of December 3,1981 need to be resolved by the applicant:
i (a) The malfunction limit (in g's) of the relay at acceptable chatter limit of 20 ms needs to be documented.
i (b) The expected maximum amplification factor at the relay location on the Fermi 2 panel needs to be clarified in order to arrive at a maximum relay acceleration with respect to'the required acceleration, 1.5 g, at the floor level of this equipment.
A Cofrentes panel, H13-P618, claimed to be applicable to that at Fermi, has been tested for amplification and found to be 7.8 at the top and 4.0 toward the center for the frequency range up to 33 Hz.
The applicant is requested to discuss the nature of the similarity between the two panels loaded with their own instruments or the rational of the dynamic relationships that might be expected'between them.
(3) For the GE Rack, the applicant has provided supporting documents to show that the resonance in side-to-side and vertical directions is higher than that in front-to-back direction.
This response is acceptable to the staff.
4 (4) Starting January 15, 1982, and every 6 months thereafter, until commercial operation, the applicant will provide an updated list of equipment which was either not cualified or not installed at the time of SQRT audit.
This commitment is
' table to the staff.
r (5) The applicant hos identified, per SQRT's request, the equipment that is included in the SQRT Equipment Summary List but is not considered to be safety related as GE-suppli'd equipment, such as steam dryer, feedwater Fermi SSER 2-3-2
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sparger, incore guide tube, and steam separator.
The applicant has com-mitted to update the list and revise FSAR Section 3.2.2 as required.
(6) The applicant's response to the floor response spectra for a 5 percent structural damping and 5 percent equ'pment damping contained in his September 4,1981 letter is acceptabie.
Floor response spectra for lower equipment dumping, including 2 percent, are to be provided by the applicant (see Section 3.7.1 of this supplement).
When such floor response spectra are accepted by the staff, review of the applicant's reassessment program of equipment seismic qualification will be completed.
(7) The applicant has agreed to submit the results of torus-attached equipment qualification under the effects of combined seismic and Mark I hydro-dynamic loads for staff review before August 1, 1982, and to have all i
necessary modifications completed before fuel load or, if justified, before the plant returns to power after the first refueling outage.
The staff will continue to review the applicant's qualification program and report its conclusion on the acceptability of the implementation of the program in a future supplement to the SER.
3.11 Environmental Qualification of Safety-Related Electrical Equipment In Supplement No. 1 to the SER, the staff concluded that additional information was needed to complete its review.
The applicant has provided requested additional information except for that requested in Supplement No. 1 Section 6.3.4.1, " Safety Concerns Associated with Pipe Breaks in the BWR Scram System." The staff evaluation of this aspect of environmental qualification will be rcported in a future supplement.
The :,taf f's safety evaluation of environmental qualification for safety-related electrical equipment is provided in Appendix F to this supplement.
The staff concludes that conformance with the requirements stated in Appendix F to this supplement will ensure compliance with the Commission Memorandum and Order (CLI-80-21) of May 23, 1980 and, therefore, the applicant's environmental qualification program is acceptable.
The staff requires that outstanding information iocntified in Appendix F Section 3 and 4 be provided within 90 days of receipt of this supplement.
Qualification of electrical equipment located in a mild environment and mechanical equipment, as described in SRP Section 3.11, Revision 2, has not been investigated because the review was initiated prior to the issuance of the revised SRP.
The review of this equipment will be performed on the same schedule as for operating reactors.
The Commission Memorandum and Order directs that by no later than June 30, 1982 all electrical equpiment in operating reactors subject to a harsh enviroment be in compliance with NUREG-0588 or the " Guidelines for Evaluating Environmental
-Qualification of Class 1E Electrical Equipment in Operating Reactors." The-Fermi 2 plant must' fulfill these requirements prior to fuel load, based on the current schedule.
If fuel load occurs before the Commission deadline, justifi-cation for interim operation with equipment not fully qualified, if any, must Fermi SSER 2 3.
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be provided.
In addition, the applicant should provide an updated qualifica-tion status of all equipment undergoing test or replacement, and notify the staff prior to full power operation that maintenance and surveillance procedures 4
have been implemented.
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i 4 REACTOR j -
4.2 Fuel System Design 4.2.3 Design Evaluation 1
Channel Box Deflection 4
In the SER, the staff concluded that additional information was needed to j
provide assurance that fuel channel box deflection will not be a problem in Fermi 2 operation.
BWR fuel channel box bulging deflection is a result of inreactor creep which,-
if allowed to proceed unchecked, could eventually create an interference 4
between the control blade and the fuel channels.
In a General Electric generic report, NED0-21354, on chanel box mechanical design and deflection, GE l
describes a channel lifetime prediction method and a backup recommendation for
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periodic channel measurements that consist of settling friction tests.
The
. settling friction tests would provide an exact profile of control rod drive-friction versus position by measuring the hydrualic pressure under the drive piston as the drive'" settles" to latch position.
4 By letter dated January 15, 1982, the applicant proposed several steps that i
could be taken to preclude excessive channel box deflection, including the following:
(1) Before a new operating cycle is begun, control rod drive friction tests shall be performed for those core cells exceeding stipulated general-
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guidelines,(discussed below) or containing fuel channels with exposures j
greater than 30,000 mwd /t (associated fuel bundle exposures).
(2)
In lieu of friction testing, fuel channel deflection measurements may be 4
used to justify use of fuel channels exceeding 30,000 mwd /t exposure for a maximum of four additional operating cycles.
(3)
In the future, analytical channel lifetime-prediction methods, benchmarked I
and backed up hy periodic deflection measurements of samples of'the highest duty fuel channels, nay be~used to ensure clearance between control _ rod blades and fuel channels without additional' testing.
In addition, the applicant has' described some. administrative guidelines that
- would be followed concerning channel location, exposure, and: residence time.
While the review of the GE generic report, NED0-21354, has not been completed,
'the staff' agrees with the applicant that the proposed actions should preclude excessive channel bowing and that they provide reasonable assurance that channel bowing will not be a problem in Fermi 2.
Should the' staff's continuing generic; review of this phenomenon (and the GE report) reveal that a
' modification to the proposed steps is necessary, the licensee will be requested
.toftake the appropriate' action.
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I 4.4 Thermal and Hydraulic Design 4.4.1 Design Evaluation As reported in the Fermi 2 SER, the applicant committed to submit.the results of its evaluatior. of the degree of conformance of the loose parts monitoring system (LPMS) with Regulatory Guide 1.133, Revision 1 (May 1981).
By letter dated
. December 16, 1981, the applicant provided this evaluation of the LPMS.
4 The staff has reviewed the letter and concludes that th'e Fermi 2 LPMS design is in conformance with the Regulatory Guide 1.133, Revision 1.
However, the applicant also is required, after plant startup, to provide a report describing i
the implementation aspects of the LPMS program.
In particular, the report should describe the plant personnel training program, procedures for system calibration, and procedures for system cperation including determination of alert levels, and diagnostic procedures to confirm a loose part.
4 By letter dated January 22, 1982, the applicant has committed to submit a l
report on.the implementation aspects of the LPMS before commercial power j
operation.
In addition, specifications for the LPMS regarding the limiting condition for operation and surveillance requirements will be incorporated in l
the Fermi 2 Technical Specifications.
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The staff concludes that the Fermi ? LPMS design is acceptable.
The staff will verify that an acceptable training program and an operational procedure are implemented.
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ENGINEERED SAFETY FEATURES 6.2 Containment Functional Design 6.2.7 Containment Leak Testing
-In Supplement No. 1 to the SER, there were two open items:
one concerning the testing requirements for the traversing incore probe (TIP) system and the other concerning the bypass leakage paths.
By letter dated November 18, 1981, the applicant provided additional information that is evaluated below.
i Traversing Incore Probe System Testing Concerning the TIP shear valves, the applicant has agreed to expand the Fermi 2 Technical Specifications to include provisions for (1) Verifying the continuity of the explosive charge at least once every 31 days.
(2) Firing one of the explosive squib charges at least once every 18 months.
The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch which has been certified by having one of that batch successfully fired.
(3) Replacing all charges according to the manufacturer's recommended service l
lifetime.
These provisions are consistent with the recommendations established by the staff in Supplement No. 1 to the SER.
The staff, therefore, finds the TIP system testing program acceptable.
Bypass Leakage Branch Technical Position (BTP) CSB 6-3* defines bypass leakage as being that leakage from the primary containment which can circumvent the secondary con-tainment boundary and escape directly to the environment; that is, bypassing the leakage collection and filtration systems of the secondary containment.
The applicant had ccamitted, as mentioned in Supplement No. 1 to the SER, to a maximum bypass leakage rate equal to 4 percent of L, where L is the maximum A
allowableleakrateintheTypeAcontainmentinteghatedleakratetest.'
For i
i the valves listed in the bypass test program, the staff had imposed the follow-ing requirements:
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(1)
Each bypass leakage path must contain two valves in series located inside the secondary containment.
- BTPs are in the staff's Standard Review Plan, NUREG-0800.
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1 (2) Each valve should be tested in accordance with the specifications of either Appendix J to 10 CFR Part 50 or Section XI of the ASME Code.
(3) Valve sets should be supplied with power from both Division I and Division II.
(4) Valves should receive diverse isolation signals.
Since the issuance of Supplement No. I to the SER, the applicant has revised his bypass leakage program.
The applicant has proposed to designate the high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) steam supply penetrations as the bypass paths for each system rather than the smaller steam drain lines that exit the secondary containment and terminate at the main condenser.
This requirement will be met without increasing the total bypass leakage permitted.
The staff finds this provision acceptable.
The other bypass leakage paths of concern are the pump discharge to the con-densate storage tank lines in the HPCI and RCIC systems.
The applicant has stated that the RCIC system line meets the requirements set forth above.
The HPCI system line meets all the requirements listed except that all HPCI valves receive only Division II power instead of being powered by both divisions.
1 However, the path from the primary coolant pressure boundary to the condensate i
storage tank contains three valves:
V8-2194, V8-2198, and V8-2200.
These valves are interlocked so that V8-2194 must be closed before V8-2198 can be opened.
The applicant has proposed adding V8-2194 to the list of valves to be tested for bypass leakage.
This is a reasonable alternative to the requirement of divisional power because in the event Division II power is lost while valve V8-2198 is open, the interlock will ensure that valve V8-2194 had been closed and, consequently, that the bypass path had been eliminated except for leakage beyond closed valve V8-2194.
The staff finds this an acceptable approach.
The remainder of the valves in the bypass leakage test program are containment isolation system valves and will be tested according to test criteria in Appendix J to 10 CFR Part 50.
The staff, therefore, finds the bypass leakage test program acceptable.
t 6.3 Emergency Core Cooling System (ECCS) 6.3.4 Evaluation Findings In the SER, the staff concluded that the core spray system for Fermi 2 provided adequate core spray flow and distribution for interim plant operation and that additional evaluation and testing were required on a generic basis.
The staff recently reviewed core spray distribution information on Japanese spray distribution tests of a simulated BWR/S configuration in steam using a 60 -sector test facility.
The test data show that central bundles receive low i
core spray flow because of maldistribution.
Although no specific data are l
available, the staff has been advised that the 360 model tests by the Japanese i
with 5/6 of_the spray nozzles blocked gave results similar to those of the i-60 -sector tests.
This information was of concern because credit is taken for l
core spray heat transfer using a minimum spray to each bundle in the GE ECCS Evaluation Model.
This results in a heat transfer coefficient of Fermi SSER2 6-2 i
2 1.5 Btu /hr-ft - F for core spray heat transfer, which is the minimum value specified in Appendix K to 10 CFR Part 50.
The Japanese data are not the first to show low flow to some sections of BWR cores.
Thus, as described above, the staff has previously considered the effect of low core spray flow to individual channels on calculated peak clad temperature (PCT).
In the staf f evaluation of General Electric Co. Topical Report NEDO-20566 Amendment 3,* it was concluded that minimum spray flow to any channel following a LOCA would not be less than half of the design flow that was demonstrated to be available by tests and calculations.
The tests and calculations did not include steam effects on nozzle spray patterns and flow l
rate.
Based en measurements of minimum bundle spray flow for each BWR size and type for one sparger only, in air, the minimum flow for BWR/2 through BWR/5 designs was calculated to be 1.3 times the flow necessary to remove decay heat by vaporization (reference flow).
Thus, the steam effects on spray distribution would not result in less than 0.65 times the minimum reference flow (or 1.3 times with both spray spargers operating).
BWR FLECHT data reported by Schraub and Leonard, June 1979* show little degradation in heat transfer for flow as low as 0.38 times the reference flow, or approximately 1 gpm.
As far as the staff has been told, the minimum flow observed for any bundle in the Japanese 60 -sector tests was 1 gpm.
The heat transfer coefficients in the GE ECCS Evaluation Model are based on the FLECHT data, and a minimum bundle flow of 1 gpm would justify the heat transfer coefficient for 2
core spray cooling (1.5 Blu/hr-ft - F) used in that model.
During the BWR core spray injection, spray injected in the upper plenum will either be distributed to the core or bypass the core and drain to the lower plenum region, which results in a rapid bottom reflooo rate.
Credit was not taken for this rapid bottom reflood effect in the GE ECCS model.
Any liquid in excess of the minimum required fer core spray heat transfer was assumed lost from the system and did not contribute to the reflood.
Preliminary results from counter-current flow-limiting (CCFL) tests performed in the 30 -Sector Stream Test Faiclity at Lynn, Massachusetts show that the spray flow injected in the upper plenum actually drains to peripheral bundles and increases the botton reflood rate.
In response to a staff request, GE presented results for reanalysis of limiting BWR/4 and BWR/S cases to assess the effect of no core spray cooling on the peak clad temperature, assuming that the core spray coolant drain; to the lower plenum and increases the reflood rate, as observed in the Lynn test.
The calculated peak clad temperature did not exceed the 10 CFR 50.46 limit of 2200 F with no credit taken for the spray cooling effect.
The staff concludes that the new information from the Japanese tests does not pose a safety concern for BWR/4 and BWR/5 reactors for the following reason 3:
(1) Only preliminary and incomplete data sre available from the Japanese tests, and it is impossible to draw final conclusions from them at this time.
