ML17309A201

Advises That NRC 810826 Draft Evaluation of SEP Topic XV-9, Startup of Inactive Loop,Flow Controller Malfunction, Reviewed.Assessment Accurately Represents as-built Condition at Plant

ML17309A201
Person / Time
Site:	Ginna
Issue date:	10/16/1981
From:	Maier J ROCHESTER GAS & ELECTRIC CORP.
To:	Crutchfield D Office of Nuclear Reactor Regulation
Shared Package
ML17258A243	List: ML17258A242 ML17309A201
References
RTR-NUREG-0485, RTR-NUREG-485, TASK-15-09, TASK-15-9, TASK-RR NUDOCS 8110210143
Download: ML17309A201 (29)

Text

air

~>>r

~ ~

0

(

r/ e

~ <<>>,

a,'t,e>>

e>> >i4NUI a>>>>4 eJJr'll ROCHESTER GAS AND ELECTRIC CORPORATION

~

89 EAST AVENUE, ROCHESTER, N.Y. 14649

>>0&N c,.

IIAIER IICr PRESIDKNT reeepmONe acre COOe

. ia 5 "6-2700 October 16, 1981 Director of Nuclear Reactor Regulation Attention:

Mr. Dennis M. Crutchfield, Chic.f Operating Reactors Branch No.

5 U.S. Nuclear Regulatory Commission Washington, D.C.

20555

Subject:

SEP Topic ZV-9, Startup of an Inactive Loop

~ R.E. Ginna Nuclear Power Plant Docket No. 50-244

Dear Mr. Crutchfield:

We have reviewed the draft topic evaluation for SEP Topic XV-9, which was provided by your letter dated August 26,

1981, and concur that the assessment accurately represents the as-built condition of Ginna.

Very truly yours, Sii02iOi43 Si10i6 PDR

• DOCK 05000244 P

PDR gffj(l'tI'.,

OCT2 0 tg8)g MISS tQCuAa ~gag RQAWLlQ~

<P,S RE00<Cg O~

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 August 26, 1981 Oocket No. 50-244 LS05-81-08-059 Nr. John E. Maier Vice President Electric and Steam Production Rochester Gas and Electric Corporation 89 East Avenue Rochester, New York 14649 Oear iver. Maier:

SUBJECT:

SEP TOPIC XV-9, STARTUP OF AN INACTIVE LOOP R. E. GINNA p(j P t'ai

~Igg~~lQ~

P Enclosed is the draft topic evaluation for XV-9.

This evaluation compares your facility with the criteria currently used by the regulatory staff for licensing new facilities.

Please inform us if you as-built facility differs from the licensing basis

~

-assumed in our assessment within 30 days of receipt of this 'letter.

This evaluation will be a basic input to the integrated safety assessment.

for your facility unless you identify changes. needed to reflect as-built conditions at your facility.

This topic assessment may be revised in the future if your facility design is changed or if NRC criteria relating to this topic are modified before the integrated assessment is completed.

Sincerely,

Enclosure:

As stated Operating Reactors Branch o.

5 Oivision of Licensing cc w/enclosure:

See next page

Hr. John E. Maier cc Harry H. Voigt, Esquire

LeBoeuf, Lamb, Lei:-.y and MacRae 1333 Hew Hafrpshire Avenue, N.

W.

Suite 1100 Mashington, D. C.

20036 Nr. thichael Slade 12 Trailwood Circle Rochester, New York 14618 Ezra Bialik Assistant Attorney General Environmental Protection Bureau Hew York State Department of Law 2 World Trade Center Hew York, New York 10047 Jeffrey Cohen Hew York State EnerrD Office Swan Street Building Core 1, Second Floor Empire State Plaza

Albany, Hew York 12223

Director, Bureau of Nuclear Operations State of New York Energy Office Agency Building 2 Empire State Plaza

Albany, New York.

12223 Rochester Public Library 115 South Avenue Rochester, New York 14604 Supervisor of the TQ'wn of Ontario 107 Ridge Road Mest

Ontario, New York 14519 Resident Inspector R. E. Ginna Plant c/o U. S.

NRC 1503 Lake Road

Ontario, Hew York 14519 Hr. Thomas B. Cochran Natural Resources Defense Council, Inc.

1725 I Street, N.

W.

Suite 600 Wash'ington, D. C.

