ML20024D472

Provides Info Requested on 830527 for two-loop Operation. Licensing Process Associated w/N-1 Loop Considerations Has Departed from Technical Issues Unique to N-1 Loop Considerations

ML20024D472
Person / Time
Site:	Beaver Valley
Issue date:	07/29/1983
From:	Carey J DUQUESNE LIGHT CO.
To:	Varga S Office of Nuclear Reactor Regulation
References
TAC-10386, NUDOCS 8308050069
Download: ML20024D472 (107)

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TA@ Telephone (412) 456-6000 Nuclear Division July 29, 1983 P.O. Box 4 Shippingport. PA 15077 0)04 Director of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Attn: Mr. Steven A. Varga, Chief Operating Reactors Branch No.1 Division of Licensing Washington, DC 20555

Reference:

Beaver Valley Power Station, Unit No.1 Cocket No. 50-334, License No. DPR-66 Request for Additional Information on N-1 Loop Operation Gentlemen: In accordance with your letter of flay 27, 1983, we are providing the infomation requested on two-loop operation. We have evaluated the questions submitted by the Reactor Systems Branch (RSB) and the Procedures and Systems Review Branch (PSRB) in your letter. The information is provided in three parts: Enclosure I quantitative basis concerning a Steam Generator Tube Rupture (SGTR) event for N-1 loop operation which was requested by RSB. Enclosure II - responses to the two questions posed by PSRB (Items 1 and 2) Enclosure III-previously committed LOCA reanalysis using the NRC approved 1981 Westinghouse Evaluation flodel for N-l Loop operation. We believe that the licensing process associated with N-1 loop considerations has departed from the technical issues unique to the N-1 loop condition towards resolution of generic multiplant issues related to N loop operation. Specifically, the issue of " qualification of the pressurizer power operated relief valves" and " operator response times" raised for the SGTR (Steam Generatur Tube Rupture) accidents during N-1 loop operations are not technical problems strictly associated with N-1 rather, these concerns are bounded by the N-Loop case for SGTR. We base this on the fact that the identical SGTR procedures utilized for N-1 loop operation would be characteristic of the actions taken by the opera-tors for the N loop condition and are therefore similar in this regard. In addition, the paraneters affecting the initial conditions'and analyzed releases resulting from a SGTR for N-1 loop operation in all cases are comparable to, or more conservative than, the N loop case. One steam generator is adequate for decay heat removal and plant cooldown for either case. Provided as Table I is a comparison of parameters for both the N loop and N-1 loop case and any comments applicable to the conditions. ( I \\ 9308050069 830729 PDR ADOCK 05000334 p PDR

Beaver Valley Power Station, Unit No.1 Docket No. 50-334, License No. DPR-66 Request for Additional Information on N-1 Loop Operation Page 2 s The radiological consequences of a Steam Generator Tube Rupture event presented in the FSAR for N loop operation are based on a calculation of the leakage of primary coolant into the secondary side of the affected steam generator and subsequent discharge of radiological effluent via secondary side relief valves. The rate of reactor coolant leakage is dependent primarily on the capacity of the ECCS system and reactor coolant temperature after reactor trip which are not adversely affected by N-1 loop operation. Similarly, since initial power level and reactor coolant system fluid volume are lower, the discharge of steam required to remove decay heat and sensible fluid energy is also reduced. The accumulated leakage in the affected steam generator is also dependent upon the time required for the operator to cool and depressurize the reactor coolant system to stop primary-to-secondary leakage. These actions can be completed coincident with a loss of offsite power using pressurizer and steam generator power operated relief valves. Neither the availability nor capacity of these components a.e reduced during N-1 loop operation. Hence, the timing of actions by the operator would not be significantly different for N-1 loop operation. Evaluation of the net effect of N-1 operation on the radiological con-sequences of a design basis tube failure leads to the conclusion that the analysis for N loop operation is applicable. The criteria of ANSI N660 " Proposed Standard for Time Response Design Criteria for Safety Related Operator Actions" state in the FOREWORD of the document that the criteria "are not intended to serve as a basis for actual operator action times, procedures, or training". Therefore, we have made no attempt to use this document as suggested to qualify the 30 minute operator response time but have utilized the SNUPPS simulator and three different licensed groups from our plant to make the detemina-tion that all requisite actions for the steam generator tube rupture can be completed within the 30 minute timeframe. Since the timing requirements for N660 were based on simulator analyses, we feel that this is an acceptable alternative. We do not believe that any document or standard can be used to quantify or qualify time response for operators during accident situations and that this can only be done through simulator training and best estimate plant response information recognizing that simulator response will vary j somewhat from actual plant thermal hydraulic response dependent on the complexity of the simulator modeling of SGTR. It was apparent from the three unannounced tests given on the simulator at SNUPPS that the primary requisite actions for SGTR, specifically, identi-I fication and isolation of the faulted generator and the subsequent cooldown and aepressurization can be performed in the 30 minute timeframe. i Maximum Time of 3 Tests 180 seconds identification of faulted steam generator l 240 seconds isolation of faulted steam generator 780 seconds cooldown and depressurization 20 minutes 1,200 seconds = t