(2) Core spray flow maldistributions resulting in flows on the order of 1 gpm l
per bundle (apparently consistent with those obtained in the Japanese 60 -
^See Appendix B.
Fermi SSER2 6-3 l
l l
sector tests) would remain consistent with the core spray cooling assumptions employed in the present GE ECCS Evaluation Model.
(3) New analyses performed by GE have shown that for limiting BWR/4 and BWR/5 cases with core spray assumed to blow down peripheral channels to increase the reflood rate as observed in the Lynn tests, the calculated peak clad temperature did not exceed the 10 CFR 50.46 limit of 2200 F with no credit taken for the spray cooling effect.
Fermi SSER2 6-4
9 AUXILIARY SYSTEMS 9.5 Fire Protection, Communication, Lighting arid Emergency Diesel Engine Systems 9.5.1 Fire Protection In the SER, the staff stated that it had not completed its review of the fire protection for the control room.
The applicant has provided additional infor-mation, and the staff has completed its review.
The staff's safety evaluation of fire protection for the Fermi 2 plant is provided in Appendix E to this suppiement, which supersedes Appendix E to the SER.
Based on its review, the staff concludes, that when the committed modifications have been completed, the fire protection program for Fermi 2 will meet the technical requirements of Appendix R to 10 CFR 50 with approved deviations, the guidelines of Appendix A to BTP ASB 9.5-1, and the requirements of GDC 3, and is, therefore, acceptable.
The staff will condition the operating license to require completion of modifications by a specified date.
9.5.7 Emergency Diesel Engine Lubricating Oil System In the SER, the staff stated it would evaluate proposed modifications to the lubricating oil system in a supplement to the SER.
By letter dated August 31, 1981, the applicant described his proposed modifications.
The emergency diesel generator system for the Fermi 2 nuclear plant consists of four Fairbanks Morse (FM) Model 3800 TD 8-1/8 opposed piston diesels.
According to the applicant, the design of the FM diesels was such that continuous lubrica-tion of the wearing parts when the engine is on standby service is not feasible.
The FM diesel is provided with a keep-warm lubrication system that operates continuously when the engine is )n standby service.
This system does not lubricate the engine wearing parts.
Therefore, when the engine is on standby service for a prolonged period of time, lubricating oil drains from the wearing parts and the engine is subject to a dry start.
Dry starts of these diesel generators under emergency conditions would result in momentary lack of lubrica-tion at the various moving parts, which could eventually lead to failures, with resultant equipment unavailability.
The applicant proposes to install the manufacturer's modification to the lube oil system.
The modification will result in the continuous lubrication of the lower portions of the engine.
The upper portions of the engine will not be prelubricated in tia standby condition.
Lubricating the upper portions of the engine in the standby condition would result in excessive amounts of oil col-lecting in the cylinders, which is an unacceptable condition to the manufacturer.
However, the modification will provide for partial filling of the upper lube oil supply header and a lube oil booster / accumulator system that will force lube oil into the upper lube oil header during starting.
Therefore, the wearing Fermi SSER2 9-1
l parts in the upper portion of the engine will get lubricated immediately upon engine starting.
Therefore, the staff finds this modification to be acceptable.
Hence the staff concern about dry starting is alleviated.
Based on its review, the staff concludes that the emergency diesel engine lubricating oil system as modified meets the requirements of GDC 2, 4, 5, and 17; meets the guidance of the cited Regulatory Guides and SRP Section 9.5.7; can perform its design safety function; and meets the recommendations of l
j NUREG/CR-0660 and industry codes and standards.
It is, therefore, acceptable.
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i Fermi SSER2 9-2
13 CONDUCT OF OPERATIONS 13.5 Industrial Security In Supplement No. I to the SER, the staff stated that the Fermi 2 revised safe-guards contingency plan and g;ard training plan were acceptable and that the revised physical security plan was under review.
The staff has reviewed these plans in accordance with Section 13.6, " Physical Security," of the July 1981 edition of the " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants" (SRP, NUREG-0800).
As a result of its evaluation, the staff identified certain portions of these plans as requiring additional information and upgrading to satisfy the requirements of Section 73.55 and Appendices B and C of 10 CFR Part 73.
The applicant filed revisions to these plans that satisfied these requirements.
The staff concludes that the revised plans comply with the Commission's regulations contained in 10 CFR Parts 50 and 73.
The approved plans, which contain information described in 10 CFR 73.21, are entitled "Enrico Fermi Atomic Power Plant Unit 2 Physical Security Plan," Revision 3, dated December 8, 1981; "Enrico Fermi Atomic Power Plant Unit 2 Safeguards Contingency Plan," Revision 1, dated July 20, 1981; and "Enrico Fermi Atomic Power Plant Unit 2 Security Personnel Training and Qualification Plan,"
Revision 1, dated July 20, 1981.
The operating license will be conditioned to require the licensee to fully implement and maintain in effect all provision of these approved plans.
An ongoing review of the progress of the implementation of these plans will be performed by the staff to ensure conformance with the performanco.2quirements of 10 CFR Part 73.
The identification of vital areas and measures used to control access to these areas, as described in the plan, may be subject to amendments in the future.
The applicant's security plans are being protected from unauthorized disclosure in accordance with Section 73.21 of 10 CFR Part 73.
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l Fermi SSER2 13-1
16 TECiiNICAL SPECIFICATIONS In the SER, the staff identified certain conditions which must be included in the Fermi 2 Technical Specifications in addition to Standard Technical Specifi-cations.
In this supplement, the staff has identified the need for additional plant-specific Technical Specifications.
The complete list of plant-specific Technical Specifications is given belo ; they are discussed further in sections of the SER or this supplement as indicated.
Technical Specifications Identified in the SER (1) Shore barrier annual inspection (2.4.2.5)
(2) Leakage tests for pressure isolation valves (3.9.6.2)
(3) Core performance (4.4.1)
(a) Prohibition of natural circulation and single-loop operation (b) Surveillance of core flow and flow-biased scram (4) Augmented surveillance tests of control rod drives (4.5)
(5) Periodic venting of ECCS and RCIC discharge lines (5.4.1, 6.3.4)
(6) Surveillance tests of low pressure coolant injection system (5.4.2)
(7) Surveillance tests of torus-to-drywell vacuum breakers and drywell-to-torus vents (6.2.1)
(8) Surveillance tests of secondary containment inleakage and drawdown time (6.2.3)
(9) Limiting containment purge to 90 br/yr (6.2.4)
(10) Surveillance tests of mainsteam isolation valve leakage control system (6.2.5)
(11) Surveillance tests of selected control systems (7.1.2)
(12) Surveillance tests of high reactor vessel level trip (7.7.2)
(13) Surveillance tests of swing bus transfer (8.3.1)
(14) Limitation on battery charger operation (8.3.2)
(15) Reg.ctorbuildingcraneinspectionandmaintenance(9.1.4)
(16) Monitoring of spent fuel rack material (9.1.5)
Fermi SSER2 16-1
(17) Surveillance tests and inspection of turbine stop valves, turbine control l
valves, and turbine bypass valves (7.7.2, 10.2.1)
(18) Response time tests of thermal power monitor (15.1)
(19) Prohibition of operation with partial feedwater heating (15.1)
(20) Annual reports of failures of relief and safety valves to close and of outages of ECCS equipment (22.2, Items II.K.3.3 and II.K.3.17)
Additional Technical Specification Identified in Supplement No. 1 to the SER (1) Leakage testing of containment isolation valves and valve operators (SER Supplement No. 1 Section 6.2.7)
Additional Technical Specifications Identified in this Supplement (1) Limiting conditions for operation and surveillance tests of the loose parts monitoring system (4.4.1)
(2) Requirement to lock cabinet enclosing indicate storage tank level instrumentation (22.2, II.K.3.22)
Fermi SSER2 16-2
22 TMI-2 REQUIREMENTS 22.2 TMI Action Plan Requirements for Applicants for Operating Licenses II.
Siting and Design II.B.3 Postaccident Sampling Capability DpeussionandConclusions In the SER, the staff stated that the operating license would be conditioned to require submittal of information in sufficient detail for the staff to evaluate compliance with this TMI-2 requirement.
The applicant has now provided detailed information.
The staff evaluation of this information is provided herein.
This section cupersedes the Discussion and Conclusions regarding post-accident sampling provided in the SER.
In Amendment 36 and in letters dated June 24, 1981 and December 18, 1981, the applicant has committed to install a postaccident sampling system that meets the requirements of item II.B.3 in NUREG-0737 and meets the sampling and analysis requirements of Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," Revision 2.
The applicant will provide the capability j
under accident conditions, to obtain and analyze samples, within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> of the time a decision is made ta sample, from the reactor coolant system, suppression pool, and containment atmosphere, and to obtain grab samples for offsite analyses.
The applicant will also provide onsite radiological and chemical
'i -
analysis capabilities to quantify hydrogen, oxygen, and radionuclide concentra-tions in containment atmosphere; radioactive isotopes (noble gases, iodine Eml I
cesium isotopes, and nonvolatile isotopes); and dissolved gases, pH chloride, and boron concentrations in liquid samples.
Offsite analysis capabilities are provided to perform chloride analysis of undiluted samples within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />.
Dissolved oxygen is verified at less than 0.1 ppm by determination of a positive hydrogen residual initially and direct oxygen analysis, consistent with as low as is reasonably achievable (ALARA) requirements.
Provisions are incorporated to reduce background radiation levels so that radionuclide i
analysis can be performed in the approximate range of 1 pCi/g to 10pCi/g.
Sample system testing and operator training will be conducted on a semi-annual basis.
Sample lines are routed to a shielded sample station in an accessible i
area.
An isolated auxiliary system is not required to be operational in order to use the postaccident sampling system.
A bellows' pump will be incorporated to permit sampling containment atmosphere under both positive and negative pres-sures.
Provisions are incorporated to purge the sample lines for reducing plate out, for minimizing sample loss or distortion, for preventing blockage of sample lines, and for appropriate disposal of samples, and 3/8 in. sample lines with environmentally qualified containment isolation valves are included as flow restrictors.
Sufficient shielding will be provided to meet the requirements of General Design Criterion (GDC) 19 in Appendix A to 10 CFR Part 50, assuming the source term defined in Regulatory Guide 1.3, " Assumptions Used for Evaluating Fermi SSER2 22-1
i the Potential Radiological Consequences of a Loss-of-Coolant Accident for Pressurized Water Reactors." Man motion / radiation exposure studies have been performed to demonstrate that sampling, transport, and analysis can also be completed consistent with GDC 19.
Provisions are incorporated to return sample flush volumes to the suppression chamber, and ventilation exhaust treatment is incorporated in the sample station to control gaseous radionuclide releases.
All equipment in the postaccident sampling facility that may be required to be operational during an accident, will have access to onsite backup power supplies.
The sampling system and containment isolation valves will be environ-mentally qualified.
By design, high pressure carrier gas in chromatographic analyses cannot enter the reactor coolant system.
The staff was concerned about the applicant's ability to obtain samples that are representative of the reactor core and suppression pool conditions.
The staff's specific concerns were i
(1) That the reactor coolant liquid sample which is taken from the downcomer area (jet pump diffuser or reactor heat removal suction) will be diluted to an uncertain degree by the reactor coolant system makeup water source.
This condition occurs when low volumes of steam are being generated, which significantly reduces the amount of moisture which leaves the core and is subsequently returned to the downcomer via the moisture separators.
This condition can result in the samples being analyzed at lower concentrations of soluble species (chloride, boron, iodine, cesium, etc.) than are actually present in the core area.
(2) That the suppression chamber samples, due to the location of the sample point relative to reactor coolant system safety valve discharge points, will either be excessively diluted or virtually undiluted resulting in erroneous estimates of core damage.
The staff has requested that the i
applicant provide information to demonstrate that these sample points are-located so that adequate mixing will occur and the samples are representa-tive of the mixture rather than only the discharged fluid.
By letter dated December 18, 1981, the applicant provided the results of an evaluation performed in conjunction with the nuclear steam supply system vendor.
These data demonstrate that at power levels down to 1 percent, samples from the reactor vessel shroud will be representative of core conditions.
At power levels of less than 1 percent, a sample that is representative of core conditions will be obtained by increasing the reactor water level by 18 in.,
which will fully flood the moisture separators, to provide a thermally. induced recirculation finw path for mixing.
In the event of large leaks where reactor
. vessel water level cannot be maintained, reverse flow thrcugh the core to the suppression pool is provided.
Representative samples will be obtained from the residual heat removal pump discharge, which will be lined up to take a suction from the suppression pool.
To ensure that suppression pool samples are rep-resentative, the applicant will operate the residual heat removal pumps for 30 minutes prior to taking a sample.
Additionally, he has conducted an evalua-t tion of safety valve discharge locations in the suppression pool and determined that they are located at positions which will facilitate obtaining a representa-tive sample.
Fermi SSER2 22-2
The applicant has submitted a summary of postaccident chemistry analytical procedures.
The staff is conducting a generic review of accuracy and sensi-tivity for analytical procedures and online instrumentation to be used for post-accident analysis.
If the staff generic review determines that a specific procedure is unacceptable, the staff will require the applicant to make appropriate mcdifications, as determined by the generic review.
Based on the above evaluation, the staff concludes that the provisions in the proposed postaccident sampling facility meet the sampling and analysis require-ments of Item II.B.3 in NUREG-0737.
However, the applicant has not completed the procedure for relating radioactive isotopes to estimated reactor core damage.
By letter of December 18, 1981, the applicant has committed to pro-vide, prior to May 1, 1982, an interim procedure for estimating core damage and a schedule for submitting the final procedure.
Prior to issuance of an operating license, the staff must evaluate and find acceptable the interim procedure for estimating core damage.
The staff will report its evaluation of the interim procedure in a future supplement to the SER.
The operating license will be conditioned to require staff acceptance and implementation of the final procedure within a specified time interval.
Upon completion of the generic review for chemistry procedures, the applicant will be required to make appropriate modifications to ensure that his chemistry pro-cedures are acceptable in the postaccident chemical / radiochemical environment.
Completion of the generic review is not a condition for full power licensing.