20006 U. S. Environmental Protection Agency Region 11 Office ATTN:

E IS COORDINATOR 26 Federal Plaza New York, New York 10007 Herbert, Grossman, Esq.,

Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Comnission Mashington, O. C.

20555 P

Or. Richard F. Cole Atomic Safety and Licensing Board-U. S. Nuclear Regulatory Commission Washington, O. C.

20555 Or.

Emmeth A. Luebke Atomic'Safety and Li'censing Board U. S. Nuclear Regulatory Commission Mashington, D. C.

20555

R. E.

GINNA TOPIC XV STARTUP OF AN INACTIVE LOOP I.

INTROOUCT ION The startup of an inactive coolant loop is examined to assure that the introduction of cooler or deborated water into the core does not lead to an unacceptable loss of fuel clad integrity or overpressurization of the primary system.

The guidelines used are contained in SRP Section 15.4.4, Startup of an Inactive Loop, at an Incorrect Temperature.

II.

EVALUATION The startup of an inactive loop at power results in a core reactivity increase when the colder water of the idle loop reaches the core.

Reactor protection for th'is event is provided mainly by administrative procedure and the inherent stability of the core.

Technical Specification 3.1.1.1 permits operation up to 130 Hwt (8.5'X of rated power) with only one reactor coolant pump operating.

An orderly power reduction is required to below 8.5% power if a pump is lost while operating at a higher power level.

Above 50% power loss of a reactor coolant pump will cause a direct reactor trip.

The licensee provided an analysis of this event in the FSAR (Reference 1).

An analog computer representation of the primary system was used.

The initial conditions and assumptions used in the analysis include:

1.

inactive loop is 20 F cooler than the active loop; 2.

large negative moderator temperature coefficient; 3.

small doppler coefficient; 4.

high heat transfer coefficient between primary and secondary so that the cold leg of the inactive loop is at saturation temperature for the steam generator secondary; 5.

instantaneous start of idle pump; and 6.

mixing with the active loop flow in the plenum.

Although no uncertainty was applied to the initial power level, these assumptions are judged to be sufficiently conservative for this analysis.

The 20'F dT was selected on the basis that at the low power level per-mitted for one-loop operation, the difference between the temperature in the inactive loop and the temperature in the active loop will not be great.

The effect of the other assumptions, such as the temperature coefficient, is to augment the reactivity and power increase caused by the cold slug.

Results of the analysis plotted in the Ginna FSAR indicate a temperature and pressure decrease of 10'F and 30 psi, respectively.

The power increased, but a pressure spike was not generated because of coolant contraction due to the temperature decrease.

The reactor stabilizes without actuation of any automatic protection features or dependence on operator action.

The relatively minor perturbations in system parameters, and the reduced power condition of the core ensure that the operating limits are not exceeded.

The effects of. startup of an inactive loop are less severe than those due to a small steam line break with one loop operable, which has been analvzyg by Mestinghouse as showy in Reference 2~

.To accommodate the steam tine break,'Technica)

Specification 3.i.i.).b requires Chat a

higher shutdown margin be maintained for one-loop operation.

Operation with less than all loops in service is the subject of SEP Topic IV-1.A, Nhich was completed by Reference 3.

III.

CONCLUSION As a part of the SKP review of Ginna, the analysis of Startup of an Inactive Loop at an Incorrect Temperature has been evaluated against the criteri a of SRP Secti on 15.4.4.

Based on thi s eval uati on, we have concluded that the consequences of startup of an inactive loop have been adequately addressed by the licensee and the acceptance criteria are satisfied.

REFERENCES 1.

"R. E. Ginna Nuclear Power Plant Final Safety Analysis Report," Section 14.1.7; September 1969.

2.

Leboeuf, Lamb, Leiby and HacRae, for Rochester Gas and Electric Corpora-tion, "Application for Amendment to Operating License;"

September 22, 1975.

3.

Letter to D. Ziemann (NRC) to L. White (RGSE);

May 29, 1979.

4y g

app

)

cl

.Ir V/+0 00 0y**4 Docket No. 50-244 UNITED STATES NUCLEAR REGULATORY COMMISSlON WASHINGTON, O. C. 20555 October 1, 1980 Hr. Leon D. White, Jr.

Yice President Electric and Steam Production Rochester Gas and Electric Corporation 89 East Avenue Rochester, New York 14649

Dear fir. White:

~ '

OD1C II-2. 8 Cate orization Generic D1 s5051 t1 on Comp,liance with A'ppendix E is being evaluated under the Emergency Pr epardness Program Office (EPPQ) effort.