Beaver Valley Power Station, Unit No.1 Docket No. 50-334, License No. DPR-66 Request for Additional Information on fl-1 Loop Operation

• Page 3 In conclusion, we believe that these tests were more conclusive with respect to operator response time than comparing existing procedures to ANSI N660 criteria since no document can adequately substitute for sound design, training, human factors considerations and live time siaulation trials in evaluating operator response to accident conditions.

On this basis, we request that the subject of operator response time for SGTR and the qualification of the Pressurizer PORVs in mitigating the consequences of a SGTR for the N-1 loop condition be evaluated as a technical issue for N loop operation to expedite resolution of N-1, unless a specific parameter (s) unique to this configuration is identified as being less con-servative than the N loop case. We have forwarded copies of the SGTR pro-cedure to Mr. P. Tam for your use. Notwithstanding this request and in concert with our desire to resolve the N-1 licensing ussues, we have evaluated the operation of the Pressurizer PORVs in the 3 loop SGTR Case for the conditions under which it is expected to operate (i.e., SGTR with Loss of Offsite Power) and included this infor-mation in Enclosure 1. If you have any questions on this subject, please contact my office. Very truly yours, J. J. Carey Vice President, Nuclear Attachments cc: Mr. W. M. Troskoski, Resident Inspector U. S. Nuclear Regulatory Commission Beaver Valley Power Station Shippingport, PA 15077 U. S. Nuclear Regulatory Commission c/o Document Management Branch l Washington, DC 20555 l I

ENCLOSURE I RSB Request for Additional Information Question: Your response to question 3 in reference 1 concerning a Steam Generator Tube Rupture (SGRT) during N-1 loop operation has been found to be inadequate. No quantitative basis was provided to sub-stantiate your assertion that a SGTR while in the N-1 loop mode would be bounded by the FSAR calculation of a SGTR while in the N loop mode. Further, you have not provided suitable justification that the 30-minute (time to equalize RCS and faulted SG pressures) assumption in the FSAR can be met while in the N-1 loop mode. If suitable quantified bases cannot be provided, you should recalculate a SGTR in the N-1 loop mode, including an analysis of offsite dose consequences. Your analysis should specifically address the following: Assumption of loss of offsite power per GDC-17 Justification for relying on non-safety related equipment for mitigation of the event (e.g., primary PORV, ADV's) should be provided. The timing of actions taken by the operator should be justified on the basis of current or proposed procedures. The time response criteria of ANSI N660 (reference 2) should be taken into account. Your response should contain calculated time variations of upper plenum pressure and temperature, saturation temperature, pressurizer level, level in the faulted steam generator, secondary relief and safety valve flows, and secoadary temperature and pressure for each steam generator. Please also provide a chronological listing of automatic actuations and operator actions, justified on the basis of current or proposed emergency procedures. If the 30-minute criterion assumed in the FSAR cannot be met, please provide justification for the current FSAR, N-loop operation SGTR analysis assumption.