1 II.K.3.22 Automatic Switchover of Reactor Core Isolation Cooling System Suction - Verify Procedures and Modify Design Discussion and Conclusion In the SER, the staff concluded that the design of the automatic switchover of the reector core isolation cooling (RCIC) pump suction meets the requirements of NUREG-0737 and that the staff would verify installation.
During an inspection of the installation, the staff questioned the applicant as to whether the instru-mentation and controls used for automatic switchover of the RCIC suction meet applicable NRC requirements.
This equipment is also used for automatic switchover of the high pressure coolant injection (HPCI) suction.
By letter dated December 1, 1981, the staff asked the applicant to clarify how the automatic switchover instrumentation for the RCIC and the HPCI systems meet applicable NRC requirements regarding protection from freezing, seismic design, and quality assurance.
l By letter dated January 6, 1982, the applicant provided a description showing l
how the switchover instrumentation meets the Instrumentation and Control System Branch (ICSB) position on the inoperability of instrumentation as a result of extreme cold weather, a justification for this instrumentation's nonseismic location, and a description of the quality level for this instrumentation.
The applicant also committed to lock the door on the cabinet enclosing the instrumenta-tion.
The staff will include in the plant Technical Specifications that the cabinet door be kept locked.
Fermi SSER2 22-3
Based on the information submitted by the applicant, the staff has concluded that the Fermi 2 design for the HPCI and RCIC automatic switchover is accept-able.
II.K.3.44 Evaluation of Anticipated Transients with Single Failure To Verify No Fuel Failure Discussion and Conclusions In the SER, the staff indicated that the results of the BWR Owners Group study was an acceptable response to this item, subject to the staff's evaluation of the results and their applicability to Fermi 2.
The staff has completed its review of this item.
By NRC Generic Letter No. 81-32, dated August 7,1981, the staf f provided its generic evaluation of the BWR Owners Group report.
In that evaluation, the staff concluded:
Since the calculations show that the core does not uncover for the worst transient combined with the worst single failure and a stuck open relief valve, and since test data from TLTA* show the calcula-tions to be conservative for an ADS
- blowdown, we find the generic responses of the BWR Owners' Group to be acceptable.
Individual licensees / applicants who reference the generic response to Item II.K.3.44 must verify that the assumptions and initial condi-tions used in the analyses are applicable or are bounding for their specific plants.
By letter dated November 18, 1981, the applicant provided acceptable verifica-tion that the assumptions and initial conditions used in the generic analyses are representative for the Fermi 2 plant.
The staff concludes that Item II.K.3.44 has been acceptably completed for Fermi 2.
III Emergency Preparedness and Radiation Protection III.A.2.2 Improving Licensee Emergency Preparedness--Long Term Discussion and Conclusions In the SER, the staf f stated that it would report on applicant's meteorological measurements program, meterolcgical model, and dose calculation method when additional information was provided.
i'-teorological measurements for postaccident monitoring are derived from the cr;ite 60-meter and 150-meter tower.
In addition, four offsite towers, part of speed and direction information as needed.
" automatic depressurization system."
Fermi SSER2 22-4
By letter dated December 16, 1981, the applicant provided reports on the lake breeze analysis method and the manual dose assessment method.
These reports are under review.
Results of the staff review will reported in a future supplement.
The staff evaluation of the Fermi 2 emergency preparedness meteorological measurements program, as described in FSAR Amendment 33, follows.
The Monroe ambient air quality monitoring network (MAAQMN), will provide wind Both the primary and duplicate systems on the 60-meter tower and backup 150-meter tower will have redundant power sources and data transmission capability to the con-trol room, technical support center, and emergency operations facility.
Additionally, a remote interrogation capability will allow access to the meteorological data from off site.
Measurements made on the 60-meter primary tower at both the 10-meter 1evel and 60-meter level include:
wind speed, wind direction, air temperature, and sigma theta.
Measurements of temperature differences between the 10-meter level and the 60-meter level to determine Pasquill stability classification are also made.
Miscellaneous measurements include precipitation at ground level and dew point temperature.
Precipitation and dew point temerature are available from the 4
]
primary tower only.
J On the 150-meter tower, wind speed, wind direction, and sigma theta are measured at 10 meters.
On the MAAQMN towers, wind speed and wind direction are measured at 10 meters.
Based on its review, the staff concludes that the meteorological measurements program for emergency preparedness satisfies the measurement and interrogation a
requirements of NUREG-0654, Appendix 2.
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Ferm 22-5
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I.
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APPENDIX A i
CONTINUATION OF THE CHRONOLOGY OF THE RADIOLOGICAL SAFETY REVIEW OF THE DETROIT FDISON COMPANY APPLICATION FOR AN OPERTING LICENSE s
FOR THE ENRICO FERM0 ATOMIC September 18, 1981 Letter from applicant concerning reheater bypass flow j
analysis.
4 September 18, 1981 Letter from applicant concerning equipment environ-mental qualification radiation profiles, i
September 23, 1981 Letter from applicant concerning Fermi 2 equipment 4
environmental qualification.
September 25, 1981 Letter to applicant concerning review of security personnel training and qualification plan.
September 25, 1981 Letter to applicant transmitting two copies of Supple-ment No. 1 to the Fermi 2 SER.
Septeraber 25, 1981 Letter to applicant requesting additional information on the physical security plan.
September 28, 1981 Letter from applicant concerning operational radiological monitoring program.
September 29-Representatives from applicant and staff meet at October 1, 1981 the Fermi 2 plant in Monroe County, Michigan, to observe construction, and obtain information regarding completion of construction and preoperational tests for use by NRC Forecast Panel (summary issued i
November 24,1981).
October 2, 1981 Letter to applicant concerning submittal of additional information.
October 5, 1981 Letter from applicant concerning applicant's position on Regulatory Guides 1.127, 8.13, 8.14, 8.15 and 8.26.
[
October 7, 1981 Letter to applicant transmitting 20 copies of Supplement No. 1 to the Fermi 2 SER.
The copies for-warded are the bound printer-contractor copies.
I October 19, 1981 Letter from applicant transmitting the safety relief valve operability test report.
Fermi - SSER2 A-1
1 October 20, 1981 Meeting summary for meeting held July 27-31, 1981, at the Fermi site to audit the results for seismic and dynamic qualification.
October 23, 1981 Letter to applicant requesting additional information evaluation of emergency response facilities.
October 23, 1981 Summary of meeting held September 16, 1981 regarding the opperating license review.
October 27, 1981 Letter from applicant concerning seismic design bases for computing combined responses.
October 29, 1981 Representatives from staff and applicant meet in Bethesda, Maryland to discuss method for generating 5 percent structural damping curves for seismic reassessment (summary issued November 3, 1981).
October 29, 1981 Summary for meeting held July 13-17 at Fermi 2 site to evaluate and document the environmental qualification.
October 30, 1981 Letter from applicant concerning response to requests for additional information in Fermi 2 operating license application.
November 5, 1981 Representatives from staff, applicant, NUS, and S&L meet in Bethesda, Maryland to discuss applicant's plans for on-site storage of low level radioactive waste (summary issued November 10, 1981).
November 13, 1981 Letter from applicant concerning postaccident sampling licensing conditions response.
Novdmber 18, 1981 Letter from applicant concerning NUREG-0798 open items.
November 18, 1981 Letter from applicant concerning NUREG-0737 Item II.K.3.44, Evaluation of Anticipated Transients Combined with Single Failure (Generic Letter 81-32).
November 18, 1981 Letter from applicant concerning qualification of inspection, examination and testing, and audit person-nel (Generic Letter 81-01).
November 20, 1981 Representatives from staff and applicant meet in Bethesda, Maryland to discuss applicant's proposed response to staff's evaluation and requests for additional information regarding Fermi 2 Emergency Response Facilities (summary issued December 14, 1981).
November 23, 1981 Letter from applicant concerning IE Bulletin 79-15, Long-Term Operability of Deep Draft Pumps.
Fermi SSER2 A-2
December 1, 1981 Letter to applicant concerning automatic switchover of the suctions for the HPCI and RCIC Systems.
December 2, 1981 Representatives from staff and applicant visit the Fermi 2 plant to audit documentation showing operability of purge valves and view installation (summary issued January 22, 1982).
December 3, 1981 Letter to applicant requesting additional information regarding seismic and dynamic qualification of mech-anical and electrical equipment important to safety.
December 3, 1981 Letter from applicant concerning control of heavy loads over or in proximity of irradiated fuel.
December 4, 1981 Representatives from staff and applicant meet in Bethesda, Maryland to discuss applicant's bases for fire protection for the Fermi 2 control room.
(summary issued January 20, 1982).
December 9, 1981 Letter to applicant concerning NUREG-0737 Item II.K.3.44, Evaluation of Anticipated Transients Combined with Single Failure.
December 14-17, 1981 Representatives from staff and applicant meet at the Fermi 2 plant to conduct an audit of environmental qualification documentation (summary issued January 26, 1982).
December 16, 1981 Letter from applicant concerning methodology for manual dose assessment for emergency preparedness and methodology for site-specific lake breeze analysis.
December 16, 1981 Letter from applicant concerning SER commitment on loose parts monitoring system.
December 21, 1981 Letter to applicant transmitting an Order extending the latest construction completion date for Fermi 2 from January 1, 1982 to June 30, 1984.
December 23, 1981 Letter from applicant concerning seismic qualification review team (SQRT) audit, response to open items.
December 31, 1981 Letter from applicant transmitting Amendment 40 to the FSAR.
January 4,1982 Letter from applicant concerning purge valve operability.
January 4,1982 Letter from applicant submitting additional information on fire protection.
Fermi SSER2 A-3
January 8,1982 Letter from applicant submitting additional information on siesmic reassessment (equipment necessary to achieve safe shutdown).
January 6, 1982 Letter from applicant concerning automatic switchover of the suctions for the HPCI and RCIC systems.
January 8, 1982 Letter from applicant submitting additional information on seismic reassessment (equipment necessary to achieve safe shutdown).
January 13, 1982 Letter from applicant transmitting Fermi 2 Emergency Plan Implementing Procedures (Draft).
January 14, 1982 Letter from applicant transmitting additional information on SQRT Audit Open Items Nos. 6 and 7.
January 15, 1982 Letter from applicant transmitting additional information AQRT Audit Follow-Up Action No. 3.
January 22, 1982 Letter from applicant concerning diesel engine lubrication aystem modifications.
January 22, 1982 Letter from applicant committing to provide a report on implementation of a loose parts detection program.
Fermi SSER2 A-4
1.
APPENDIX B 4
1 a
BIBLIOGRAPHY Documents referenced on or used to prepare this supplement may be obtained as indicated on the inside front cover.
Correspondence between the Commission and the applicant (including the applicant's Final Safety Analysis Report and l
Environmental Report), as well as the Commission's rules and regulatory guides, vendor documents, and industry codes and standards, may be inspected at the NRC Public Document Room, 1717 H St., NW, Washington, DC 20555.
Correspondence between the Commission and the applicant may also be inspected at the Monroe County Library, 3700 South Cester Road, Monroe, MI.
4 Applicant and Vendor Documents Detroit Edison Co.,'" Evaluation of Selected Control Panel Components Subjected to a Postular Exposure Fire," November 1981.
General Electric, Generic Report, NED0-21354, September 1976 General Electric Co., Topical Report, APED-5529, " Core Spray and Core Flooding i
Heat Transter Effectiveness in a Full-Scale Boiling Water Reactor," F. A. Schraub l
and J. E. Leonard, Janury 1979.
i General Electric Co., Topical Report, NE00-20566 Amendment 3, 'GE Analytical
'Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, "Effect of Steam Environment on Core Spray Distribution," April-1977.
Industry Codes and Standards American Society of Test Engineers, E-119, " Fire Test of Building Construction
'and Materials."
l-Institute of Electrical and Electronics Engineers Standards 334-1971, 832-1972,
-383-1974.
National Fire Protection Association Standards, 12, 12A, 13, 14, 15, and 20.
U.S. Nuclear Regulatory Commission NUREG Reports
]
0588-Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment, December 1979.
i 0654-Criteria for Preparation and Evaluation of Radiological Emergency Response-Plans and Preparedness in Support of Nuclear Power Plants, Revision 2.
4 0737-Clarification of TMI Action Items, November 1980.
0800-Standard Review Plan (formerly NUREG-75/087), July 1981.
Fermi SSER2 B-1
U.S. Nuclear Regulatory Commission Regulatory Guides 1.101 Emergency Planning for Nuclear Power Plants 1.97 Instrumentation for Light-Water-Cooled Nuclear Rev. 2 Power Plants To Assess Plant and Environs Conditions During and Following an Accident.
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i APPENDIX C NRC STAFF CONTRIBUTORS AND CONSULTANTS This supplement to the SER is a product of the NRC staff and its consultants.
The NRC staff members listed below were principal contributors to this report.
A list of consultants follows the list of staff.
NRC STAFF Name Title Branch R. K. Anand Systems Engr.
Chemical Engineering R. Giardina Reactor Systems Engr.
Power Systems i
Y. Gene Hsii Sr. Nuclear Engr.
Core Performance J. E. Kennedy Sr. Equipment Equipment Qualification Qualification Engr.
D. Kunze Plant Protection Analyst Physical Security Licensing J. C. Lane Containment Systems Containment Systems Engr.
A. Lee Sr. Mechanical Engr.
Equipment Qualification W. T. LeFave Sr. Auxiliary Systems Auxiliary Systems Engr.
J. Levine Meteorologist Hydrologic and Geotechnical Engineering J. Mauck Reactor Engr. -
Instrumentation & Control Instrumentation Systems C. McCracken Sr. Chemical Engr.
Chemical Engineering i
B. Siegel Reactor Engr.
Operating Reactors Br2 S. B. Sun Nuclear Engr.
Core Performance C. P. Tan Structural Engr.
Structural Engineering M. Tokar Reactor Engr.
Core Performance i
CONSULTANTS Name Organization J. Behn Gage-Babcock & Assoc.
M. Hinton EG&G Idaho, Inc.
J.'R'se EG&G Idaho, Inc.
