ML20006B099

Licensing Change Request 89-213-UFS to Updated FSAR for Fermi 2

ML20006B099
Person / Time
Site:	Fermi
Issue date:	11/01/1989
From:	DETROIT EDISON CO.
To:
References
NUDOCS 9001310406
Download: ML20006B099 (35)

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B) Technical Expert Date i Z/6*l C) Nuclear Generation Unit Head Adli Date /#-h - N D) General Director, Nuclear Engineering [ INA Date /#

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.c these changes are changes to the sections as

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,.. O~and coordination of refueling.

The Operations Support Engineer v

1 1 l shall have an SRO license.

The General Supervisor - Radwaste is responsible for the supervi-sion, both technical and operational, of plant activities involv-ing radioactive waste processing.

The Radweste Supervisor -

Operations reports to the General Supervisor - Radwaste.

2 The Startup Engineer - Test Phase was functionally responsible for the checkout and initial operation and preoperational testing and the ascension-to-power testing programs.

l 2l The General Supervisor - Chemistry is responsible for directing the sampling of plant fluid systems, for the chemical laboratory, I

for prescribing-the procedures to be followed for sample preparation and analysis and result reports, and for environ-The General Sup mental regulatory compliance.

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evaluates results, reports, and laboratory techniques.

Chemical Engineer, the Chemist, and an engineer report to the 2

I General Supervisor - Chemistry.

N The Chemical Engineer is responsible for the operation,.mainte-nance, and calibration of the plant chemical processing and I

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water treatment equipment.

The Chemist is responsible for maintaining the radiological-and

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1 p chemical parameters of the plant within the requirements of the l (g g Technical specifications and the Radiological Effluent Technical if Specifications.

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sueerintendent - Maintenance and Modifications 13.1.2.3.2.2 j

g The Superintendent - Maintenance and Modifications is respon-l 32 sible for the maintenance of the plant and all associated l

t equipment and the modification to plant structures, systems, l

ReDorting to the Superintendent - Maintenance h ~

and ecuipment.anc Modifications areTthe General Sup/ Instru ervisor

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The Superintendent - Maintenance and Modification is responsible 1

for the maintenance of plant structures, systems, and equipment.

In this capacity, compliance with the Technical Specifications l

related to maintenance, written procedurer', and work practices is ensured; and duties include the lubrication program, the preven-tive maintenance program, instrument spare parts, routine cali-i bration, instrument and control troubleshooting, and the J

j standards calibration program.

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<O The Assistant Superintendent - Maintenance and Modification is responsible for the Measuring and Test Equipment (M&TE) Program, Maintenance Procedures, i

Maintenance Engineering Support, and Maintenance programs such as Preventive Maintenance and Environmental Qualification. Reporting to the Assistant j

Superintendent - Maintenance and Modification are the General Supervisor-Maintenance Support, the Principal Engineer - Maintenance, and the M&TE Supervisor.

The General Supervisor - Electrical Maintenance is responsible for ensuring that electrical maintenance is performed in accordance with procedures. Reporting to the General Supervisor - Electrical Maintenance are the Supervisors - Electrical Maintenance.

The General Supervisor - Mechanical Maintenance is responsible for ensuring that mechanical maintenance is performed in accordance with procedures. Reporting to the General Supervisor - Mechanical Maintenance are the Supervisors - Mechanical 1

Maintenance.

I i O The General Supervisor - Instrument and Control is responsible for repair, _

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maintenance and calibration of the instrument and control devices located in the Plant in accordance with procedures. Reporting to the General Supervisor -

Instrument and Control are the Supervisors - Nuclear Instrument and Control.

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e orders to the work crew a suring that the. work is cone in Responsibilit'ies also s

accordance'with approved oc ures.