Response

i With respect to the issue of SGTR coincident with loss of offsite power, we cannot identify any parameter specific to N-1 that would rep-resent a less conservative condition that the N loop case and therefore are evaluating it as an N Loop problem. Attachments 1 through 5 are our justifications for relying on non-safety related equipment for miti-gation of a SGTR event coincident with loss of offsite power. The containment environmental conditions under which the pressurizer PORVs are expected to operate during a SGTR coincident with a loss of offsite power for the 3 loop case envelope the expected N-1 condition. i These results and assumptions are presented as Attachment 1. Presented as Attachment 2 is our risk assessment of the probability of a steam generator tube rupture event coincident with loss of offsite power for N loop operation. Attachment 3 provides a comparison between I .L

m_ m __ ENCLOSURE 1 Page 2 ~ the factors affecting SGTR for the N and N-1 loop. Attachment 4 details the conservatisms relative to the Unit 1 Technical Specifications and FSAR assumptions for SGTR event. Attachment 5 is a lcgic diagram which represents the various and diverse means of depressurizing the reactor coolant system during a SGTR. In considera-tion of the low probability of this event, analyses conservatism with respect to 10CFR100 releases, the expected containment atmosphere during the event, substantial capital already expended and the diverse depressurization means available, we feel that it does not warrant backfitting the pressurizer PORVs with fully qualified safety grade equipment. We recognize that this accident was a design basis consideration for the SGTR, and, on a generic basis, the Final Safety Analysis Report does not provide substantial detail on which to justify the 30 minute operator response time. The Westinghouse Owner's Group Procedures Sub-committee will be addressing the 30 minute time response through the l validation and verification efforts on the Em'ergency Response Guidelines to qualify this time frame. In addition, we will evaluate the final version of our plant specific emergency procedure for SGTR when our simulator is completed. This schedule will be consistent with the Control Room Design Review effort submitted in our Generic Letter 82-33 dated April 15, 1983. We have had three of our licensed groups timed on the SNUPPS simulator utilizing our current procedures to qualify the tiue to identify and isolate faulted steam generator and perform the subsequent depressurization/cooldown. These results were addressed in the cover letter and must consider the fact that our operators are not familiar with this control board (SNUPPS), as this is Duquesne Light's first use of this facility and that the modeling of the actual thennal hydraulic response will vary dependent on the simulator. Copies of the procedures for SGTR and safety injection have been forwarded to Mr. P. Tam for your i use. i; The September 1983 meeting of the Westinghouse Owner's Group (WOG)

i has an agenda item on the SGTR 30 minute response time issue that will be i

voted on to address this problem on behalf of the NT0L's (Seabrook, Shoreham, Harris, Catawba), who have open items in their safety evaluations in t this regard. This issue was also identified as an open item (#16) in the 1 l Safety Evaluation of " Emergency Response Guidelines" in the June 1,1983 D. G. Eisenhut letter to J. J. Sheppard of the WOG. The proposed WOG j agenda items relative to SGTR include: justification of operator response time, consequences of delays (i.e., steam generator overfill), qualification of equipment used in mitigation of i j accidents, and limiting single failures. l We expect that this issue will be adequately funded, and we believe the results will be comparable to our preliminary analyses of the SGTR event i i and justify the limited conditions under which the pressurizer PORVs must lj operate. This effort will be responsive to the concerns expressed in the April 4,1983 D. G. Eisenhut memorandum to the Commission entitled " Board Notification Regarding the Need for Rapid Primary System Depres-surization Capability" in PWRs (BN-83-47) and the R. J. Mattson memorandum i l-