P. Brown Brookhaven National Laboratory J. Curreri Brookhaven National Laboratory A. J. Philippacopoulos Brookhaven National Laboratory M. Reich Brookhaven National Laboratory S. Sharma Brookhaven National Laboratory M. Subudhi Brookhaven National Laboratory Fermi SSER2
.C-1
l APPENDIX D ERRATA TO SUPPLEMENT NO. 1 TO THE FERMI-2 SER Page 1-2, Line 5:
Change "Rothburg" to "Rothberg" Line 6:
Change " Tam" to " Tan" Line 10:
Change "Anaud" to "Anand" Line 11:
Delete "R. Giardina" Page 3-4, Line 15:
Change "this" to "the" Page 3-7, Line 17:
Change "this supplement" to "the SER" Page 6-8, Line 25:
Change "(c)" to "(3)"
Page 18-2, Line 6:
Change "3.1(c)" to "13.1" Page 18-3, Line 5:
Change " licensed" to " issued" Page A-2, Line 27:
Change "NRC to act" to "W. Colbert (Detroit Edison Company) to commit" Page A-3, Line 9:
After this line insert " June 10, 1981 letter from applicant transmitting Amendment No. 36 to the amended and substituted application for licenses."
Page A-6, Line 18:
Change " seismic qualification review audit" to " analysis of buried piping."
Fermi SSER2 D-1
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APPENDIX E*
FIRE PROTECTION REVIEW FERMI 2 BY THE OFFICE OF NUCLEAR REACTOR REGULATION CHEMICAL ENGINEERING BRANCH I.
INTRODUCTION The staff has reviewed the Fermi 2 fire protection program reevaluation and fire hazards analysis submitted by the applicant dated October 1977, including Amendment 40.
The Enrico Fermi Atomic Power Plant is a on?-unit site.
The Ferrni 2 reevaluation was in response to a staff request for the applicant to review the fire protection program against the guidelines of Appendix A to Branch Technical Position (BTP) ASB 9.5-1, " Guidelines for Fire Protection for Nuclear Power Plants." As part of its review, the staff visited the plant site to examine the relationship of safety-related components, systems, and struc-tures in specific p ant areas to both combustible materials and to fire detec-l tion and suppression systems.
The overall objective of this review was to ensure that in the event of a fire at Fermi 2, personnel and the plant equipment would be adequate to safely shut down the reactor, to maintain the plant in a safe shut-down condition, and to minimize the release of radioactivity to the environment.
The review included an evaluation of the automatic and manually operated water and gas fire suppression systems, the fire detection systems, fire barriers, fire doors and dampers, fire protection administrative controls, and the fire j
brigade size and training.
On October 27, 1980, the Commission approved for publication in the Federal Register a new rule, S50.48, and its Appendix R to 10 CFR Part 50, delineating 2
certain fire protection provisions for nuclear power plants.
The staff used the contents of this rule in the evaluation of.the Fermi 2 fire protection program.
The applicant has been informed that all fire protection modifications have to be implemented prior to fuel load.
The staff consultants, Gage-Babcock and Associates, Inc., participated in the review of the fire protection program and in the preparation of this safety evaluation report, and concur with the staff findings.
II.
FIRE PROTECTION SYSTEMS DESCRIPTION AND EVALUATION A.
Water Supply Systems The water supply system consits of two fire pumps connected to a 12-in. carbon i'
- This appendix supersedes Appendix E to the SER.
Fermi SSER2 E-1
steel, coated and wrapped pipe yard main loop.
There is one electric-motor-and one diesel-driven fire pump, and both are rated at 2500 gpm at 150 psig head each.
The fire pumps are UL listed.
The controllers are not listed but meet the general design and function requirements of listed controllers.
Their design and installation conforms to general guidelines of National Fire Protection Association Standard (NFPA) 20, " Standard for the Installation of Centrifugal Fire Pumps." The pumps are located in se general service water pump house with the diesel fire pump enclosed in a 2 nour fire rated enclosure with automatic sprinklers.
The pumps take suction from the general service water pump header which is supplied from Lake Erie.
The fire main system is connected and is pressurized by the general service water system but is capable of complete isolation from the general service water system with a check valve provided to prevent flow from the fire main loop into the general service water system.
The pumps discharge into two separate connections to the underground 12-in. yard main 100p.
'?
The general service water pumps operate continuously, maintaining a pressure of 150 psig.
The fire pumps start automatically on low header pressure.
If the water supply system pressure falls to 130 psi, the electric-motor-driven fire pump would start automatically.
If the pressure falls to 110 psi, the diesel-engine-driven pump would start.
The diesel pump will also start by loss of power to its controller.
The pumps can be stopped only at the pump controller panels located in the immediate area.
Separate alarms are provided in the control room to monitor pump operation, prime mover availability, and failure of a fire pump to start.
A loss of offsite power will result in only the diesel fire pump being operational.
A closed fuel valve will stop the diesel engine once the fuel in the line is consumed.
At the request of the staff, the applicant by letter dated June 18, 1981, agreed to lock open the supply valve from the elevated diesel fire pump fuel oil storage tank located outside the service water building.
The largest single fire suppression system water demand for areas that need to be protected is 1500 gpm.
Adding 500 gpm for hose streams creates a total water demand of 2000 gpm.
Either of the two tire pumps can deliver the required water flow.
A total of 11 yard hydrants are provided at intervals not exceeding 300 feet.
A fire hose is provided for eacn hydrant.
Sufficient hose will be provided to cover all areas between hydrants with adequate capacity and pressure.
All valves in the fire protection water supply system are locked open and are under administrative controls.
All valves in the fire protection system will be periodically checked to verify position.
The water supply valves meet the requirements of Appendix A, Section C.3.b and are, therefore, acceptable.
The staff finds that the water supply system can deliver the required water demand with one pump out of service.
Based on its review and the applicant's commitment to lock open the fuel supply valve for the diesel fire pump, the staff concludes that the water supply system is adequate, meets the guidelines of Section C.2 of Appendix A, and is, therefore, acceptable.
Fermi SSER2 E-2
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B.
Sprinkler and Standpipe Systems The wet pipe sprinkler systems and deluge systems are designed to the requirements of NFPA Standard 13, " Standard for Installation of Sprinkler Systems," and NFPA 15, " Standard for Water Spray Fixed Systems." The manual hose stations are connected separately to the underground water supply loop or building loop header.
Appropriate sectional and diversion valves are provided so that primary and secondary fire pretection water supplies will always be j
available should a single break develop.
(See also residual heat removal building.) The water supply valves to the sprinklers are locked open and are under administrative controls.
In addition, the sprinkler and standpipe systems have water flow alarms which alarm in the control room.
The areas that have been equipped with sprinkler or spray systems or will be equipped, as stated in the applicant's letter dated June 18, 1981, as the result of the fire hazards analysis include the following:
l Reactor Building Torus Room; Zone 1, Elevation 560' Basement NE Corner Room, Zone 2, Elevation 540' HPCI Turbine and Pump Room, Zone' 3, Elevation 540' Corridor Area, Zone 4, Elevation 562' 1
First Floor, Zone 5, Elevation 583' (railroad bay)
Second Floor, Zone 6, Elevation 613' (cable trays)
Zone 7, Elevation 641'-6" Ventilation system, Zone 9, Elevation 684'-6" j
Emergency lighting, Zone 11, Elevation 643'-6" Auxiliary Building I
Basement Zone 1, Elevation 551' and 562' Mezzanine and Cable Tray Area, Zone 2, Elevation 583-603' Ventilation Equipment Area, Zone 15, Elevation 677' (manual water spray)
Cable spreading room, Zone 7, Elevation 630'-6" (manual open head i
spray system)
Residual Heat Removal Complex Fuel Oil Storage Tank Room Radwaste Building Baled Waste Storage Area Voltage Regulator (automatic deluge)
Turbine Building Reactor Feed Pump Turbine Turbina Oil Reservoir Main Lube Oil Reservoir
-Oil Storage and Turbine Oil Tank Rooms Fermi SSER2 E-3 i
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t-Second Floor Pipe Space Hydrogen Seal Oil Unit (automatic deluge)
Outside Areas North Main Transformer (automatic deluge)
South Main Transformer (automatic deluge)
North System Service Transformer (automatic deluge) t South System Service Transformer (automdtic deluge)
General Service Water Pumphouse Diesel Fire Pump Room Manual hose stations are located throughout the plant to ensure that an effective hose stream can be directed to any safety-related area in the plant.
The standpipe systems are consistent with the requirements of NFPA la,
" Standpipe and Hose Systems for Sizing, Spacing, and Pipe Support Requirements."
The staff concludes that the sprinkler and standpipe systems meet the guidelines of Appendix A to BTP ASB 9.5-1 and are therefore, acceptable.
C.
Gas Fire Suppression Systems The areas that have.been equipped with a low pressure carbon dioxide system include the following:
RHR Building Emergency Diesel Generators SGTS The areas that will be equipped with an automatic total flooding Halon system include the following:
Auxiliary Building Room Outside Division II Switchgear Room, Elevation 641' Cable Spreading Room Computer Room Under Computer Room Floor Relay Room Cable Tunnel Cable Tray Area Zone 5, Elevation 613'-6" Zone 8, Elevation 631' Miscellaneous Room, Zone 11, Elevation 643'-6" Automatic CO systems are activated by heat and/or smoke detectors.
Detection 2
devices activate alarms to indicate the presence of a fire and activate control equipment to initiate discharge of fire extinguishing agents'.
A time delay of sufficient time to enable personnel to leave the area is provided for each system. Activation of the system may also be accomplished manually at local points.
Fermi SSER2
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The residual heat removal complex has its own low pressure C0 2 system con-sisting of two 6-ton storage tanks, one for each division of the residual heat removal complex.
There is a separate discharge system for each diesel generator room.
1 A Halon 1301 total flooding system is used as the primary extinguishing agent in the nonsafety-related computer room underfloor spaces.
Products of 1
combustion detection activate automatic discharge of Halon to the space.
The i
system is activated by ionization and photoelectric detectors on a Class A fire j
alarm circuit.
At the request of the staff, the applicant, by letter dated June 18, 1981, agreed to provide an automatic total flooding Halon 1301 system l
for the cable spreading room in the auxiliary building because of the personnel l
safety factors associated with a total flooding CO2 system.
4' The staff is concerned that a fire in the computer room could spread into the control room.
At the request of the staff, the applicant, by letter dated i
l June 16, 1981, agreed to provide a total flooding Halon 1301 automatic fire I
suppression system for the computer room located within the control room complex.
The staff has reviewed the design criteria and bases for the CO2 and Halon fire suppression systems and concludes that these systems satisfy the provisions of Appendix A to BTP ASB 9.5-1 and are in accordance with the applicable portions
{
of NFPA Standards 12 and 12A.
They are, therefore, acceptable.
1 D.
Fire Detection Systems The fire detection systems consist of the detectors, associated electrical l
power supplies, and the annunciation panels.
The types of detectors used are i
ionization (products of combustion), thermal, photoelectric and heat sensing cable.
Fire detection systems give an audible and visual alarm which annunciates in the plant control room.
Local audible and/or visual alarms are 1
also provided.
The fire detection systems will be installed in all areas i
having safety-related equipment and/or safety related cables.
These include the control room area, and new and spent fuel pool storage areas, and areas of cable concentration.
The fire detection systems are installed according ts NFPA 72D, " Standard for
- he Installation, Maintenance, and Use of Proprietary Protection Signalling Systems." Those fire detection systems that are used to actuate suppression 4
systems are a Class A system defined in NFPA 720.
All other redundant j
safety-related division areas have a cross-zoned Class B' system.
The two fire detector groups are powered from separate non-Class.1E motor control centers.
l Each motor control center is fed from opposite divisional _ Class lE switchgear, j
Normal offsite power provides the primary supply for the detectors.
Upon loss of offsite power, the detectors are automatically connected to the onsite
~
All fire' alarm circuits and alarm bell circuits are electrically supervised.
~
!l' The staf f was concerned that smoke detection is not provided in the auxiliary building, elevation 615' northeast corner in the stairway adjacent to the relay i
Fermi SSER2 E-5
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room.
At the request of the staff, the applicant, by letter dated June 18, 1981, agreed to install additional smoke detectors.
- The staff has reviewed the fire detection systems to ensure that fire detectors are adequate to provide detection and alarm of fires that could occur.
These systems are intalled with due consideration for the use of detector spacings less than those recommended for smooth, unobstructed ceilings.
The staff has
~
also reviewed the fire detection systems design criteria to ensure that they conform to the applicable sections of NFPA 720.
The staff concludes that the design and the installation of the fire detection systems meet the guidelines of Appendix A to BTP ASB 9.5-1, and are, therefore, acceptable.
l III. OTHER ITEMS RELATED TO FIRE PROTECTION PROGRAMS t
A.
Fire Barrier and Fire Barrier Penetrations r
l Walls that separate safety-related buildings are 3-hour-fire-rated walls.
The floor / ceiling assemblies separating areas in buildings containing safe shutdown systems - are 3-hour-fire-rated barriers.
For fire areas not having a 3-hour-fire-rated assembly, the staff analyzed each individually with respect to its fuel load, fire suppression and detection systems, and proximity to safe shutdown equipment and concluded that the fire-rated assemblies provided were adequate for the areas affected, meets Section D.1.d and D.1.j of Appendix A to l
BTP ASB 9.5-1, and, therefore, is acceptable.
l The existing cable tray and pipe penetrations are not now sealed.
The applicant has agreed to provide 3-hour seals that meet specific UL designs.
All fire penetration sells used in the penetration cable trays, conduits, and piping will pass the penetration qualification tests including the time l
temperature exposure curve specified by ASTE E-119, " Fire Test of Building l
Construction and Materials." The staff has concluded that fire seals meet the j
guidelines of Appendix A to BTP ASB 9.5-1, and, therefore, are acceptable.
By letter da'.ed June 18,.1981, the applicant has agreed to provide 1-hour fire-rated barriers for conduit and/or cable trays of one of two redundant divisions that are within 20-feet of each other.
In some areas'1-hour barriers are provided for both divisions, or a 3-hour barrier is provided for one division.
In addition, the applicant, by letter dated June 18, 1981, agreed i
that the cable tray supports will be protected to achieve the same fire rating as the cable tray or conduit itself.
The fire rated barriers will be provided in the following areas:
Auxiliary Building l
L Basement, Zone 1, elevation 551' and 562'.
