ML19341C989

Evaluation of Operation of Beaver Valley Power Station, Unit 1 W/One Loop Isolated, Technical Evaluation Rept

ML19341C989
Person / Time
Site:	Beaver Valley
Issue date:	02/28/1981
From:	Lyon R EG&G IDAHO, INC., EG&G, INC.
To:	Panciera V Office of Nuclear Reactor Regulation
References
CON-FIN-A-6267 EGG-EA-5350, TAC-10386, NUDOCS 8103040661
Download: ML19341C989 (16)

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% Assistance Report EGG-EA-5350 February 1981 EVALUATION OF OPERATION OF BEAVER VALLEY POWER

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. This is an informal report intended for use as a preliminary or working document Prepared for the U.S. Nuclear Regulatory Commission Under 00E Contract No. DE-AC07-76ID01570 0 g g E g 'd8ho FIN No. A6267 $Q 81030406W

Ass,s_.,_ N1C Research am.' Technical Flev 18 m

. N Assistance Report INTERIM REPORT Accession No.

Report No. EGG-EA-5350 Contract Program or Project

Title:

(N-1) Loop Operation of Beaver Valley and Zion 1/2

Subject of this Document:

Evaluation of Operation of Beaver Valley Power Station Unit 1 with One Loop Isolated Type of Document:

Technical Evaluation Report

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Author (s): C ,r-R. E. Lyon .3q 4 08., ! [(e-r ,/g

~f Date of Document: .%

February 1981 \

.;. O Responsible NRC Individual and NRC Office or Division:

V. W. Panciera, NRC-DSI This document was prepared primarily for preliminary or internat use. it has not received full review and approval. Since there may be substantive changes, this document should not be considered final.

EG&G Idahc, .'nc.

Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C.

Under DOE Contract No. DE AC07-761D01570 NRC FIN No. A6257 INTERIM REPORT e

CONTENTS 1.0 INht000CT, ION .................................................... 1 2.0 TECHNICAL DISCUSSION ............................................ 2 3.0

SUMMARY

AND RECOMMENDATIONS ..................................... 7

4.0 REFERENCES

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1.0 INTRODUCTION

The Beaver Valley Power Station Unit No.1, operated by Duquesne Light Co., is a three loop PWR designed by Westinghouse. The reactor coolant system is equipped with loop isolation valves wnich can isolate a steam generator and reactor coolant pump from tna primary system, permitting continued operation at reduced power with one loop snut down. Beaver Valley nas been operating with a condition in their operating license whicn prevents them from operating with a loop isolated. Duquesne Light Co. has submitted a reouest for amendment to their operating licensel which provides analyses to establisn tne response of tne plant to transients and accidents wnen operating witn one loop isolated. This report documents the review which nas been performed on those analyses and the acceptability of permitting Beaver Valley Unit No.1 to operate with one loop isolated.

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2.0 TECHNICAL DISCUSSION Ine Beaver Valley Power Station Unit No.1 has been analyzed to determine tne response of the plant to specific transients when operating with one loop isolated. Tnese analyses are documented in either, or both, -

of two locations. Some of tne analyses were documented in the Final Safety Analysis Report (FSAR) and additional analyses were submitted in License Amendment Recuest No. 35.I The results of these analyses are discussed in tne following sections. The analyses are grouped into three categories, s was done for the Final Safety Analysis Report, and, in general, have been submitted and reviewed for isolated loop operation to the same level of detail as was done for normal operation.

2.1 Core and Coolant Boundary Protection Analysis Analyses in tnis area cover minor transients which should result in, at most, a reactor trip and whicn should not result in any loss of function of fuel cladding and reactor coolant system boundaries. Tne analyses wnich .

were performed are listed in Tab?e 1. The basic acceptance criteria for the majority of tnese transients, as given in tne Standard Review Plan,3 are tnat tne reactor system pressure remains below 110% of design pressure and tnat the DNB Ratio remains above 1.30, therefore these parameters are listed in Table 1, along with other parameters which may be of interest.

A review of the results of the analyses shows that, for all transients wnere data is given, the maximum RCS pressure is less than 1l0% of design ,

i pressure and the DNBR is greater than 1.3. In general, the transients for ~

isolated loop operation are similar to, but slower and less severe, than tne corresponding transients for normal operation. Tne peak pressure is lower in all cases except for the excessive load increase, where it is sligntly ($10 psi) higner for isolated loop operation. Maximum pressure .

for botn isolated loop and normal operation was a result of the loss of load witn turbine trip.