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a eral Foreman ~ Nucl a Ins nt and fontrol is the Foreman -

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The Foreman - Nucle natrument- 'n Control is responsible for direct ng the d y to-day,.mork f

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The Supervisor -

in nance Su port a responsible for provid-ing technical se y e and operat ng su t for all phases of i y.N. /

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The ventive ntenance Coordinator is r ponsib'le for th's of the spare part preventive 2

a4.controlheliabilityDataS tem (NPRDS),.and adminis ratio:

ear Plant maintenan du tenance area of grgramsinthema N

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esponsibil t to help e 13.1.2.3.2.3 Director - Nuclear security i

The Director - Nuclear Security is responsible for the physical security of Edison nuclear power plants and the-facilities, l.

material, equipment, and construction associated with them.

The j

physical security responsibility includes developing and imple-menting the Physical Security Plan, the Safeguards Contingency Plan, the Security Personnel Training and Qualifications Plan, i

3-the Safeguards Information Protection Program, and security i

13.1-11 REV 2. 3/89

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Plant Manager - Muclear*

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Superintendent - Operations.

Plant Manager (Section 4.2.1)

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13.1.2.6 oualifications of Plant Personnel l'

The resumes of the initial appointees to the managerial and j

supervising technical positions for Fermi 2 were included in the original FSAR.

The qualification summaries of personnel cur-rently filling these or e uivalent positions are maintained e

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[ ] Operating License [ ] Tech Specs [ ] Environmental Protection Plan E5 E

[ ] Tech Spec Clarification

8) Section(s), Table (s), Figure (s), etc. Affected (Attach marked-up pages)

O C) Reference and Source Documents Attached

[ ] Significant Hazards Consideration

[ ] Environmental Evaluation

[ ] Enviromental impact / Categorical Exclusion

[ ] Other D) is UFSAR change required?

[ ] Yes [ ]No LCR No.

E) Priority STATUS NRC approval required by (date):

,- t An [ ] Emergency [ ] Exigent condition will occur if not approved b) N"p "l2/ > '

(State date):

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//.:Lo-Er9 Explanation CONTROLLED REV.

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F ********* **** ******** * ** * * *.E A R T 3 : A PP R O V A L S* *" ; " " " " "" " " " " " " " " " " "" - - ********

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8) Technical Expert R. J. BeaudryMhb.

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D) General Director, Nuclear Engineering [ ]NA

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E) Plant Manager Date F) Other Date s

G) Director, Nuclear Licensing h

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Revision 0 Page 1 of 5 F; ;;; ;;;;;;PART 1: UFSAR, PLAN, OR Hs0GitAN HETISION [

JEA Nl A) Document EDP-9724, Rev. O B) Section(s), Table (s), Figure (s), etc. Affected ( Attach marked-up pages)

Flaure 10.4-9 sheets 1 and 2., M L n ~4 WI W.C $ - l-C) Reason for (hange EDP-9MF, Rev. O installs 3-way air operated valves in the operating air sup WY to heater drain control valves and impacts UFSAR D) Reference and b r oe Acuments (Identify)

Figures.

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Procedure _N/A ABN i

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)Yes [X ]No Other Drawings, Design Calculations, Correspondence, etc.

1:::::::;;;;;;;;;;;;;PART 2: OPERATING LICENSE NA M [ IJNA ;;;;;;;;;l A) Document

[

] Operating License [

] Tech Specs

[-

) Environmental Protection Plan

(

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C) Reference and Source Documents Attached STATUS

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[

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{

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) Other DISL 'No-M D) Is UFSAR change required?

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[ ] Operating License [ ] Tech Specs [ ] Environmental Protection Plan

] Tech Spec Clarification B) Hection(s), Table (s), Figure (s), etc. Affected (Attach marked-up pages)

C) Reference and Source Documents Attached

[ ] Significant Hazards Consideration

[ ] Environmental Evaluation

[ ] Enviromental Impact / Categorical Exclusion

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STATUS

] Yes [ ]No LCR No.