(Provide a 1-hour-fire-rated wall or fire-rated barrier for Division II~.)
Mezzanine and Cable Tray Area, Zone 2, elevation 603'-6" cable trays and conduit.
Relay Room, Zone 3, elevation 613'.
Cable Tunnel, Zone 5, elevation 613'5".
( A 1-hour-fire-rated barrier
. will be provided on both redundant divisions.)
t Fermi SSER2-E-6 1
~.
v 4
l Stairwell, northeast corner of relay room, Zone 3, elevation 613'6".
l (Provide a 3-hour-fire-rated barrier on one division.)
Cable Spreading Room, Zone 7, elevation 630'-6" (A 1-hour fire-rated barrier will be provided on both redundant divisions of cable trays and conduit.)
Cable Tray Area, Zone 8, elevation 631' (Provide a 1-hour-fire-rated barrier around Division I cable irays and conduit in NE corner of room [H-11]. )
Miscellaneous Room, Zone 11, elevation 643'-6" (G-11 and 11).
In addition to the cable tray, a 1-hour-fire-rated partition should be provided along the 11 line to approximately 4' west of the Division I motor control center and then 6' north of the cabinet.
Reroute Division I conduit to maintain a 20' separation.
Control Center Ventilation Equipment Room, Zone 14, elevation 677'.
(Provide a 3-hour fire rated barrier for one division.) By letter dated June 18, 1981, the applicant agreed to provide 3-hour fira rated barrier for Division I circuits in the Division II area Ventilation Equipment Area, Zone 13, elevation 659'-6".
(Provide a 1-hour-fire rated barrier around Division I trays.
Reroute Division II conduit to maintain 20' separation.)
Zone, 5, elevation 613'-6".
Provide a 1-hour-fire-rated barrier for Division I conduit.
Reactor Building Second Floor, Zone 6, elevation 613'-6" southeast corner (E-F and 10-11).
Ventilateun System, Zone 9, elevation 684'-6".
Third Floor, Zone 7, elevation 641'-6".
First Floor, Zone 5, elevation 503'-6".
High Pressure Coolant Injection Pump and Turbine and Control Rod Drive Pump Room, Zone 3, elevation 540'.
(Provide 1-hour barrier on Division I conduit.)
8.
Fire Doors and Dampers The staff has reviewed the placement of fire doors and verified that all doorway openings to areas containing safe shutdown equipment or circuits are provided with fire doors with ratings commensurate with the fire ratings of the wall.
Fermi-SSER2 E-7
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The applicant has provided 3-hour-fire doors and dampers wherever ventilation ducts or openings penetrate 3-hour-tire-rated walls or ceiling / floor assemblies.
Dampers are not now installed t.o the manufacturer's specifications.
The staff is concerned that a fire in the immediate area of these fire dampers will collapse the ventilation duct and consequently pull the fire damper out of the wall. At the request of the staff, the applicant, by letter dated June 18, 1981, agreed to reinstall all fire dampers according to the manufacturer's instructions.
Based on its review and the commitments, the staff concludes that the fire doors and dampers will be provided in accordance-with the guidelines of Appendix A to BTP ASB 9.5-1, Section D.l.j, and are, therefore, acceptable.
IV.
EMERGENCY LIGHTING The applicant has installed self-coatained 8-hour battery pack emergency lighting in all areas of the plant which could be manned to bring the plant to a safe cold shutdown and in access and egress routes to and from all fire areas.
The staff concludes that the UEergenc Appendix A to BTP ASB 9.5-1, and the'y lighting meets the requirements of provisions of Section III.J of Appendix R to 10 CFR Part-50.
It is, therefbre, acceptable.
V.
FIREPROTECTIONFORSPECIFIC$REAS A.
Control Room Complex l
Summary of Review The control room complex is separated from the turbine and reactor building as well as other areas of the plant by 3-hour-fire-rated walls and floor / ceiling assemblies.
5 Originally the peripheral rooms (including the computer room) located within i
the control room complex did not haie 1-hour-fire-rated walls and doors to separate them from the control room.
Also, no automatic fire suppression system was provided for these rooms.
The staff was concerned that fire in one of these rooms may. spread into the control room.
?
l At the request of the staff, the applicant,;by letter dated June 16, 1981, committed to provide 1-hour-fire-rated walls and doors to separate the peripheral rooms with a potential of a fire hazard from the control room.
The applicant will also provide a total flooding Halon 1301 system for the computer-room.
The area under the raised floor of the computer room is presently protected by a total flooding automatic Halon system.
At the time the SER was prepared, the % [icant h'ad proposed to provide two app divisional remote shutdown panels'and'had performed a fire test to demonstrate that a fire external to the main control panels;would not result in damage to redundant shutdown functions in the control roorq,that would, in turn, cause the
-(
l Fermi SSER2
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3
loss of functions of the redundant functions at the remote shutdown panels.
The staff identified four deficiencies with the test:
(1) The mock-up panels did not simulate the plastic components mounted on the control ro0* panels.
(2) The fire configuration was altered during the test due to the distortion of the fuei pan.
(3) The mock-up panels did not simulate the control room panel ventilation system and this sytem had not yet been designed.
(4) The effects of fire suppressants on the components were not demonstrated By letter dated July 31, 1981, the applicant provided the results of the fire test and provided additional information regarding the four deficiencies.
After reviewing this letter, the staff asked the applicant to demonstrate that (1) the involvement of the plastic components on the control panel in the fire will not cause the loss of integrity of the control panel surface and (2) the ventilation system will not result in a common mode failure of redundant circuits due to a fire in the ventilation system or a pathway for a fire inside one cabinet to the inside of a redundant cabinet.
By letter dated November 24, 1981, the applicant provided additional information in regard to these concerns in a report, " Evaluation of Selected Control Panel Components Subjected to a Postulated Exposure Fire," dated November 1981.
On December 4, 1981, the staff met with the applicant to discuss this issue and to summarize the design features now proposed.
By letter dated January 4,1982, the applicant provided additional information in regard to the combustibility of materials in the control room and the development of emergency shutdown procedures.
Design Features The shutdown circuits in the control room are contained within three pairs of cabinets.
The control cabinets are mounted on a 4-in.-high concrete pad.
Each cabinet contains only cables from one division of shutdown systems.
The redundant division is contained in the adjacent cabinet.
Each set of cabinets is separated from the other sets by several feet.
A Division I remote shutdown panel with controls to achieve cold shutdown conditions is provided in the Division I switchgear room; a Division II remote shutdown panel with controls to achieve hot shutdown conditions will be installed in the Division II switchgear room.
These remote panels are evaluated in a Section VI of this report.
In the control room, redundant components in adjacent cabinets are separated from each other by 3/16-in.-steel panels that are free of p'enetrations.
The rear doors of the cabinets are also 3/16-in.-steel without penetrations.
The components inside the cabinet are mounted toward the front of the cabinet so that about 2 feet of space free of combustibles is at the rear of the cabinet interior.
On the front of the cabinet, the portion of the cabinet below the operating panel is louvered; however, a panel of 1-in.-thick marinite has been Fermi SSER2 E-9
fastened on the inside of the panel to close these openings.
Components on the front of the control panel are exposed to a fire in the control room.
The handles and face plates for the control switches are plastic as are a few indicator faces.
The other faces for annunciator, instruments, and recorders will ba glass.
All control cabling within the control panel has been qualified to IEEE 383 flame test.
There are no power cables in the control cabinets.
The normal control switches on the main control panel and their counterparts on the remote shutdown panel have common circuits.
However, the normal instrumen-tation on the main control panel is independent of the counterpart instrumentation on the remote shutdown panel.
A common ventilation duct supplies air for the redundant control cabinets from above.
A separate duct will be provided in each ct: binet.
A fire damper will be installed where the duct enters the cabinet.
The duct takes tha air to the bottom of the cabinet.
The air exits the cabinet through a vent at the top of the cabinet into the control room atmosphere.
There will be no carpet in the control room.
The interior wall finishes, thermal insulating materials, radiation shielding materials, and sound proofing material are noncombustible; that is, the flame spread, smoke, and fuel contribution are rated at 25, 50, and 50 respectively.
Fire detection has been provided in the ventilation system, the entire control room, above and below the dropped ceiling, and in the main control cabinets and consoles.
Standpipe hose stations and portable extinguishers are provided for manual fire suppression.
A water-type 2 gal portable fire extinguisher is also provided.
Summary of Proposed Modifications The applicant has proposed the following modifications:
(1) A Division II remote shutdown panel will be installed in the Division II switchgear room.
(2) Each of the six control cabinets containing the shutdown systems in the control room will have a 1-in-thick marinite panel installed to close the louvered portion of the front panel.
(3) The plastic faces of the annunciator panels will be replaced with glass.
(4) A separate inlet ventilation duct will be provided for each of the air control panels and a fire damper will be installed in the inlet duct of each cabinet.
(5) Emergency shutdown procedures will be developed to safely shut down the plant.
Evaluation The design features deviate from the staff guidelines (Section III.G of Appendix R to 10 CFR Part 50) in that (1) redundant components in the control room are not separated by either 20 ft or a 1-hour fire barrier, (2) a fixed Fermi SSER2 E-10
suppression system is not installed in the control room, and (3) the alternative shutdown system is not independent of the control room.
The staff accepts the lack of a fixed suppression system on the basis that the control room is required to be manned at all times by at least two persons.
In the' event of a fire, manual fire suppression would be effective and prompt.
The operators provide a continuous fire watch in the control room and such a fire watch is an acceptabla alternative to a fixed suppression system.
The staff accepts the lack of separation by 20 f t or by a 1-hour fire barrier around one division becau*e the majority of the in situ fire load is contained within the control cabinets and because the circuits that are common with the remote shutdown panels have a high damage threshold.
A fire within a cabinet which affects one division should not affect the counterpart components of the other division in an adjacent cabinet or at the other division's remote shutdown panel.
In addition, the applicant will develop emergency procedures to safely shut down the plant.
These procedures could be used to accomplish shutdown if the common circuitry were damaged.
As shown on drawing 6A721-2407, there are three sets of two cabinets that contain shutdown circuits.
The most important set contains the controls for the heat removal systems.
These sets of cabinets are separated by approximately 10 ft.
The staff accepts, the 10-ft separation, 3/16-in. steel cabinets, and the IEEE qualified cable as providing adequate protection.
As part of the NRC fire protection research program, the staff has exposed non-IEEE qualified cable within steel conduit to a 5 gal Heptane exposure fire about 1-1/2 ft distant and demonstrated that the non-IEEE qualified cable was not damaged by such an exposure.
The staff has considered the effects of an exposure fire of transient combustibles that involves in situ combustibles on the face of the adjacent control panels.
In addition, the applicant has performed tests and analyses that are useful in estimating the effects of an exposure fire on the circuitry within the panels.
These tests and analyses considered an exposure fire of 1 gal of Heptane in a 2-ft square pan and oven tests which exposed the components of concern to environmental temperatures of 600 F.
The exposure fire was selected after consideration of transient combustibles the staff review teams have occasionally observed in control rooms during staff site surveys; for example, trash carts filled with crumpled paper, drawing racks with vertical sheets of paper, maintenance and cleaning materials, cardboard cartons of waste paper, and temporary cables for test instrumentation.
The staff recognizes that it is impossible to predict what sort of transient combustibles will be involved in a real fire; however, the staff deems the test fire a conservative simulation.
The staff also recognizes that exposure fires which do not ignite in situ combustibles pose little threat; therefore, the staff was concerned that during the applicant's fire test, the plastic components on the face of the control panel were nct replicated and that the fire duration was shortened due to " boil over" of the Heptane and the distortion of the fuel pan.
Therefore, the fire test was not conclusive in demonstrating that such plastics would not ignite.
The staff asked the applicant to demonstrate what would happen if the plastics ignited if their temperature was raised to 600 F.
Instead, the applicant performed an oven test that exposed the plastic to a 600 F temperature for Fermi SSER2 E-11
several minutes.
During this test the plastic reached only 475 F and did not ignite.
However, the applicant performed calculations of a similar exposure fire which predicted that the plastic components would not ignite.
The tests did show that the components of interest within the panel--that is, the d
switches and their attached cabling--would reach temperatures of only 300-400 F
]
under the test conditions and that such temperatures for periods of a few minutes would not cause loss of function.
The applicant states that the exposed plastic on the control panel will ignite at about 500-600 F.
During the exposure fire test the steel panel heated at a rate of about 100 F per minute.
However, no temperatures were measured on the plastic.
In the oven test of the mounted switch, the steel panel heated at about 25 F per minute; however the plastic heated at about 140 F per minute during the first minute and at SC'F per minute for the next 5 minutes.
At about 6 minutes the temperature if the plastic stayed approximately constant at 475 F while the steel panel continued to rise.
In both the oven end the exposure fire tests, the temperature of the body of the switch raised only 10-20 F.
If these heating rates are extrapolated to the fire test conditions, an exposure fire lasting about 3 or more minutes may lead to ignition of the plastic on the face of the control panels.
Thus, if all such plastic had been included in the panel mockup and if the fire duration were not shortened by
" boil over" and pan distortion, or if the quantity of fuel were slightly greater, or if the area of the fuel pan in the test were slightly less, ignition of the plastic may have occurred.
If the plastic did ignite, the temperature inside the panel would not be signi-ficantly increased.
The plastic face plate is part of the switch mounting in the panel; however, even if this burned away the rigidity of the cables connected to the switch would probably not allow it to drop within the cabinet.
If portion of the burning plastic fall into the cabinets, they do not have sufficient heat capacity to damage the cables.
Qualified cables require exposure to a significant volume of flame for several minutes before they are damaged.
Fire tests for a fire within a cabinet were not performed.
The staff assumes such a fire would involve the cables within the cabinet that would cause the loss of functions of the components in the cabinet as well as their counterparts on one of the remote shutdown panels.
The applicant states that the power distribution system does not have counterpart switches on the remote shutdown panels.
The effects of smoke and fire suppressants on the cables and components within the main control room were not demonstrated.
However, smoke should not damage the components of concern, and as long as panel integrity is maintained, redundant components should not be exposed to suppressants in a manner that will cause short circuits.
The staff accepts the lack of independence of the shutdown control panels in the main control room from those in the switchgear rooms because it is not clear that such independence significantly increases the post-fire shutdown capability of the plant.