ENCLOSURE 1 Page 3 to D. G. Eisenhut dated March 27,1983 entitled " Board Notification Regard-i ns, PORV's". In the event that funding is not approved for this issue,. we intend to stand on the existin5 design of the Pressurizer PORVs since our pre-liminary review indicates that these valves should function under the limited challenge imposed on them by the operating environment in contain-ment during a SGTR. Duquesne Light Company has spent $288,750 for testing the Masonellan-type power operated relief valves through the EPRI Test Program to satisfy the NUREG-0737, Item II.D.l.A requirement. Results of the testing conducted at the Marshall Steam Station, where eleven different evaluation tests were perfomed with a total of 63 cycle operations, showed that in all cases the valves opened and closed on demand with no failures or damage to the valve and that the lowest recorded closing pressure was 2205 psig. To satisfy the concerns of cold overpressure, we have upgraded and perfomed numerous modifications involving: NUREG-0578 Pressurizer Safety and Relief Valve Position Indication Acoustic Monitoring Modifications, Subcooled Margin Meter, the RCS Vent Modifications, and the Reactor Vessel Level Instrumentation Modification Installation of pressure switches in each pressurizer safety relief valve for monitoring pilot assembly leakage Pressurizer Spray Valves Replacement Modification (scheduled for the fourth refueling outage) Upgrading the Pressur'izer Safety Relief Valves RCS Overpressure Protection Modification Total expenditure of the above listed modifications, to date, is $7.3 million. { We have diligently followed and analyzed industry experience on these valves and are currently modifying the valves' air system based on our review of the Westinghouse Tech. Bulletin (NSD-TB-82-02) on the potential for vent port restriction due to orifices or elbows in the air line, which was dated April 15, 1982. l With due consideration for the substantial testing, modifications, man-rem exposure and capital investments associated with these valves i to satisfy multi-plant issues, Duquesne Light Company does not intend 1 to further enhance the qualifications and acceptability of these valves based on their capability to perform their intended function under the expected limited service condition for SGTR. We have not identified any failures of these valves in our review of events (1) that were service induced or caused by adverse environmental conditions. I

ENCLOSURE 1 Page 4 The NRC staff review of NUREG-0651 concluded that during the SGTR cases, "no significant offsite doses or system performance inadequacies have occurred..., and only minor procedural and equipment deficiencies were noted". Moreover, no loss of offsite power occurred prior to or following a SGTR event. References 1. NUREG 0651 Evaluation of Steam Generator Tube Rupture Events Tube Rupture Events NUREG 0886 Steam Generator Tube Experience NUREG 0909 NRC Report on the January 25, 1982 Steam Generator Tube Rupture at R.E. Genna Nuclear Power Plant Kemeany Report on the Accident at Three Mile Island NUREG CR-3226 Station Blackout Accident Analyses / 9 4 l t .lt i ? l ll o' l l t

ATTACHMENT 1 We have evaluated the conditions under which the pressurizer PORVs must operate for a steam generator tube rupture coincident with a loss of offsite power. Our calculaticns indicate that the pressurizer discharge volume recuired for depressurization to the faulted steam generator pressure is relatively small in terms of total energy release and duration (<5 minutes) for the N loop condition. We have perfonned a limited parametric review and iterative mass and energy calculations over incremental tire periods and determined that the containment atmosphere is not significantly affected to the point that PORV operability is challenged during the period of PORY operation, as the energy is substantially absorbed by the pressurizer relief tank surge volume. The assumptions and initial conditions utilized in performing the mass energy releases for SGTR are: 1. The pressurizer relief tank pressure, level and temper-ature are all at their alarm settings prior to the release (78% level, 22.7 psia,125F) 2. PORV capacity 210,000 lbm/hr 0 2250 psia C = 46 critical flow y coef. 0.9 f==.02448ftf/lbmhf=642.3 BTU /lbm 3. Pressurizer liquid V h = .2269 ft /lbm 1153.7 BTU /lbm Pressurizer steam Vg based on a break flow stabilization pressure of 1750 psia t 4. RCS Temperature reduced to 497F prior to depressurization in accordance with SGTR procedures. 5. Charging Pump Mass input (See figure 14.2-3 Updated FSAR attached) 3 6. 1400 ft pressurizer f 1300 ft pressurizer relief tank l 7. Initial PRT Internal Energy = 93.225 Yf =.0162 ft /lbm 3 BTU l lbm I 8. Pressurizer level at time of PORV actuation = 50% 9. Curves for the pressurizer and relief tank are attached l } Since the three PORVs fail closed and their respective isolation valves are being qualified under the EQ Program, we believe the valves are capable of performing their intended and limited safety function over the time frame of interest for a SGTR during the loss of offsite ji power scenario. .[ _. -,, _ _, _ ~,.,.,,.