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1 l-I Although the decrease in DNBR was greater for some isolated loop

. transients than for the corresponding normal configuration, the minimum DNBR reached was always higner for the isolated loop, due to the nigher initial DNBR (2.2 vs. 1.7). The minimum DNBR tor botn isolated loop and

normal operation was.for the uncontrolled rod cluster control assembly witndrawal at power.

Tnere are two areas whicn may require additional information from tne

, licensee. Tne first of tnese is tne uncontrolled boron dilution i transient. Inis transient nas-been analyzed for the normal operating mode j out not for the isolated loop. The acceptance criteria for this transient as given in SRP 15.4.5 is tnat during power operation, a minimum of h 15 Ininutes must be available from the time an alarm makes tne operator

! aware that tne dilution is in progress until loss of snutdown margin occurs. For normal operation, the FSAR states that 15 minutes is the time

, for recriticality after reactor trip. For operation with an isolated loop,

this time would be somewhat reduced because of the smaller volume being j- diluted.

] The second area concerns the accidental depressurization of the

reactor coolant system. This transient has also been analyzed for normal

{ operation but not for isolated loop operation. Tnis is not the limiting transient for normal operation and conseouently is probably not a problem

! for isolated loop operation, however, there is not sufficient information

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to confirm this. Also, there nave been no corresponding size small break analyses which nave been performed and which might give some indication as j to tne acceptable response to tnis transient.

2.2 Standby Safeguards Analysis

i Analyses in this area cover transients which are more severe out very infreouent and may lead to a breacn of fission product barriers. This j series of transients contained tne one event where the isolated loop operation results in a more severe transient than for normal operation.

] For the reactor coolant pump locked rotor, the maximum pressure was I 2725 psig witn an isolated loop compared with 2675 for normal operation. l i l 1

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i Tnis value is still below tne SRP 15.3.3 acceptance criteria of 110% of

design pressure. -DNS was assumed to occur for Doth modes of operation,

{- -witn'the resulting clad oxidation and peak clad temperature less for tne isolated loop mode.

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• i Three of tne transients whicn were analyzed for normal operation were l not analyzed for an isolated loop, but should be acceptable. The steam -

' generator tube rupture was analyzed only for radiological consequences in the normal mode. Radiological consequences of the tube rupture with one 4

loop isolated snould be less severe. Also, inadvertent loading of fuel in an improper position was not analyzed with an isolated loop, but due to the higner initial DNBR, the results snould still be acceptable. Tne same 4

reasoning would also apply to the single RCCA withdrawal at power.

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Tne remaining initiating events gave results whicn were less severe l for the-isolated loop mode than for the normal mode. There is, nowever one

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area of concern nere. Tne steam line break transients with an isolated loop are generally similar to or slower than the corresponding transients ,

i for normal operation. However, as snown in Table 2.5-2 of License Amendment Reouest No. 35, tne time for 20,000 PPM boron to reacn the core is.significantly shorter with an isolated loop. This becomes even more significant wnen it is noted tnat the time given for normal operation is f tne time to reacn the loops, whereas for isolated loop operation it is the

! time to reacn the core. Tnere is notning in the discussion of the transient whicn would explain this difference, unless it might be that

different codes were used to analyze the two modes of operation.

! 2.3 Loss-of-Coolant Accident The Westinghouse approach to analysis of isolated loop operation for large and intermediate loss-of-conlant accidents has been reviewed and approved on a generic basis.4 Tne approval was conditional on two items, o

the first Deing that at least two nodes are used for the postulated -

inactive loop break. Tne Beaver Valley analysis meets this criteria. Tne f

second item was tnat momentum flux is accounted for between the two cold leg nodes. It is not apparent from the information provided for Beaver S

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Valley that this has been done. Reference is made to a Westinghouse 6

letter which discussed the effect of using one or two nodes to model the l inactive loop break. Later personal communication between NRC and Westinghouse. confirmed that momentum flux was account (d for in the four

Inop model, resulting in a longer end-of-blowdown time for the two node j ,

model. However, the E08 time for the'three loop model is the same for both

! one node and two nodes, indicating tnat momentum flux may not nave been .

taken into account.