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1) NSRG Approval (Operating License Amendments) [ ' NA Date l

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pressure-control valves are provided to maintain the gland steam

~ header pressure constant at approximately 2 psig, by either sup-plying or dumping steam as required.

If pressure rises above 5 psig, excess steam is discharged to the condenser by a relief valve.

Temperature and pressure gages are installed in a local panel.

Test flow orifices are provided to monitor operation of the system.

10.4.4 Turbine Bypass syste_m 10.4.4.1 Introduction The Fermi 2 bypass system is a composite of passive and active components that provides a steam path following a turbine-generator trip.

The Fermi bypass system has two key features the live steam supply to the turbine reheaters, which has a nominal flow of 10 percent of nuclear boiler rating, and the electro-hydraulic control (EHC) redundant bypass valves, which are each designed to bypass a nominal 12-1/2 percent of 105 percent reactor flow to the condenser.

Immediately following a turbine or generator trip from rated power, the bypass system will have a nominal capacity of 35 per-cent of nuclear boiler rating.

Following a typical trip, the live steam supply is eventually isolated and the pressure control system maintains the setpoint pressure by modulating the bypass

• valves.

10.4.4.2 Summary The Fermi bypass system is designed in such a manner that the loss of the bypass system would require multiple random failures in the system.f Because this is very unlikely, the turbine trip

( without cypass transient has been analyzed as the turbine trip with a single bypass valve failure and, as such, this event is no l

longer the limiting transient with respect to minimum critical l

f power ratio (MCPR) limits.

i Edison maintains that the design of the Fermi 2 bypass system i

obviates the need to consider the turbine trip without bypass to

{D be part of the design basis.

However, at the request of the NRC, i

Edison performed a sensitivity analysis-of the turbine-generator

4 trip with po'stulated failure of bypass flow to the condenser.

p The analysis was, based on input parameters specified in bg Table 15.0-1 and used the ODYN ca=nntar code.

The results are_

summarized in subsection 15.2.3/fHE Tech 4.HcAt 1steciFic.MTiod l'oA (P)cPR **g BAsto oN West R.esv&.Ts Awa TM peo uTHomavAL aAasA evtVT:

The anarysis et reneater steam zlowf which is M tant in tne

)

l I

turbine trip analysis, is described further in subsection v

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15.E.E.1.E.3 manerator Toad RmJeetien Witbaot auemme

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Go discussed in Subsection 10.4.4. Fermi E has a unieue bwpass swetem.

Ltse of the bwpass swatem would require sultiple random failures.

However. as identified in Table 10.4-1. loss of the SDP 130 UDC-l J

fcoding the swstem, causes both bwpass valves to close. Edison

-cointained that this is verw unlikelv and need not be considered part of the design basis of the Fermi E plant. Therefore, the anslvsis of turbine generator trip' described in this section includes cases with-bwpass and with a single failure in the bwpass owstem that doos not Anelude turbine trip without bwpass.

Ft the request of the NRC. Edison performed a sensitivitw analvsis of

~

turbine generator trip with postulated failure of bwpass flow to the t

o candenser. See UFSAR Subsection 15.2.3.1.E.3. Turbine Trip Without Bwpass, for details.

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As discussed in subsection 10.4.4, Fermi 2 kas a unique bypass Imsas of the bypass system would require multiple random system.

failures.4, Edison maintained that this is very unlikely and need l

os os conv3dered part of the design basis of the Ferah 2 plant.

.Therefore, the analysis of turbine generator trip described in l

this section includes cases with bypass and with a single failure 4

in the bypass system but does not include turbine trip without bypass.

l q

i i

At the request of the NRC, Edison performed a sensitivity i

I analysis of turbine generator trip with postulated failure of bypass flow to the condenser.

The analysis was based on input parameters specified in Table 15.0-1 and used the CDYN computer i

i~*/

code.

The resul ar===--rimed in Table 15. 2. 3-1 anA..,ir~P At.

i o

.tiaure 15.2.