ATTACHMENT 1 Page 2 Our preliminary analyses indicate that the PRT would not rupture during a SGTR, however, if the PRT did rupture, the containment temper-ature expected for this short tem energy release is approximately 230*F which would decay quickly due to cessation of the release and installed containment cooling systems. If the pressurizer relief tank surge volume cooling capacity, structural heat sinks, time delay associated with attaining equilibrium containment temperatures and time delay until any hypothetical pressurizer relief tank rupture cccurs, are considered for the duration that the valve must function under of offsite power, it is highly improbable (i.e,410 gGTR with a loss ) that at least one of the three installed PORVs would not function under these conditions. References 1. BVPS Updated FSAR 2. Wark, K "Themodynamics" McGraw/ Hill New York 1927 3. Keenan, J. H. and Keyes, F.G. " Steam Tables" John Wiley and Sons, Inc. USA,1969 4. BVPS OM Chapter 53, Procedure E-3 5. Westinghouse E-Spec 676270 6. Masoneilan Handbook for Control Valve S1 zing Masoneilan Inc.1977. 7. " Thermodynamics" Abbot and Van Ness, Schaums Outline,1972

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ATTACHMENT 2 Risk Assessment Evaluation for SGTR Coincident with Loss of Offsite Power The following is a risk assessment addressing SGTR concerns using typical probability values. 1. An abbreviated method following NUREG/CR-2934 (SANDIA), Review and Evaluation of the Indian Point Probabilistic Safety Study (IPPSS),se evaluation)quence to core melt for Tube Rupture (preliminary is used for this case. This event addresses SGTR coincident with a stuck open secondary safety value. In the event of core meltdown, this may result in a direct radioactive material release to the atmosphere. The SGTR frequency used is i from EPRI/NP-2330, dated January,1982. No reduction in Tube Rupture probability (AVT) for secondary chemistry.for BV-1 was attempt volatile treatment The four incidents reported in NUREGs-0651 and 0909 were for plants that initially used phosphate secondary chemistry. The staff in-dicated in NUREG-0651 that AVT has somewhat alleviated the concern over the recurrence of a Point Beach Unit-1 type tube rupture incident. A) S/G TR x Failure of HPI x Secondary Safeties x At least System (ASP *) demanded to open. one safety (AssumesPORVs fails to closed by Procedure) close. 4.0 E-2 x 1.3 E-3 x 1.0 x.01 5.2 E-7 = B) S/G TR x Secondary Safeties x At least one x Failure of demanded to open. safety fails RHR Pumps to close. (ASP *) 4.0 E-2 x 1.0 x.01 x 1.2 E-3 4.8 E-7 = C) A+B= 1.0 E-6

• Value from Accident Sequence Precursor Study, NUREG-2497 (0RNL) 2.