There are several locations in the plant where a fire would cause the loss of one division of shutdown systems (for example, switchgear room and power transformer).
The other locations are more likely to have a large fire than the control room.
Therefare, providing independence of Fermi SSER2 E-12
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i l
the remote panels would only affect the least likely fire location, and the benefit of such a modification is marginal.
Because of the uncertainitie; concerning (1) whether the exposure fires are bounded by the test fire and (2) whether the fire suppressants could extend damage to the redundant cabinets, the staff asked that the applicant develop emergency procedures to establich safe shutdown conditions in the event the 4
power distribution system was disrupted by a control rc:cm fire and the control functions were lost at the control room and remote shutuown panels.
By letter dated January 4, 1982, the licensee committed to develop and implement such procedures.
The staff will verify that procedures are implemented.
Conclusion Based on the above evaluation, the staff concludes, that the fire protection features of the control room meet the guidelines of Appendix A, Section F.2 to BTP ASB 9.5-1 and, therefore, are acceptable.
B.
Cable Spreading Room The cable spreading room is separated from the balance of the plant of 3-hour-fire-rated walls and floor / ceiling assemblies.
The room has isolation dampers in the supply and exhaust ducts to contain the automatic Halon fire supppresion system.
These ducts also have automatic fire danpers to isolate the room from the remainder of the plant.
The room has an existing automatic total flooding carbon dioxide system.
At the request of the staff, for personnel safety, the applicant has agreed to install an automatic total flooding Halon fire suppression system which will be activated by ionization and photoelectric detectors on a Class A fire alarm circuit.
Manual fire suppression capability is provided by standpipe, with hose stations and portable fire extinguishers.
In addition, installed smoke detectors will initiate an early warning alarm in the control room prior to the Halon system actuation.
The staff was initially concerned that a fire could affect redundant shutdown systems located in the cable spreading room.
The applicant has proposed to wrap both redundant divisions of cable trays and conduit with a 1-hour-fire rated barrier.
Because access by manual hose stations is limited, at the request of the staff, the applicant, by letter dated June 18, 1981, agreed to provide a fixed piping open head manual spray system for the entire room.
The system will be controlled from outside the cable spreading room.
l The fire protection for the cable spreading room meets the guidelines of Appendix A to BTP ASB 9.5-1 and is, therefore, acceptable.
C.
Containment and Reactor Building The drywell atmosphere of the containment will be inerted with a 97 percent concentration of nitrogen, hence eliminating any potential fire hazard from lubricating oil or hydraulic fluid systems.
Containment and reactor building Fermi SSER2 E-13
fire protection features include hose stations, fire detectors, fire extin-guisheis, automatic sprinklers, and fire control barriers.
Ionization smoke detectors are distributed throughout the drywell with alarm and annunciation in the control room.
Because the containment is inerted, provisions of III.0, of Appendix R to 10 CFR Part 50 are met.
The staff has reviewed the applicant's fire protection for the artas inside the containment and the reactor building and concludes that the fire protection meets the guidelines of Appendix A to BTP ASB 9.5-1.
It is, therefore, acceptable.
D.
Emergency Diesel Generator Rooms The residual heat removal complex located in a separate detached building contains the emergency diesel generators, diesel oil storage tanks, and residual heat removal service water pumps, as well as other safety-reltaed equipment and cables.
Train I is separated from Train II by a blank 3-hour fire rated concrete wall.
The deisel fuel oil storage tanks are separated from the other areas of 3-hour fire rated walls and protected by an automatic sprinkler system.
Smoke detectors with alarms and annunciation in the control room are provided for the Division I and II pump rooms and in the diesel generator switchgear room.
At present a single feed from the underground fire main provides the primary and secondary fire protection for the residual heat removal complex.
A break or a closed valve in this line would elminiate all automatic and manual fire protection in the building.
At the request of the staff, the applicant, by letter dated June 18, 1981, agreed to provide a second feed from the outside underground into the buildi'g to a common header and properly valved to the extent that one of the two suppression systems will always be available.
Based on its evaluation and the commitments, the staff concludes that the fire protection for the diesel generator rooms meets the guidelines of Appendix A to BTP ASB 9.5.1 and is, therefore, acceptable.
E.
Other Plant Areas The applicant's Fire Hazards Analysis addressed other plant areas not specifically discussed in this report.
The applicant has comc. tted to install additional detectors, portable extinguishes, and automatic sprinklers prior to fuel load.
The staff finds that the fire protection for these areas with the commitment made by the applicant is in accordance with the guidelines of Appendix A to BTP ASB 9.5-1 and is, therefore, acceptable.
VI.
FIRE PROTECTION OF SAFE SHUTOOWN CAPABILITY At the request of the staff, the applicant evaluated the fire protection for equipment, cables, and conduits required for safe shutdown that are within 20 f t of each other.
The results of this evaluation were provided in the Fermi SSER2 E-14
applicant's letter dated June 29, 1981.
The applicanthas proposed to instal prior to fuel loading 1-hour fire barriers and fixed suppression systems as described in Sections II.8, II.C, and III.A of this report.
The staff requested similar protection be provided for redundant circuits in the following areas:
Reactor Building Fire Floor, Zone 5, elevation 583'-6", west side outside containment.
Reactor Building Second Floor, Zone 6, elevation 613'-6" general area of reactor vessel level and pressure instrument rack, Division II.
By letter dated June 18, 1981, the applicant stated that these redundant circuits are for control of valves used to achieve cold shutdown; hence, no fire barriers or fixed fire suppression is provided.
The staff concurs with the applicant.
The staff evaluation for safe shutdown capability included a review of the applicant's commitments in his June 19 and July 31, 1981, letters that relate i
to upgrading fire protection in various areas.
The applicant's fire study used I
computer analyses to locate the proximity of redundant cables in essential l
shutdown and power circuits.
The applicant's response to Q 021.32, dated July 31, 1981, proposed an additional remote shutdown panel using Division II equipment.
It also provided sufficient information to demonstrate that adequate redundancy exists for the two remote shutdown panels to reach hot shutdown using Division I or Division II equipment.
Cold shutdown can be achieved through remote Division I equipment but not by remote Division II equipment.
Reentry to the control room is required for cold shutdown using Division 11 equipment.
For both divisional shutdowns it is presumed that the operator initiates a manual scram before leaving the control room.
The applicant has shown that in the event of a fire in the control room, at least one train of shutdown cooling systems could be relied upon to reach hot shutdown:
safety relief valves, residual heat removal system (containment cooling mode), and reactor core isolation cooling system (Division I) or high pressure injection system (Division II).
For cold shutdown, residual heat removal is aligned in the shutdown cooling mode via the Division I shutdown panel or from the control room for Division II.
The staff also reviewed instrumentation on each of the two remote shutdown panels as well as instrumentation not on the shutdown panels, and concludes that sufficient information is available to achieve hot and cold shutdown.
The instrumentation provided includes:
suppression pool temperature, reactor vessel pressure, drywell pressure, reactor vessel level, RCIC flow, RHR flcw, and various valve position indications.
Other process variable that may be useful, but are not normally necessary for safe shutdown may be determined in the relay rooms through the use of portable instruments directly measuring the variables.
These variables include service water flow and pressure, storage tank level, and suppression pool level.
Based on its review, the staff concludes that fire protection of safe shutdown capability, with the approved deviation from Appendix R (remote shutdown panels not electrically independent of the control room) will meet the technical requirements of Sections III.G and III.L of Appendix R to 10 CFR 50, and, therefore, is acceptable.
Fermi SSER2 E-15
VII. ADMINISTRATIVE CONTROLS AND FIRE BRIGADE The administrative controls for fire protection consist of the fire protection organization, the fire brigade training, the controls over combustibles and ignition source, the prefire plans, and procedures for fighting fires, and quality assurance.
The fire brigade will be composed of five members per shift.
To have proper coverage during all phases of operation, members of each shift crew will be trained in fire protection in accordance with NRC guidance, including Regulatory Guide 1.101, " Emergency Planning for Nuclear Power Plants." The applicant has agreed to implement the fire protection program contained in the staff supplemental guidance, " Nuclear Plant Fire Protection Functional Responsibilities, Administrative Controls and Quality Assurance,"
dated August 29, 1977, including (1) fire brigade training, (2) control of combustibles, (3) control of ignition sources, (4) fire fighting procedures, and (5) quality assurance.
The applicant will implement the plant administrative controls and procedures before fuel loading.
The applicant will have a five-member fire brigade which meets the staff guidelines, and is, therefore, acceptable.
The staff concluJes, that, with the commitments, the size of the fire brigade, the necessary equipment, and the adequacy of the training, training will conform to the recommendations of the National Fire Protection Association, to Appendix A to BTP ASB 9.5-1, and to the supplemental staff guidelines.
They are, therefore, acceptable.
VIII. TECHNICAL SPECIFICATIONS The applicant has committed to follow the Standard Technical Specifications.
The staff finds this acceptable.
IX.
APPENDIX R STATEMENT On October 27, 1980, the Commission approved for publication in the Federal Register a new rule, 50.48, and Appendix R to 10 CFR Part 50, delineating certain fire protection provisions for nuclear power plants.
The staff used the technical requirements of this rule in the evaluation of the Fermi 2 fire protection program.
By letter dated June 9, 1981, the applicant nade the commitment to meet the technical requirements of Appendix R to 10 CFR 50, or to provide equivalent protection.
The applicant his proposed and the staff has found acceptable the following deviations:
(1) manual suppression in lieu of fixed fire suppression in the control room (Section V.A of this report)
(2) divisional separation in_the individual control cabinets in combination with redundant remote shutdown capability and emergency shutdown procedures in lieu of alternate shutdown capability independent of the control room (Section V.A. of this report)
Fermi SSER2 E-16
l Based on its evaluation, these commitments, and these approved deviations, the staf f concludes that the fire protection program for the control room meets the technical requirements of Appendix R to 10 CFR 50.
X.
CONCLUSION Based on its review, the staff concludes that when the committed modifications have been completed, the fire protection program for Fermi Unit 2 will meet the i
technical requirements of Appendix R to 10 CFR Part 50 with approved deviations, the guidelines of Appendix A to BTP ASB 9.5-1, and the requirements of GDC 3, and is, therefore, acceptable.
The staff will condition the operating license to require completion of modifications by a specified date.
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APPENDIX F ENVIRONMENTAL QUALIFICATION SAFETY EVALUATION REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION EQUIPMENT QUALIFICATION BRANCH DETROIT EDISON COMPANY ENRICO FERMI ATOMIC POWER PLANT UNIT 2 DOCKET NO. 50-341 Fermi SSER2 F-1
CONTENTS Page 1
Introduction...................................
F-5 2
Background..................................................
F-5 2.1 Purpose..............................................
F-5 2.2 Scope.................................................
F-6 3
Staff Evaluation............................................
F-6 3.1 Completeness of Safety-Related Equipment...............
F-6 3.2 Service Conditiors...................................
F-7 3.3 Temperature, Pressure, and Humidity Conditions Inside Containment.........
F-7 3.4 Temperature, Pressure, and Humidity Conditions Outside Containment....
F-8 3.5 Submergence...........................................
F-8 3.6 Chemical Spray.........................................
F-8 3.7 Aging..................................................
F-8 3.8 Radiation (Inside and Outside Containment).............
F-9 3.9 Outstanding Equipment..................................
F-9 4
Qualification of Equipment............................
F-9 4.1 Equipment Requiring Immediate Corrective Action........
F-10 4.2 Equipment Requiring Additional Information and/or Corrective Action.....................................
F-10 4.3 Equipment Considered Acceptable or Conditionally Acceptable.............................................
F-11 5
Conclusions.................................................
F-11 APPENDIX A Equipment Requiring Immediate Corrective Action....
F-13 APPENDIX B Equipment Requiring Additional Information and/or Corrective Action..................................
F-14 APPENDIX C Equipment Considered Acceptable or Conditionally Acceptable.........................................
F-21 APPENDIX D Safety-Related Systems.............................
F-22 Fermi SSER2 F-3.
l l
SAFETY EVALUATION REPORT BY THE 0FFICE OF NUCLEAR REACTOR REGULATION EQUIPMENT QUALIFICATION BRANCH FOR DETROIT EDISON COMPANY ENRICO FERMI ATOMIC POWER PLANT UNIT 2 DOCKET NO. 50-341 ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT 1
INTRODUCTION General Design Criteria 1 and 4 specify that safety-related electrical equipment in nuclear facilities must be capable of performing its safety-related function i
under environmental conditions associated with all normal, abnormal, and accident plant operation.
In order to ensure compliance with the criteria, j
the NRC staf f required all near-term Operating License (OL) applicants to reassess and evaluate their environmental qualification documentation for-their safety-related electrical equipment.
2 BACKGROUND By letters dated February 5 and 21, 1980, the NRC Office of Nuclear Reactor Regulation (NRR) requested Operating License applicants to review and evaluate the environmental qualification documentat' ion for each item of safety-related electrical equipment and to identify the degree to which their qualification program complies with the staff's positions as described _in NUREG-0588, " Interim Staff Position on Environmental Qualification of Safety-Related Electrical j
Equipment." The applicants were directed to provide a submittal reporting the l
results of this review.
s Subsequently, Commission Memorandum and Order CLI-80-21 (issued on May 23, 1980) states that NUREG-0588 forms the requirements that applicants must meet regarding environmental qualification of safety-related electrical equipment in order to satisfy those aspects of 10 CFR 50, Appendix A, General Design Criterion (GDC) 4.. This order requires that licensees have qualified safety-related l
equipment installed in their plants by June 30, 1982.
IE Bulletin 79-01B " Environmental Qualifica+. ion of Class 1E Equipment," issued January 14, 1980, and its supplements dated February 29, September 30, and October 24, 1980 established environmental qualification requirements for operating _ reactors.
This bulletin and its supplements were provided to OL applicants for consideration in their review.
j In response to the staff request, the applicant provided equipment qualifica-tion information by letters dated June _26 and December 1, 1981.
.2.1 Purpose 1
The' purpose of this SER is to identify equipment whose qualification.documenta-tion does not provide sufficient assurance that the equipment is capable of i
Fermi SSER2 F-5 L
i performing the design function in hostile environments.