Compliance wiih Appendix I is being evaluated as part of the NRC Appendi x I revi ew ef ort-8y letter dated November 16,

1479, we sent you the list of tonics which would not be reviewed by your plant in the Systematic Evaluation Program (SEP).

These topics were deleted from the SEP review because they were not applicable to your plant or they were being revi ewed as part of an ongoing generic issue outs1de the SEP.

The following topics should be added to the hovember 16 listing.

Removal of these additional topics from the SEP review results

rom 1) related Generic NRC activities and 2) non-applicability of topics to certain facilities.

The following is a list of such topics and indicates their disposition:

'-2.D Generic Compliance is beino evaluated under the EPPG effort.

4 ~

Del eted Th s tpp'.c was orici..ally intended for 8!R'."

VI-e XV-11 ln Generic Deleted Deleted Mill be reviewed by l:RC as part of Till Tasr Action P 1 cn(:l"REu-0660) as specified in i!ay 7, 1980 letter to all operating reactor licensees.

This item applies specifically to SWR's.

Thi s iteI.", applies spec', fica 1 ly tp 8l!R'.

The radi ol ogi ca 1 cor sequences of steam line breaks for PM?'s is part of Topic,

~ ~ ~

l ~

a 0

Hr. Leon D. White

<<2>>

October 1, l980 The following two topics will not be reviewed by SEP in their entirety due to overlap with other NRC activities:

Disoosition Those portions dealing with plant emergency plans will be evaluated as part of the EPPO effort.

Those portions of the auxiliary systems being reviewed as part of Lessons

Learned, the Tf>7 XV-7 XV-7 XV-8 XV-8 XY-8 XV-'l2 DESIGN BASIS EVENT Decrease in feedwater temperature Increase in feedwater flow Increase in steam flow Inadvertent opening of steam generator relief/safety valve Start-up of inactive loop System malfunction causing decrease in boron concentration Loss of external load Turbine trip Loss of condenser vacuum Steam pressure regulatory failure Loss of feedwater flow Feedwater system pipe break Steam line break inside containment Steam line break outside containment Loss of AC power to station auxiliaries Loss of all AC power Loss of forced coolant flow Primary pump rotor seizure Primary pump shaft 'break Uncontrolled rod assembly with-drawal at power Uncontrolled rod assembly with-drawal low power start-up Control rod misoperation Spectrum of rod ejection accidents SRP 15.1.1 15.1.2 15.1.3 15.1.4 15.4.4 15.4.6 15.2.1 15.2.2 15.2.3 15.2.5.

15.2.7 15.2.8 15.1.5 15.1.5 15.2.6 15.3.1 15.3.3 15.3.4 15.4.2 15.4.1 15.4.3 15.4.8

~Grou VII XV-19 Spectrum of loss of coolant accidents 15.6.5 Grouo VIII

~Grou It

~Grou 'I XV-21 XV-20 XV-15 XV-14 Drop of cask or heavy equipment Radiological Consequences of Fuel Damaging Accidents (inside and outside containment)

Inadvertent opening of PMR.

pressurizer relief valve Inadvertent operation of ECCS or

'VCS malfunction that causes an incr ease in coolant inventory 15.7.5 15.7.4

'15.6.'1 15.5.1 15.5.2

Enclosure 6

~Grou XI

~Grou XII RELATED TOPIC XV-17 DESIGN BASIS EVENT Not applicable to PlIIRs Steam Generator Tube Failure SAP 15.6.3

TOPIC Xtj'-j2(SYSTB5)

SEE TOPIC XV-1 FOR SYSTEMS EVALUATION

TOPIC XV-32(IjOSES)

SEE TOPIC XV-2 FOR DOSE EYALUATION

TOPIC XV-D SEE TOPIC II-4.E

TOPIC X/-16 SEE TOPIC XV-2 FOR DOSE EVALUATION

TOPIC XV-17(SYSTENS)

SEE TOPIC XY-1 FOR SYSTEMS EVALUATION

TOPIC N/-17(KSES)

SEE TOPIC XV-2 FOR DOSE EVALUATION

TOPIC N/-18 SEE TOPIC II-2.B

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 October 2, 1981 Docket No. 50-244 LS05 10-001 Mr. John E. Maier, Vice Presidert Electric and Steam Production Rochester Gas It. Electric Corporation 89 East Avenue Rochester, New York 14649

Dear Mr. Maier:

SUBJECT:

R.