YnK Tic Muic.n srecopic.Mfiw et.4 M l

1 ama ato o TM st A tsuLT.S AND TNC p.e9 HQWDeAL GAAOR eh!Me j

onal conuriia'Gry m.ThAN analyses or renester steam Indicated that the reheater steam flow relation ohnwn in Figure l

15.0-5 and used in the analysis of turbine generator trip is conservative.

The results also indicated that the peak fission power occurs within 1 see of the start of the turbine trips of 4

4 interest.

Therefore, the primary effect of reheater steam flow l

j is seen during the early stages of the transient, probably within the first se:ond.

Thus, the espected early peaks in reheater l

flow calculated in these supplemental analyses would be espected to reduce peak reactor power and heat flus compared with the reheater steam flow relation shown in Figure 15 0-5, confirming l

the conservatism of the use of Figure 15.0-5.

1 15.2.3.2 secuence of Events and systems operation

)

i 15.2.3.2.1 secuence of Events

.urbine Generator Trip 15.2.3.2.1.1 A turbine trip at high power produces the sequence of events listed in Table 15.2.3-2.

15.2.3.2.1.2 Turbine Generator Tris With single Failure in the

.v. ass avet.m h turbine trip at high power with a single f ailure in the bypass system (failure of one bypass valve) produces the sequence of events listed in Table 15.2.3-3.

15.2.3.2.2 systems operation I

l 15.2.3 2 2.1 Turbine Generator Trip With typass h

All plant control systems maintain normal operation unless spe-i\\ A emily designated to the conte ry-

• p. flow through the g -

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Option A 1.2 Option B

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[ } Operating License [ ] Tech Specs [ ] Environmental Protection Plen r

1 Tech Spec Clarification I

B) ILection(s), Table (s) Figure (s), etc. Affected (Attach marked-up pages) l C) Reference and Source Documents Attached S

[ ] Significant Harards Consideration

[ ] Environmental Evaluation

[ ] Enviromental impact / Categorical Exclusion

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I U

)Yes [ } No LCR No.

S E) l'riority k

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(State date):

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e'

1. b[nu,,Date b

H) OSRO Approval (Tech Spec Amendments [

NA Date 9 2 8I I) NSRG Approval (Operating License Amendments)

/)NA Date

• 6//4 torm FIP-RA2-01 Att 1 P1/1030189 DTC: lE TCLCR for UFSAR File: 1735

{

DTC: [ ] TDLCR for other

--.-.\\

ePFECTfVENESS mEWEW meterence LCR I i i f I-l/ 1 # 1Fl-IvlrIsl Q " - - - - - - - - - " " - -

" " " - - - - - " PART 1 :

R. vision # Pag. 2 of 9 F --

UFSAR PQNA -

- ---- l l

A) Quality Assurance Program

[ IYes [ ] No Does the change (s) cease to satisfy the eftteria of 10CFR50, Appendix B i

and the UFSAR program commitments previously accepted by the NRC7 Provide the basis for occh change on Attachment 2, Page 2 B) P6re Protection Program

[ IYes [ ] No Does the change (s) significantly decrosse the level of fire protection in the plant?

[ IYes [ ] No Does the change (s) result in failure to complete Fire Protection Program opproved by the NRC prior to license issue?

Provide the basis for each change on Attechment 2, Page 2.

!""PART 2: RADIOLOGICAL EMERGENCY RESPON5E PREPAREDNES PLAN [/ENA -l A) [ ] Yes [ ] No Does the change (s) decrease the effectiveness of the RERP Plan?

! ] Yes [ ] No Does the RERP Plan, as changed, cease to meet the standards of 10CFR50.47(b) and 10CFR50 Appendix E?

' Provide the bests for each change on Attachment 2, Page 2.