The staff position relating to any identified deficiencies is provided in this report.
l 2.2 Scope The scope of this report is limited to an evaluation of the equipment which must function in order to mitigate the consequences of a loss-of-coolant accident (LOCA) or a high energy-line-break (HELB) accident, inside or outside containment, while subjected to the hostile environments associated with these accidents.
3 STAFF EVALUATION The staff evaluation of the applicant's response included an onsite inspection of selected Class 1E equipment, audits of qualification documentation, and an j
examination of the applicant's report for completeness and acceptability.
The criteria described in NUREG-0588 Category II form the basis for the staff i
evaluation of the adequacy of the applicant's qualification program.
j The NRC Office of Inspection and Enforcement (IE) will perform an onsite verification inspection of selected safety-related electrical equipment.
The inspection will verify proper installation of equipment, overall interface integrity, and manufacturer's nameplate data.
The manufacturer's name and model number from the nameplate data will be compared to information given in the Component Evaluation Work Sheets (CES) of the applicant's submittal.
The staff performed an audit of the applicant's documentation for qualification of safety-related electrical equipment on July 13-17, 1981 and December 14-18, i
1981.
The first audit identified several deficiencies.
The second audit in December determined that no significant program deficiencies existed and that the applicant had corrected the deficiencies identified in the first audit.
3.1 Completeness of Safety-Related Equipment In accordance with NUREG-0588, the applicant was directed to (1) establish a list of systems and equipment that are required to mitigate a LOCA and an HELB and (2) identify components needed to perform the function of safety related display information, post accident sampling and monitoring, and radiation monitoring.
The staff developed a generic master list based upon a review of plant safety analyses and emergency procedures.
The instrumentation selected includes parameters to monitor overall plant performance as well as to monitor the performance of the systems on the' list. -The systems list was established on l
the basis of the functions that must be performed for accident mitigation (without regard to location of equipment relative to hostile environments).
The list of safety related systems provided by the applicant was reviewed against the staff-developed master list.
Based upon information in the applicant's submittal, the staff has verified and determined that the systems included in the applicant's submittal are those reqJired to achieve or support:
(1) emergency reactor shutdown, 4
(2) containment isolation, (3) reactor core cooling, (4) containment heat Fermi SSER2 F-6 8
s
,_4.
removal, (5) core residual heat removal, and (6) prevention of significant release of radioactive material to the environment.
The staff therefore concludes that the systems identified by the applicant (listed in Appendix D) are acceptable.
The list of display instrumentation, post-accident sampling and monitoring, and radiation monitoring equipment provided by the applicant will be reviewed against the final emergency procedures.
The applicant identified 150 types of equipment which were assessed by the staff.
These 150 equipment types represent approximately 1500 individual equipment components.
Of the 150 types of equipment, 5 are conditionally qualified (two of these only in selected areas of the plant), and the balance will be either retested or replaced in accordance with schedules provided by the applicant.
The staff has reviewed the deficiencies identified in unquali-fied equipment and agrees with the corrective action plans proposed.
Qualifi-cation of replacement components not yet identified will be reviewed prior to fuel load.
3.2 Service Conditions Commission Memorandum and Order CLI-80-21 requires that the "For Comment" NUREG-0588 is to be used as the criteria for establishing the adequacy of the safety-related electrical equipment environmental qualification program.
This document provides the option of establishing a bounding pressure and temperature condition based on plant specific analysis identified in the applicant's Final Safety Analysis Report (FSAR) or based on generic profiles using the methods identified in these documents.
On this basis, the staff has assumed, unless otherwise noted, that the analysis for developing the environmental envelopes for Enrico Fermi Unit 2 relative to the temperature, pressure, and the containment spray has been performed in accordance with the requirements stated above.
The staff has reviewed the l
qualification documentation to ensure that the qualification specifications envelop the conditions established by the applicant.
Equipment submergence has also been addressed where the possibility exists that flooding of equipment may result from HELBs.
3.3 Temperature, Pressure, and Humidity Conditions Inside Containment The applicant has submitted the LOCA/MSLB profiles used for equipment qualifica-tion purposes.
The peak pressure and temperature resulting from these profiles are as follows:
Max Temp ( F)
Max Press (psig)
Humidity (%)
LOCA 300 56 100 MSLB 340 56 100 The staff has reviewed these profiles and finds them acceptable for use in equipment qualification.
That is, there is reasonable assurance that the actual temperatures and pressures for the postulated accidents will not exceed these profiles anywhere within the specified environmental zone (except in the break zone).
Fermi SSER2 F-7
3.4 Temperature, Pressure, and Humidity Conditions Outside Containment The applicant has provided the temperature, pressure, humidity, and applicable environment associated with an HELB outside containment.
The following area outside containment has been addressed:
(1) Reactor Building The temperature values provided by the applicant were below the screening criterion of saturation temperature at the calculated pressure.
The applicant provided detailed assumptions and calculations for these temperatures which were reviewed by the staff and found to be acceptable.
3.5 Submergence The maximum submergence levels inside containment have been established and assessed by the applicant.
Unless otherwise noted, the staff assumed for this review that the methodology employed by the applicant is in accordance with the appropriate criteria as established by Commission Memorandum and Order CLI-80-21.
The applicant's value for maximum submergence is 576' 6".
Equipment below this level was identified by the applicant.
The applicant indicated that some equipment with the potential for submergence will perform its function before becoming submerged, and that failure after submergence will not affect any other safety-related equipment or system or mislead an operator.
Other equipment is or will be qualified for submergence.
The effects of flooding on safety-related equipment outside the drywell are discussed in Appendix C of the FSAR.
In some areas, the safety function of submerged equipment is performed by redundant equipment in another location isolated from the event.
In other areas, operator actions terminate the leak before safety-related equipment becomes submerged.
The staff's findings with respect to the flooding analysis provided by the applicant are discussed in Section 3.6.2 of the Fermi 2 SER.
3.6 Demineralized Spray Demineralized water spray may be utilized for containment heat removal following an accident.
The effect of demineralized spray on the operability of safety-related electral equipment was considered in the applicant's review of documention for qualification of equipment.
3.7 Aging i
NUREG-0588 Category II delineates two aging program requirements.
Valve operators committed to IEEE Standards 382-1972 and motors committed to IEEE Standard 334-1971 must meet the Category I requirements of the NUREG.
This-requires the establishment of a qualified life, with maintenance and replacement schedules based on the findings.
All other equipment must be subject to an l
aging program which identifies age-susceptible materials within the component.
Additionally, the staff requires the applicant to:
i Fermi SSER2.
-F-8 r
--- 1
(1) Establish an ongoing program to review surveillance and maintenance ecords to identify potential age related degradation.
(2) Establish component maintenance and replacement schedules which include considerations of aging characteristics of the installed components.
The applicant has established a limited qualified life for each qualified equipment type through test and/or analysis.
In addition, the applicant has developed a program plan for surveillance and maintenance to ensure that equip-ment will not degrade sooner than predicted.
The staff has reviewed this plan and finds it acceptable.
Surveillance and maintenance program procedures are to be implemented before full power operation.
Until these procedures are implemented, aging will remain an open item.
The applicant is requested to notify the staff when procedures are implemented.
. 3. 8 Radiation (Inside and Outside Containment)
The applicant has provided values for the radiation levels postulated to exist following a LOCA.
The application and methodology employed to determine these values were presented to the applicant as part of the NRC staff criteria con-tained in NUREG-0588 and in the guidance provided in IEB-79-01B, Supplement 2.
The staff review determined that the values to which equipment was qualified enveloped the requirements identified by the applicant.
The value required by the applicant inside the drywell is an integrated dose of 2. 0 x 109 rads.
The applicant has provided the analysis including basis, assumptions, and a sample calculation used to determine the radiation levels.
The staff has reviewed the analyses and concludes that the applicant's overall approach should result in reasonable estimates of the radiation qualification values and is acceptable.
A required value range of 3.25 x 102 to 7.2 x 106 rads has been used by the applicant to specify limiting radiation levels equipment outside containment.
This value considers the radiation levels influenced by the source term methodology associated with post-LOCA recirculation fluid lines and is acceptable.
- 3. 9 Outstanding Equipment Modifications as a result of the TMI Action Plan will be qualified to meet NUREG-0588 Category I requirements at the time of installation or when the.
equipment is required to be operable, per the established schedule of NUREG-0737 (" Clarification of TMI Action Plan"), unless directed otherwise by the staff.
4 QUALIFICATION OF EQUIPMENT The following subsections present the staff's assessment, based on the applicant's submittal, of the qualification status of safety related electrical equipment.
The statf has separated the safety related equipment into three categories:
(1) equipment requiring replacement based on this review, (2) equipment requiring additional qualification information and/or corrective action, and Fermi SSER2 F-9
(3) equipment considered acceptable if the staff's concern identified in Section 3.7 is satisfactorily resolved.
An appendix for each subsection of this report provides a list of equipment for which additional-information and/or corrective action is required.
Where appropriate, a reference is provided in the appendices to identify deficiencies.
It should be noted, as in the Commission Memorandum and Order, that the deficiencies identified do not necessarily mean that equipment is unqualified.
However, they are cause for concern and may require further case-by-case l
evaluation.
4.1 Equipment Requiring Replacement Prior to Startup Appendix A identifies equipment (if any) which the staff review has determined requires replacement prior to plant startup.
There is no equipment in this category for Fermi 2.
Equipment items which have been scheduled for replacement as a result of the applicant's review are presented in Section 4.2.
4.2 Equipment Requiring Additional Information and/or Corrective Action Appendix B identifies equipment in this category, including a tabulation of deficiencies.
The deficiencies are noted by a letter relating to the legend (identified below), indicating that the information provided is not sufficient for the qualification parameter or condition.
Legend R
- radiation T
- temperature QT qualification time RT - required time P
pressure H
- humidity CS - chemical spray A
- material-aging evaluation; replacement schedule; ongoing equipment su veillance S
- submergence M
- margin I
- HELB evaluation outside containment not completed QM qualification method RPN - equipment relocation or replacement; adequate schedule not provided EXN - e.<empted equipment justification inadequate SEN - separate effects qualification justification inadequate t
QI qualification information being developed RPS - equipment relocation or replacement schedule provided l
l As noted in Section 4, these deficiencies do not necessarily mean that the equipment is unqualified.
However, the deficiencies are cause for concern and require further case-by-case evaluation.
The staff has determined that an acceptable basis to exempt equipment from qualification, in whole or part, can be established provided the following can be established and verified by the applicant:
l l
Fermi SSER2 F-10
- 1 V
(1) Equipment does not pertorm essential safety functions in the harsh environ-ment, and equipment failure in the harsh environment will not impact safety related functions or mislead an operator.
(2a) Equipment performs its function before its exposure to the harsh environment, and the' adequacy of the time margin provided is adequately justified, and (2b) Subsequent failure of the equipment as a result of the harsh environment does not degrade other safety functions or mislead the operator.
(3) The safety-related function can be accomplished by some other designated equipment that has been adequately qualified and satisfies the single-failure criterion.
(4) Equipment will not. Le subjectso to a harsh environment as a result of the postulated accident.
4.3 Equipment Considered Acceptable or Conditionally Acceptable Based on the staff review of the applicant's submittal, the staff identified the equipment in Appendix C as (1) acceptable on the basis that the qualification program adequately enveloped the specific environmental plant parameters, or (2) conditionally acceptable subject to the satisfactory resolution of the staff concern identified in Section 3.7.
For the equipment identified as conditionally acceptable, the staff determined that (1) The applicant has not completed the evaluation of plant equipment material to ensure that no known materials susceptible to degradation because of aging have been used.
t (2) Although the applicant has established a plant surveillance and maintenance program, implementation procedures have not been completed.
The applicant is, therefore, required to inform the staff of both completion of the aging program and implementation of the surveillance and maintenance program.
5 CONCLUSIONS The staff has determined that the applicant's listing of safety-related systems and associated electrical equipment whose ability to function in a harsh environ-ment following a postulated accident is complete and acceptable except as-noted in Section 3 of this report.
The staff has also determined that the environmental service conditions to be met by the electrical equipment in the harsh accident environment are. appropriate, except as noted in Section 3 of this report.
The staff has reviewed the qualification of safety-related electrical equipment to the extent defined by this SER and finds the qualification status or correc-tive action plan for achieving qualification acceptable.
The majority of equipment will have additional testing or be replaced, as indicated in Fermi SSER2.
F-11
._. ~.
l Section 4.2.
Equipment whose qualification has been demonstrated through the test or replacement program shall have central files with qualification docu-mentation established two months prior to fuel load.
The NRC will conduct an audit of this new information prior to the fuel load date.
Any remaining out-standing items from Section 4.2 must have justifications for interim operation submitted to the staff for review two months prior to fuel load.
In addition, updated component evaluation worksheets for newly qualified equipment must be provided by the same date.
Based on these considerations, the staff concludes that conformance with the above requirements and satisfactory completion of the corrective actions by the Commission deadline will ensure compliance with the Memorandum and Order of May 23,1980.
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Fermi SSER2 F-12
APPENDIX A EQUIPMENT REQUIRING REPLACEMENT PRIOR TO PLANT STARTUP (CATEGORY 4.1)
No equipment in this category Fermi SSER2 F-13
APPENDIX B EQUIPMENT REQUIREING CORRECTIVE ACTI0tl OR ADDITIONAL INFORMATION (Category 4.2)
Equipment Description Manufac urer Model No.
Deficiency Relay Agastat EGPI-001 RPS Relay Agastat GPIC QI Terminal Blocks Allen-Bradley 1492-CA1 RPS Cable Anaconda EP QI Solenoid Valve (I/C)
ASCO NP8320 QI Solenoid Valve ASCO HT8317 RPS Solenoid Valve ASCO THT8262 QI Solenoid Valve ASCO WP8316 RPS Solenoid Valve ASCO HVA-90-405-2A QI Dual Solenoid Valve ASCO 832322 RPS Solenoid Valve ASCO 8316C37 RPS Solenoid Valve ASCO 8320A10 RPS Solenoid Valve ASCO 8320 RPS Solenoid Valve ASCO 8320A90 RPS Solenoid Valve ASCO 8321A3 RPS Delay Timer Auto-Timing 3198006QIC QI
& Control Solenoid AUVAL C4988-15 RPS Manifold Assembly Pressure Barksdale B2T-M12SS RPS Switch Pressure Barksdale D2H-M80SS RPS Switch D2H-M150SS RPS Differential Barton 288 RPS Pressure Switch Differential Barton 289 RPS Pressure / Flow Switch Flow Barton 368 RPS Transmitter Flow / Pressure Barton 384/S321 RPS Transmitter Fermi'SSER2 F-14
i Equipment Description Manufacturer Model No.