Loss of Offsite Power (LOP) and S/G Tube Rupture j The Beaver Valley Unit-1 FSAR Accident Analysis documents this event. The three cases discussed below provides an assessment j of the probability magnitude. Reasons why the occurrence of this event is considered highly unlikely are also presented. A) LOP and SGTR-coincident occurrence by unrelated causes. The coincident LOP and SGTR probability is considered negligible when the events are not causally related. I A2-1 y 5mi. -m.-i-- m w - -

ATTACHMENT 2 PAGE 2 From WASH-1400, addressing LOP and LOCA, "Since the time of interest for this event is of the order of one minute, the likelihood of losing offsite power by a failure which is not causally related to the LOCA is negligible." It is assumed that the time span of interest for a SGTR would be comparable to that for a LOCA. B) LOP during SGTR event by unrelated causes. In this case, it is assumed that a SGTR event has occurred and at same time after the event there is a LOP due to an unrelated cause. In NUREG/CR-2497 (ORNL) the frequency of LOP was evaluated to be 4.1 E-2/yr. (4.8 x 10 /hr) in which was included { the chance of rectifying the initiating event. A value of 1.8 E-3/ demand (NUREG CR-2497) was used for failure of emergency power. Therefore, the probability for total loss of A/C power is calculated to be 7.4 E-5/yr. If it is assumed that LOP occurs within 30 minutes after tube rupture, the time to equalization of primary and secondary pressures, the probability can be estimated to be on the order of i magnitude of E-9. For comparison, WASH-1400 assessed a point estimate of the failure rate of offsite i power to be 2 E-5 failures per hour and the probability that both diesel generators Unit I will trip out to be 1 i 10 E-2. Although the methodology is somewhat different, j it is noted that the WASH-1400 probability of total loss of electric power after a LOCA for 1 hour is 2.0 E-7. C) LOP at S/G TR-Causally related For this event consideration is given to the possibility { of a SGTR resulting in a turbine generator trip and sub-sequent transient instability of the transmission grid i due to the loss of generation. WASH-1400 assessed that the probability of losing offsite power due to LOCA induced power system transient is 10 E-3 and the failure of 2 diesel generator sets is E-2. These values are used below. Of concern here is the possible reduction of available depressurization capability. Upon LOP, based l on Beaver Valley Unit-1 Operating Procedures, the pres-surizer PORVs would be used to reduce primary system pressure thereby equalizing the pressures between the primary system and the secondary side of the affected steam generator. This action would serve to attenuate break flow into the secondary release path. The pro-bability of PORY failure to operate upon demand is E-3 (upperbound, WASH-1400). The probability for this event can be calculated by: { A2-2 i f I

.= - -.. -...--.:..= ATTACHMENT 2 ~ PAGE 3 LOP x PZR PORY Failure I E-3 x E-3 = E-6 assuming that the probability that the PZR PORV demand is one and emergency power is available. The total loss of AC power can be calculated by: LOP x Failure of Emergency Power E-3 x E-2 E-5 = These probabilities are considered conservatively high estimates for the following reasons: 1) It is somewhat doubtful that LOP would occur at i all since none of the Tube Rupture Events reported in NUREGs 0651 and 0909 indicated a LOP. 2) From NUREG-0886 (2/82) "The probability of the 4 design basis accident occurring during normal operation is small, and the probability that the accident would occur during the short period ! 4 of time between the detection of a leak and that i exceeding the Technical Specification leak rate limit and plant shutdown is even smaller." 3) No credit is taken for corrective action. Restora-tion of offsite power would increase the probability i of full depressurization capability. Data from Appendix III of WASH-1400 approximates that a 64% l restoration of offsite power within 30 minutes of i an event such as a LOCA. On July 28, 1978, a total loss of offsite power occurred at BV-1 due to i j a main transformer fault and improper relay operation. Offsite power was restored in 17 minutes. 4) Only one of three PORVs is considered operable. t 1 5) Beaver Valley Unit-1 has operated with AVT secondary l chemistry. A2-3 1