E.

GINNA - SEP TOPIC XV-19 (SYSTEMS)

LOSS OF COOLANT ACCIDENTS RESULTING FROM SPECTRUM OF POSTULATED PIPING BREAKS WITHIN THE REACTOR COOLANT PRESSURE BOUNDARY On July 27, 1981 we transmitted to you a draft assessment of SEP Topic XV-19 (systems)..

In your letter of September 15, 1981 you provided comments in the form of a revised topic assessment.

The staff has evaluated the suggested revisions and we consider that they provide substantial additional information, but do not alter the staff's conclusions.

Therefore, we will use your revised assessment (enclosed) as our final evaluation.

We now consider Topic XV-19 (systems) to be complete.

The enclosed safety evaluation will be a basic input to the integrated safety assess-ment for your facility.

The assessment may be revised in the future if your facility design is. changed or if NRC criteria relating to this topic are modified before the integrated assessment is completed.

Sincerely, Enclosure;"-

As stated cc w/enclosure; See next page Dennis M. Crut hfield, Chief Operating Reactors Branch No.

5 Division of Licensing

~--

Mr. John E. Maier CC Harry H. Yoigt, Esquire

LeBoeuf, Lamb, Leiby and MacRae 1333 New Hampshire

Avenue, N. W.

Suite 1100 Mashington, 0. C.

20036 Mr. Michael Slade 12 Trailwood Circle Rochester, New York 14618 Ezra Bialik Assistant Attorney General Environmental Protection Bureau Hew York State Department of Law 2 World Trade Center New York, New York 10047 Jeffrey Cohen Hew York State Energy Office Swan Street Building Core 1,.Second Floor Empire State Plaza Albany, Hew York 12223 Director, Bureau of Nuclear Operations State of New York Energy Office Agency Building 2 Empire State Plaza

Albany, New York 12223 Rochester Public Library 115 South Avenue Rochester, New York 14604 Supervisor of the Town of Ontario 107 Ridge Road West

Ontario, Hew York 14519 Resident Inspector R. E. Ginna Plant

,c/o U. S.

gRC 1503 Lake Road

Ontario, New York 14519 Mr. Thomas B. Cochran Natural Resources Defense Council, Inc.

1725 I Street, N. M.

Suite 600 Mashington, D. C.

20006

~ ~

U. S. Environmental Protection Agency Region II Office ATTN:

Regional Radiation Representative 26 federal Plaza New York, New York 10007 Herbert Grossman, Esq.,

Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Comnission Mashington, 0. C.

20555 Dr. Richard F. Cole

.Atomic Safety and Licensing Board U. S. Nuclear Regulatory Comnission Mashington, D. C.

20555 Or. Emmeth A. Luebke Atomic Safety and Licensing Board U.

S Nuclear Regulatory Comnission Mashington, D. C.

20555

~

R E ~

GINNA SEPTEMBER'981 TOPIC XV-19:

LOSS OP COOLANT ACCIDENTS RESULTING PROM SPECTRUM OP POSTULATED PIPING BREAKS WITHIN THE REACTOR COOLANT PRESSURE BOUNDARY INTRODUCTION The capability of the R

E-Ginna Emergency Core Cooling

~ System to miti'gate the consequences of a spectrum of Loss of Coolant Accidents (LOCAs) is evaluated to assure that pipe breaks in the reactor coolant system (RCS) do not result in a loss of core cooling capability.

Detailed acceptance criteria for Emergency Core Cooling System (ECCS) performance are contained in 10 CPR 50.46 and in Standard Review Plan Sections 15.6.5, 6.3 and supporting appendices.

The five main criteria for accep-tance are:

~,

1 Peak clad temperature less than 2200oF 2-Maximum cladding oxidation Less than 17%

3.

Total hydrogen generation less than LS of total zirconium in the active fuel region 4.

Maintenance of eoolable geometry 5.

Long term coo'lability A spectrum of break sizes up to and including a double-ended break of the largest pipe at various break Locations is examined using an approved eval.uation model. which conforms to the requirements of Appendix K to 10 CPR 50 to verify that the acceptance criteria are met for a variety of postulated loss of coolant accidents.

II EVALUATION The Ginna power plant ECCS provides emergency core cooling

- waMr at three delivery pressures.

The high pressure

safety injection (HPI) system delivers borated water at up to 1400 psi (see Pig. 2-1 of Ref.