Deficiency Cable Boston Bostrad 7E QI Teminal Blocks Buchanan 427,430 RPS
& End Piece Fuse Block &
Buchanan 358/330 RPS End Section Fuse Block 2P Bussman 2919 RPS Fuse Block 1P Bussman 4513 RPS Fuse Bussman FNM RPS Fuse Bussman FRS RPS 1.2KW Space Chromalox S-1202 RPS Heater Electrical Conax 7087 QI Penetration Thermocouple Conax Copper Constantan RPS Terminal Blocks Connectron NU-2 QI l
Control Crittenden B13410 QI Transformer Pressure Switch CUSCO 604V1 RPS Differential CUSCO 67408005 RPS Pressure Switch Hydrogen Cell Delphi B5G-1B6-C QI 0xygen Cell Delphi B6G-186-C QI Air Cool Fan Delphi Armstrong 11AV QI Heat Exchanger Pressure Dwyer 3000-0 RPS Differential Switch Power Supply Elma 5965A RPS Temperature Fenwall 22810 QI Switch Temperature Fenwall 35003-0 RPS Switch Controller Thermo Element Fenwall 35680-4-255 RPS Temperature E/P Fisher 546 QI Converter Fermi SSER2 F-15
T t
Equipment Description Manufacturer Model No.
Deficiency Radiation GE 237X731G001 QI Monitor Detector i
Radiation GE 194X927G011 QI Monitor i
Intermediate GE 112C3144G008 QI Range Detector Power Range GE 163C1154G002 QI Detector Voltage Pre GE 112C2218G001 QI Amp i
Guide Tube GE 136B1302G002 QI Valve l
Assembly Cable GE Vulkene QI Supreme Cable GE Flame'-Resistant QI Vulkene j
. Cable GE EPR/ Neoprene QI Heater, 200W GE 2A9078102 QI Selector GE CR2940VB203F QI Switch Push Button GE CR2940WA202C QI i
Switch Pump GE CR206B1 QI Starter Differential GE 555111 RPS Pressure / Flow
-Transmitter l
Pressure GE 556120 RPS Transmitter Differential Hays-Republic T-00252A-4 RPS Pressure Transmitter 90KW Heater HESCO HET501 RPS i
- Door Interlock Solenoid Hoffman AEK460 QI
- 24KW Air Heater:
INDEEC0 26SSSF54 RPS l
I Fermi SSER2-F-16 i
Equipment Description Manufacturer Model No.
Deficiency Circuit Breaker ITE HE38015 QI Circuit Breaker ITE HE38030 QI Circuit Breaker ITE HE3B060 QI Circu t Breaker ITE EF3-8015 QI i
Circuit Breaker ITE FJ3-B150 QI Circuit Breaker ITE EF3-A010 QI Circuit Breaker ITE EF3-L050 QI Fused Disconnect ITE 5641-DA, DB, DC QI Switch Fused Disconnect ITE D1054 QI Switch Non-Reversing Starter ITE 5641-DACAB QI Non-Reversing Starter ITE 5641-DBDAB QI Non-Reversing Starter ITE 5641-DCEAB QI Contactor ITE A103C12 QI Non-Reversing Starter ITE A203C12 QI Reversing Starter ITE 5641-R-DACAB QI 5641-R-DBDAB QI 5641-R-DCEAB QI 2-Speed Motor ITE 5641-SW-DCEAB QI Starter Thermal Overload ITE G30T QI Control Transformer ITE 2032-T3, T4, T6, T10 QI Relay ITE J10, J20 QI Contactor ITE 5642-DXCAB QI 5642-DXDAB Motorized ITE T01-F120 QI Operator Transient ITE F20C12 QI Suppressor Motor Control ITE 5640, 5600 QI Center Wire ITT-Royal SIS RPS Electric Terminal Kulka MAI-60, GDI-30F, MDG QI Blocks Fermi SSER2 F-17
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i e
Equipment Description Manufacturer Model No.
Deficiency Temperature /
i.eeds &
Electromax III QI i
Flow Controller Northrup E/P Converter Leeds &
10970-3 RPS j
Northrup Flow Transmitter Leeds &
1913 RPS Northrup Valve Operator Limitorque SMB, SBD Q1 (inside containment)
Valve Operator Limitorque SMB, SMC, SBD QI Temperature Switch Love 56-838 QI Terminal Blocks Marathon 6012-DJ-5V RPS 6006-DJ-SV Current Midwest 4CT16 QI Transformer E.ectric Product i
Temperature NECI N145C3224 RPS Element j
Cable Okonite Okonite Okolon QI j -
Thermocouple Omega ICSS-14U-12 RPS Relay Potter KRPI-IAG QI Brumfield Temperature Pyco 22-1021-02-06 RPS Detector Cable Raychem Flamtrol QI EECW Pump Motor Red Band 8A0 RPS Electric Fan Motor Reliance P21G71PFZ RPS-Fan Motor Reliance P28611CG15EZ RPS Fan Motor Reliance 278919A1,'A4, A8, A9 RPS Pump Motor Reliance 721804-Y8 QI Level Switch Rnbertshaw 83843 QI Level Switch Robertshaw 83844-B1 QI Cable (single Rockbestos Firewall III QI conductor)
. Cable (coaxial)
Rockbestos Firewall III QI i
I
-Fermi SSER2 F-18
l i
t p
Equipment Description Manufacturer Model No.
Deficiency Pressure Transmitter Rosemount 1151 QI Pressure Transmitter Rosemount 1152 RPS Trip Unit Rosemount 5100U QI Solenoid Valve Ross 2773A5905 RPS Pump Motor Rotron DR313 RPS Relay Rowan 2190E QI Relay Rowan 2190-E02AA QI Relay Rowan 2190-E20AA QI Relay Rowan 2190-E21AA QI l
l Control Rowan 8025-L7 QI Transformer Fuse Shawmut TRI/0NIC QI 200 Amp Fuse Shawmut K-5 QI Fuse Shawmut A6Y QI Solenoid Valve Skinner L20B5150 RPS Quick Break Fuse Square D QMB QI Disconnect Circuit Breaker Square D Q0130VH RPS Current Relay Square D 9055-B0-115 QI Circuit Breaker Square 0 Q0T1515 RPS Key-Operated Square D 9001-KS-11-B-18 RPS Selector Switch Pressure Switch Static-0-Ring SN-AA3-X95PT
.RPS 6NAA21 Pressure Switch Static-0-Ring 6P3-K3M4C QI Flow Switch Static-0-Ring 15R3-KYIC QI p
Solenoid Valve Target Rock 72V QI Solenoid Valve.
Target Rock 7567F QI Solenoid Valve Target Rock 78U QI Position Switch Target rock Position Switch QI
,e of 72V.
,j.
f l
. Thermocouple Thermo Electric Type K RPS l
Thermocouple Thermo Electric Type T RPS
(
l Autotransformer Time-Trol-1182 QI
~
-Assembly g
l SCR Power Controller Time-Trol 1066ZC3-125EX QI-4 Fermi SSER2 F-19 n-s g..
Equipment Description Manufacturer Model No Deficiency Temperature United 93 QI Switch Electric Level Switch United J300K/455 QI Electric Resistance Weed 601-1A-3-C-RPS Temperature 1-5-0-0 Detector Resistance Weed RTD-550-1669 QI Temperature Detector Terminal Blocks Weidmuller SAK (Phenolic)
QI Relay Westinghouse ARB300VAC QI Motor Starter W/0L Westinghouse A200M2CAC,1CAC RPS Fan Motor Westinghouse 182T (SBDP)
RPS Fan Motor Westinghouse 05-20H4-TBDP-MKB RPS 05-15H4-TBDP-MKB Blower Motor Westinghouse TBFC RPS 20 HP Fan Motor Westinghouse AF73 RPS Control Westinghouse MTA,MTC QI Transformer Power Westinghouse A201K2CA RPS Contactor A201K0CA RPS Circuit Westinghouse FB3010, FB3030 QI Breaker FB3060, FB3070 QI Fermi SSER2 F-20
APPENDIX C EQUIPMENT CONSIDERED ACCEPTABLE OR CONDITIONALLY ACCEPTABLE (CATEGORY 4.3)
Equipment Description Manufacturer Model No.
Deficiency Solenoid Valve (0/C)
ASCO NP8320 A
Pressure Transmitter Rosemount 1153, Series B A
Terminal Blocks Weidmuller SAK (Melamine)
A Pump Motor GE SK A
Cable (0/C)
Okonite Okolon A
s m.,
b
+
- Fermi SSER2 F-21
APPENDIX D 3
Safety-Related Systems Function System 1.
Emergency Reactor Shutdown Reactor Assembly (811)
Switchgear (R14)
Nuclear Boiler (B21)
CRD Hydraulic Control (Cll)
Neutron Monitoring (C51)
Reactor Protection (C71)
Motor Control Center and Distribution Cabinets (R16)
Wire and Cable (R 34)
Standby Emergency Power System (R30)
DC Systems (R32)
Vital Power (R31)
Heating, Ventilation and Air Conditioning (X41) 2.
Containment Isolation Nuclear Boiler (B21)
CRD Hydraulic Control (Cll)
Process Radiation Monitor (011)
Residual Heat Removal (Ell)
Core Spray (E21)
High Pressure Coolant Injection (E41)
Reactor Core Isolation Cooling (E51)
Radwaste (G11)
Reactor Water Cleanup (G33)
Torus Water Management (GS1)
Emergency Equipment Cooling Water (P44)
Motor Control Centers and Distribution Cabinet (R16)
Wire and Cable (R34)
Containment (T23)
Containment Atmosphere (T48)
Primary Containment Monitoring (T50)
Emergency Equipment Service Water (P45)
Standby Emergency Power System (R30)
IThe NRC staff recognizes that there are differences in nonmenclature of systems because of plant vintage and engineering design; consequently some systems performing identical or similar functions may have different names.
In those instances it is necessary to verify the system (s) function with the applicant.
Fermi SSER2 F-22
Function System Switchgear (R14)
DC Systems (R32)
Vital Power (R31)
Heating, Ventilation and Air Conditioning (X41) 3.
Reactor Core Cooling Remote Shutdown (C35)
Nuclear Boiler (821)
Reactor Recirculation (B31)
Residual Heat Removal (Ell)
Core Spray (E21)
High Pressure Coolant Injection (E41)
Reactor Core Isolation Cooling (E51)
Emergency Equipment Cooling Water (P44)
Motor Centrol Centers and Distribu-tion Cabinets (R16)
Wire and Cable (R34)
Heating, Ventilation and Air Conditioning (T41)(X41)
Emergency Equipment Service Water (P45)
Compressed Air (P50)
Standby Emergency Power System (R30)
Switchgear (R14)
DC Systems (R32)
Vital Power (R31) 4.
Containment Heat Removal Residual Heat Removal (Ell)
Emergency Equipment Cooling Water (P44)
Motor Control Centers and Distribu-tion Cabinets (R16)
Wire and Cable (R34)
Heating, Ventilation and Air Conditioning (T41) (X41)
Standby Emergency Power System Containment Atmosphere Cooling (T47)
Remote Shutdown (C35)
Emergency Equipment Service Water (P45)
Switchgear (R14)
DC Systems (R32)
Vital Power (R31)
Fermi SSER2 F-23
Function System 5.
Core Heat Removal Residual Heat Removal (Ell)
Emergency Equipment Cooling Water (P44)
Motor Control Centers and Distribution Cabinets (R16)
Wire and Cable (P34)
Heating, Ventilation and Air Conditioning (T41) (X41)
Remote Shutdown (C35)
Emergency Equipment Service Water (P45)
Standby Emergency Power System (R30)
Switchgear (R14)
DC Systems (R32)
Vital Power (R31) 6.
Prevention of Significant Process Radiation Monitor (D11)
Release of fiadioactive Radwaste (G11)
Materials to the Environment Motor Centrol Centers and Distribution Cabinets (R16)
Wire and Cable (R34)
Standby Gas Treatment (T46)
Switchgear (R14)
Standby Emergency Power System (R30)
DC Systems (R32)
Vital Power (R31)
Heating, Ventilation and Air Conditioning (X41) (T41)
Fermi SSER2 F-24
NRC roRv 335 U.S NUCLE AR REGUL ATORY COMMISSION
- 1. REPORT NUVBE R (Ass,gneo Dy DDC/
I BIBLIOGRAPHIC DATA SHEET Supplement flo. 2 4 TI T LE AN D SUBTS T LE IA dd Volume No, o f appropronte) 2 (Leave blank)
Safety Evaluation Report related to the operation of Enrico Fermi Atomic Power Plant, Unit tio. 2 3 RECIPIENT'S ACCESSION NO.
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- 5. D ATE REPORT COMPLE TED MON TH l YEAR 53;.. of 3qma l..
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9 PE HFORMING ORGANIZATION N AME AND MAILING ADDRESS (loctuar 2,0 Code!
DATE REPORT ISSUED
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'"l"
10 Washington, DC 20555 6 'L ' *
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a! 'i tional infcrration subni tted iar the anplican. rcrarlinc cuts ta'dir." revieu issacs i 'enti fic_ in Lpoler.ent
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tI N E Y WOR DS AND DoctA'E N T AN AL YSIS 17a DE SC RIP TO RS 11t> # DE N TIF IE RS OPE N E N DE D TE RYS 18 AV AIL ABILITY ST ATE ME NT 19 SE gv yjf
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NnCFony 335 47 77,
UNITED STATES NUCLEAR REGULATORY COMMISSION W ASHINGTON. D. C. 20555 OFFICI AL SUSINESS C " "' "
PE N ALTY FOR PRIV ATE USE,5300 u
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