A summary of tne results of the Loss-of-Coolant analyses are given in Table 3. .Both the peak clad temperature and the metal water reaction are nigner for tne isolated loop tnan for normal operation, but are still i

within tne acceptance criteria of 10 CFR 50.46. The peak values for both modes of operation occurred for a 0.4 DECLG break. The isolated loop analysis consit ered both an active and an inactive loop break and confirmed tnat tne active loop break is the limiting case.

Neitner the Beaver Valley submittal nor the Westinghouse generic analysis nas considered a small oreak LOCA. Westingnouse has taken the position5that the response of a 4-loop plant operating with one loop isolated is very similar to a 3-loop plant. They further state that "since all N-loop small break analyses show a great deal of margin to the regulatory limits, there is adeouate assurance that small break LOCA analyses for plants operating in the (N 1) loop condition with loop stop valves closed will not be limiting". Tnis would appear to be an adeouate justification, however there is one area of concern. As noted above, the isolated loop analyses result in the hignest peak clad temperatures and metal water reaction. In addition, it appears, from the limited data available, that tnis difference increases for smaller break sizes. If this trend continued into the smaller breaks, then the statement by Westingnouse may not be adequate justification.

2.4 Technical Specification Cnanges G

Several changes in tne reactor protection system are required for isolated loop operation. Tnese are as follows:

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1. Cnange the P-8 interlock setpoints to <71% of rated thermal power.

2. Trip the dT overtemperature and di overpower cnannels for t*,e inactive loop.

3. Reduce tne dT overtemperature trip setpoint.

Tnese actions are to be completed witn the reactor suocritical.

Tnese requirements are listed in tne Tecnnical Specifications, Section 3.4.1.1.

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3.0

SUMMARY

AND RECOMMENDATIONS W Duquesne Light Co. has submitted analyses whicn, in general, support I

their reauest for~ license amendment to allow operation of Beaver Valley

.- Power Station Unit No.1, w'.th one loop isolated. The analyses cover all

classes of transients and were provided for most of the initiating events

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. considered in the original FSAR.

1 f f or tne transients covered under the Core and Coolant Boundary.

Protection Analysis, results from tne normal operation analyses were limiting. In tne area of Standby Safeguards Analysis, the isolated loop results were closer to the normal operation results, and in one case (RCS f

pressure on reactor coolant pump shaft seizure) tne isolated loop value was limiting. For tne Loss-of-Coolant Analysis, the isolated loop results Decame limiting. In all cases, the acceptance criteria of tne corresponding ,ections of tne Standard Review Plan were met.

, During tne course of tne review several areas which may reouire j additional information were identified. These are as follows:

1. Most of tne isolated loop analyses used the LOFTRAN code, in contrast to other codes which were used for the normal

operation. The applicability of this code for this application needs to be verified.

2. Tne applicability of the 15 minute time to loss of shutdown margin for a boron dilution event needs to be established for

) Beaver Valley. If it is determined to be applicable, then Duauesne Light snould be reauested to provide the appropriate i analysis for isolated loop operation.

3. It is recommended that Duauesne Light be reauested to provide an analysis of tne accidental depressurization of the main steam system with an isolated loop.

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4. Duquesne Light should be requested to explain the apparent discrepancy in the injection time for 20,000 PPM boron during tne steam line break.

5. It is felt tnat insufficient information exists to determine tne ,

adeauacy of the plant to a small oreak LOCA. It is recommended that the appropriate analysis be requested from Duquesne Lignt to -

verify tne response of the plant to a small break LOCA.

Subject to satisfactory resolution of the above concerns, it is felt tnt sufficient information nas been presented to allow operation of the Beaver Valley Power Station Unit No. I with an isolated loop.

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4.0 REFERENCES

1. Letter from C. N. Dunn, Vice President, Operations, Duquesne Light, to Mr. A. Schwencer, Chief, Branch No.1, Divisica of Operating Reactors,

. dated October 27. 1978.

2. Final Safety Analysis Report, Beaver Valley Pe..er Station Unit No.1.

3. U. S. Nuclear Regulatory Commission, Standard r.eview Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, NUREG-75/087, dated Marcn 1979.

4. Letter from J. F. Stolz, Cnief, Light Water Reactors Branch No.1, Division of Project Management, to Mr. Thomas M. Anderson, Manager, Nuclear Safety Department, Westinghouse Electric Corporation, dated February 28, 1979.