8 for SI pumped flowrate as a function of RCS pressure asuming 5't degradation from design head).

This is different than current PWR's which use safety grade cha'rging pumps which are capable of injection at operating pressure, about 2235 psi.

Intermediate pressure passive injection is provided by the accumulators which are held at 700 psi by nitrogen

gas overpressure.

The HPI system and thy accumulators discharge into lines to each cold legs Low pressure cooling water from the refueling water storage tank is delivered by the residual heat removal (RHR) system which becomes available at 140 psi.

The low pressure injection flow is pumped directly into the upper plenum of the reactor vessel through two separate nozzles.

More complete descriptions of these systems are provided in the Ginna safe shutdown report, (Ref. 1), and in the Final Safety Analysis Report (Ref. 7).'witchover from injection to recirculation mode is covered in SEP Topic VI-7.B.

The Ginna core currently contains 117 fuel assemblies designed and fabricated by Exxon Nuclear Company and 4 mixed oxide fuel assemblies designed and fabricated

'y Westinghouse Electric Corporation.

Analysis for the large pipe breaks was performed by Exxon Nuclear for the Cycle 8 fuel reload in Reference 2, with the staff evaluation presented in Reference 3.

The limiting large break was reanalyzed in January 1980 (Ref. 9) to include an NRC model for fuel clad swelling and the. incidence.of fuel clad rupture.

The LOCA evaluation for the 4 Westinghouse assemblies is presented in Refs.

10 and ll with NRC approval in Ref. 12.

The effect. of the low pressure injection point being the vessel upper plenum i.nstead of the cold legs is addressed in SEP Topic VI-7.A.2.

This topic has been deleted from.consideration in SEP since it is generic.

The small break analysis was performed by Westinghouse for Ginna during the initial Appendix K reviews (Ref. 8).

Since the small breaks were clearly demonstrated to be non-limiting, later reloads re-evaluated only the large break spectrum.

In response to the NRC's Bulletins and Orders Task Force, additional small break analyses were performed on a generic basis.

The Westinghouse calculations of Reference 4 were reviewed by the staff in Reference 5.

--Lar e-Break Anal sis'he cycle 8 fuel reload safety analysis (Reference 2) examined six different pipe breaks in the cold leg.

Hot leg breaks were not examined.

Three double area guillotine breaks with discharge coefficients (CD) of 1.0; 0.6, and 0.4 were analyzed.

The other three breaks

considered were split breaks with discharge coefficients of 1 ~ 0, 0.6, and 0 ~ 4.

The selection of breaks for this analysis was justified, based on previous evaluations, which clearly identified the cold leg split and guillotine breaks as the most limiting.

The assumptions and computer codes used in the

LOCA, analyses are covered in Reference 2.

Some of the more important assumptions include:

1.

Initial power at 102'5 2.

Reactor trip is neglected for large breaks 3

All accumulator water bypasses core until termination of bypass 4.

Linear Heat Generation Rate of 13.76 kw/ft 5.

Total peaking factor is 2.32 6

Fuel at beginning of life (BOL) conditions

'These and the other assumptions used for these analyses were in accordance with 10 CFR 50.46 and Appendix K, 10 CPR Part 50 and have been shown (Ref. 2) to result in conservatively high peak cladding temperatures.

Results The limiting case of the six breaks examined in the ECCS analysis for Ginna presented in Ref.

2 was the double-ended cold leg guillotine break with CD=0 4.

The peak clad.temperature predicted was 1922 F, con-siderably below the 10 CFR 50.46 limit of 2200 P.

Clad oxidation (peak and total) was also well within limits.

It should be noted that guillotine breaks with a discharge coefficient smaller than 0.4 are not required in accordance with Reference 6.

The analyses to determine the effect of using the NRC's model for fuel clad swelling and the incidence of clad rupture was performed using the models described in References 13-16.

As described in Ref. 9, the revised model for fuel clad swelling and the incidence of rupture resulted in a peak clad temperature increase of 1 F for Exxon Nuclear fuel.

-'Phut, analyses-presented in Ref. 2"remain valid.

Small Break Anal sis As discussed above, plant-specific small break analyses were not performed by Exxon Nuclear because it had been shown in previous Westinghouse analyses for Qinna (Ref. 8) that the small breaks would not be the limiting case.