4 p n -- " -- " "-- " ="" PART 3: SECURITY PLANS LPQNA ----------- "- ---- ----- " """"- l A) Document A) [

. ] Yes [ ] No Does the change (s) decrease the effectiveness of the Physical Security Plan or Security Personnel Training and Quellfication Plan prepared i

pursuant to 10CFR50.34(c) or 10CFR737 i

[ ] Yes [ } No Does the change (s) decrease the effectiveness of the first four categories of Informational Background, Generic Planning Base, Licensee Planning O

Base, and/or responsiblity metrix of the Safeguards Contingency Plan I

propered pursuant to 10CFR50.34(d) or 10CFR737 Provide the basis for each chen 3e on Attachment 2, Pas e 2.

F" " = "

"""" "" PART 0: PROCESS CONTROL PROGRAM [X]N#"*"****"*

l A) [ ] Yes [ ] No Does the change (s) reduce the overall conformance of the solidified weste product to existing criteria for solid westos in accordance with Technical Specification 6.137 Provide the basis for each change on Attachment 2, Pele 2.

F "- - "- " " -- - - - - ""-- "- " PART 5 : ODCM LK. M A """ -

---

"----""l A) [ ] Yes [ ] No Does the change (s) reduce the accuracy or reliability of the dose l

calculations or setpoint determinations in in accordance with Technical l

Specification 6.147 Provide the basis for each chan 3e on Attachment 2. Page 2.

F - - - - - -- - - - - - - - - - - = = P ART 0: APPROvAtta==="m===amaa====*ma l A) Originator IA Mh'

/ * ~2M/9

[

Date f 11 8 f

3) Technical Empert -

d Date C) Ouality Assurance (For Security Plans only) A//4 Date f4(,A 9 lD) OSRO (Not required for UFSAR Changes)

1. /#

Date 9/m/f-9l crm FIP-RA2-01 Att 21/2 030189 DTC: N TCLCR for UFSAR File: 1735 DTC: [ ] TDLCR for other j

vf BPPECTfVENESS REVIEW DOCUMENTAft0N Referense LCR 1 II TI-1/1 lf I f l -

l til F l 8 l Revision # Pese 3 of 7 Document Listed below is each change by section and page; the rosson for the thenge; and the basis for Concluding that the revised plan or program cont 6nues to 6ettsfy the cetterle for that plan or program.

Beetlen/Page Chenge Beele 48 8'3 l0'3 A49,5I,5SlC's (CAptNG bATA od EDG.t

/ER CAtcut Aftod Mo.5003 U^*"A 9 byb,4,t.if ?) ARE VPp4 7'*D PsR. cALcu1A-g,,, g 7,74 g t,p g o g 4.u.gt Tsod No 5003 KEV $.

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PERMi t pFSAR 1

TABLE 4.3-5 ENERGENCY M EstL GENERA"0R 14ADING SE00ENCE:

LOSS-0 '-C0Y. ANT ACCIDEN" AND

.055 0F 0 'FSITE FOWER

' DIW EION I EDG5

,1 e 12 J (DIVI 5 ION II EDG5 I De a J

Time (se'c)

Overall E

EDG 11 (13)

EDG 12 (14)

Accident occurs Accident occurs I

0 Diesel starts Diesel starts

• 3 i

s.

Rated speed and voltage Rated speed and voltage 13 13 0

EDG breaker closes EDG breaker c1'oses 13 0

Auxiliary 400-V Auxiliary 400-V s

transformers energised, transformers energised instrumentation l

13 0

RHR pumps and MOVs RER pumps and MOVs 18 5

Core spray pumps and Core spray pumps and MOVs MOVs Emeroency lichtina/Eisl[ce**

18 5

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• 30 25 ECCS and auxiliary room ECCS and auxiliary room cooling cooling 48 35 Air compressor and dryer l

58 45 EDG service water and EDG service water and i

auxiliaries auxiliaries l

68 55 control center air l

conditioning fans and FER TN15 LeR chiller, control room recirculation emergency I

makeup tan

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