, 5. Letter from C. Eicheldinger, Manager, Nuclear Safety Department, Westinghouse Electric Corporation, to Mr. J. F. Stolz, Cnief, Light Water Reactors Project, Division of Project Management, dated June 27, 1977.

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TABLE 1. CORE AND COOLANT BOUNDARY PROTECTION ANALYSIS Maximum Pressure Minimum Section Transient (PSIG) DNBR Otner 1.1 Uncontrolled RCCA Bank Withdra.21 6.5NF 0.75TF 820 F (AT)(l)

Subcritical 1.0NF 0.25TF 730 F (AT)(l) 1.2 Uncontrolled RCCA Bank Witndrawal 2350 1.35 1.2NF(l)

At Power 2300 1.86 0.8NF(l) 1.3 RCCA Misalignment 2275 >l.3 Not Analyzed 1.4 Uncontrolled Boron Dilution 57 min /129 min /15 min (2)

Not Analyzed E

1.5 Partial Loss of Forced Reactor 1.6 Coolant Flow 1.98 1.6 Startup of Inactive Reactor 2270 >1. 3 Coolant Loop 6 minute; to loss of shutdown margin 1.7 Loss of Load Turbine Trip 2530 1.6 2510 2.1 1.8 Loss of Normal Feedwater No water I.V. 1150 ft3 (3) relief 880 ft3(3) 1.9 Excessive Heat Removal, Feed- +50 1.65 water System Malfunction I.V. 2.1 1.10 Excessive Load Increase 2240 1.48 1.12 NP(4) 2250 2.2 0.76 NP(4)

TABLE 1. CORE AND COOLANT BOUNDARY PROTECTION ANALYSIS (Continued)

Maximum Pressure Minimum Section Transient (PSIG) DNBR Other 1.11 Loss of Offsite Power Limited by 1.8 1.12 Turbine Generator Accident Not Applicable to tnis Report 1.13 Accidental Depressurization of I.V. >1. 3 -0.5%(5)

Main Steam System I.V. >1.3 -1.0%(5) 1.14 External Environmental Causes Not Applicable to tnis Report 1.15 Accidental Depressurization of I.V. 1./1 Reactor Ceoiant System Not Analyzed 5

(1) NF Neutron Flux (Fraction of Nominal)

TF Thermal Flux (Fraction of Nominal)

AT Average Clad Temperature (2) Time to loss of shutdown margin (Refueling /Startup/At Power)

(3) Pressurizer volume (4) Nuclear Power (Fraction of Nominal)

(5) Reactivity (%oK/K)

insLE 2. STAN33Y S A?EG'JARD5 Aited 515 Maximum Pressure Minimum Section Transient (PSIG) DNBR Other 2.1-2.3 Not Applicable to tnis Report 2.4 Steam Generator Tube Only Radiological Consecuences Analyzed Rupture Not Analyzed 2.5 Major Secondary 1.V. >1.3 Pipe Ruptures I.V. >l.3 0.85%((

0.50% I) l) 2.6 Rod Ejection Accident 2565^F(2) 2038 F(2) 2.7 Reactor Coolant Pump 2675 2031*F(2)

Locked Rotor 2725 <2031 F(2)

2.8 Inadvertent Loading of Accer'able Fuel in Improper Position No'. Analyzed 2.9 Comp,lete Loss of Forced 1.48 Reactor Coolant Flow 2.02 2.10 Single RCCA Withdrawal <5%(3) at Power 2.11 Minor Secondary Not Analyzed4)((4)

Pioe Rupture Not Analyzed (1) Reactivity (%a(/K)

(2) Peak Clad Temperature (3) % Fuel Rods Failed (4) Large secondary pipe rupture meets the acceptance criteria for small break, nence small break was not analyzed.

TABLE 3. LOSS-OF-COOLANT ACCIDENT Break Size Peak Clad Metal / Water

& Location Temperature (*F) Reaction %

1.0 DECLG 1857 2.3 0.6 DECLG 1969 3.5 0.4 DECLG 2014 4.6

• 6" Cold Leg 1729 '.5 4" Cold Leg 1456 0.6 3" Cold Leg 1586 1.E Active Loop 1980 4.5 0.6 DCCLG Active Loop 2155 8.3 0.4 utCLG Inactive Loop 2106 6.4 0.4 DECLG O

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