Westinghouse analysis yielded a limiting small break

size of 4 inch diameter with a peak clad.temperature of 1688o Small Break Anal ses - Post TMX Generic analyses of small break LOCAs were submitted by the Westinghouse Owners Group in response to NRC Bulletins and Orders Task Force requirements.

The staff has accepted these analyses as a basis for providing information on plant response and as an aid to developing guidelines for operator action.

The generic analyses included consideration of the reduced head HPZ system.

The staff considers these generic analyses to be repre-sentative of the response for Ginna to a postulated small break LOCA.

Results - Small Break - Post TMX As a result of the review of these analyses, the staff expressed concern about the applicability of current evaluation models and their application to the expanded scope of small break LOCA analyses now being considered.

As part of the TMX Task Action Plan, which is beyond the scope of the SEP review, Westinghouse is to revise and resubmit the small break analysis methods for staff approval.

Plant specific calculations, using these revised methods will then, be required to show compliance with 10 CPR 50.46.

These analyses should place special emphasis on accidents which actuate the HPX-ZZZ. CONCLUSION The loss of coolant accidents analyzed for the Ginna nuclear power plant meet the acceptance criteria.

New" small break LOCA analyses using revised evaluation modelswill be conducted as part of the TMX Task Action Plan and will not, be included as part of the SEP review.

The impact of upper plenum low head safety injection is being conducted by review of SEP Topic VI-7.A.2, "Upper Plenum Injection" and NRR Generic Task D-05.

Zt is thus not included as part of this SEP topic.

REFERENCES 1.

Safe Shutdown Systems for R.

E. Ginna Nuclear Power

Plant, SEP Topic VII-3, May 13, 1981.

2.

"ECCS Analysis for R. E. Ginna Reactor with ENC WREM-IZ PWR Evaluation Model."

Exxon Huclear Company Report, XH-NF-77-58, December 1977.

3.,

"R. E. Ginna Nuclear Power Station Cycle 8 Reload Safety Evaluation Reports

May 1, 1978.

4 "Report on Small Break Accidents for Westinghouse HSSS System" Westinghouse Nuclear Energy Systems

Report, WCAP-9600, June 1979 5.

"Generic Evaluation of Feedwater Transients and Small Break Loss of Coolant Accidents in Westinghouse Designed Operating Plant",

NUREG-0611, January 1980.

6-Status Report by the Directorate of Licensing in the Matter of Westinghouse Electric Company ECCS Evaluation

'Model Conformance to 10 CFR Part 50, Appendix K, October 15, 1974 and the Supplement, dated November 13, 1974.

7.

"R. E. Ginna Nuclear Power Plant Final. Safety'Analysis Report" September 1969.

8.

Application dated September 3,

1974 and submitted September 6,

1974 from RQ&E to the HRC.

9.

Letter dated January 10, 1980 from L. D. White, Jr.,

RQ&E to Dennis L. Ziemann, USHRC re ECCS Models.

10.

Application dated December 14, 1979 and submitted December 20, 1979 from RG&E to the HRC.

11.

Letter dated February 20, 1980 from L. D. White, Jr.,

RG&E to Dennis L. Ziemann, HRC 12.

Amendment No. 32 to the Ginna license transmitted by letter dated April 15, 1980 from Dennis L. Ziemann, NRC, to L.

D-White, Jr.,

RQ&E.

13...",Exxon Nuclear Company NREM-Based-Generic PWR.ECCS Evaluation Model Update ENC WREN-IIA," XH-HF-78-30, August 1978.

14.

"Exxon Huclear Company WREM-Based Generic, PWR ECCS Evaluation Model, " Zf-75-4'1:

a.

Volume I, July 1975 b.

Volume ZZ, Augus 1975 c.

Volume ZZZ, Rev'sion 2, August 1975

'd ~

e ~f.

go h

1, ~

I Supplement 1,

Supplement 2,

Supplement 3,

Supplement 4,

Supplement 5,

Supplement 6,

Supplement 7,

August 1975 August 1975 August 1975 August 1975

'Revision 5, October 1975 October 1975 November 1975 15.

"Exxon Nuclear Company >hEM-Based Generic PWR. ECCS Eva1uat3.on Model Update ENC WREM-II," XN-76-27, July 1976; Supplement 1, September 1976; Supplement 2, November 1976.

16 "Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model Updated ENC WREM-IIA:

Responses to NRC Request for Additional Information," 3QT-HF-78-30(A) I XN-NF-78-30, Amendment 1(A), May 1979.