.u _. 7-B=ver Vallcy Power Station, Unit N3.1 Dockst N3. 50-334, Licensa N3. DPR-66 Requ2st for Additional Infcrmatien Pace 3 ATTACHMENT 3 PARApe:TER cot @ARISDN OF M AND N-1 CASE (FIUP VALVES CIDSED) tL9eENT I Pressurizar PORY Flow Rate 210,000 lbs/hr at 2350 psig saturated steam No affect - 3 fvDV's installed l RCS Pressure 2235 psig initial No affect Nuclear Power 1004 plus calorimetric N loop h kes M-1 SGTR more conservative f 604 plus calorimetric N-1 loop l l Decay Heat tower for N-1 loop due to lower initial power using hkas N-1 SGrk more conservative j ANS 5.1 recay mat RCS Temperature lower 7,,9 for N-1 Kakes N-1 SGrR more conservative RCS Mass Inventory IAss for N-1 loop due to isolated loop inventory hkes N-1 SGTR more conservative since less mass has to be cooled during post accident Coolant Activity Limits rascricted by Technical Specification 3.4.8 Short term increased 1 131 activity for N-1 permitted but compensated b, lower decay heat }

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Fission Product Inventory Imse for N-1 loop due to lower equilibriusa levels hkes N-1 SGru more consetsative ' l of fission products Steam Generator Tube Size .875" No affect Steam Pressure Higher for N-1 due to lower power hkes initial conditions for N-1 SGrR more conversative due to lower tube A P, less break flow J l initial l r I } Steam Generator Level 444 plus instrument errors No affect as post trip staan pressures (would be the same, ie. same mass ) Break Flow Slightly less for N-1 initially due to higher hkes N-1 SGTR more conservative steam pressure during N-1, lower tube differential pressure ECCS Flow Rates Same for N and N-1 loop case No affect cooldown j Faster for N-1 due to less mass in primary to be cooled i and less decay heat hkes N-1 SGTR more f' conversative Site hteorology Same for N and N-1 loop case i; No affect Offsite Power Availability Same for N and N-1 loop case No affect 3 ; Secondary Heat Removal Steam Dump, ICRV's, and Safety Valves avaliable for both No affect, but due to less decay Cases heat the offsite release would be less which would make N-1 moru conversative

~ ATTACHMENT 4 Conservative FSAR Assumptions of SGTR Event 1. Analysis of tube thinning at Prairie Island established that even a tube with a 65% wall reduction did not rupture, while tube plugging is required at BVPS 1 for any tube with a wall reduction of 40%. (NUREG 0651, pg. C-6 and BVPS Technical Specification 3.4.5) 2. BVPS SGTR Analysis assumes 15X15 fuel while a 17X17 is actually used. This is conservative since the diffusion of radioactive isotopes in the fuel is temperature dependent and the 17X17 fuel operates at a lower temperature. Therefore, the release of fission products from the pellet to the clad gap is reduced. (FSAR Section 14.2.4) 3. Radiation monitors have been installed in the main steam relief system in conjunction with NUREG-0737, Item II.F.1 which would expedite operator identification of the faulted steam generator I without benefit from the Air Ejector and Blowdown Radiation Monitor as described in the FSAR. l 4. A qualified pressurizer and vessel head vent system has been installed to meet the requirements of NUREG-0737 Item II.B.l. This system would provide a limited depressurization rate due to a 7/32" orifice being installed in-line but would be available on a loss of all A.C. pending NRC review of the procedures previously submitted. 5. The radiological consequences of a SGTR event presented in the FSAR Section 14.2.4 produces a dose at the site boundary of 300 mrem whole body and 900 mrem thyroid which is substantially within the limits of 10CFR100, even if it is assumed the operator delays in taking action when warned by alarms and instruments. Therefore, as long as a single phase steam release is maintained, the accumulated release would be conservatively within 10 CFR 100 limits even if we assume the primary depressurization and isolation times already established during actual SGTR events and documented in NUREG 0909. Equalization times for actual events obtained from NUREG-0909 are summarized below: Point Beach 108 minutes S u r ry 2............... 30 minutes Ginna 180 minutes t Prairie Island............ 61 minutes 6. The activity release through a faulted steam generator, which i is limited by the concentration in the reactor coolant assumed l to result from 1% failed fuel, is highly conservative based on the fact the worst case coolant activity measured throughout l I r w

ATTAC& TENT 4 Page 2 s reactor operations at BVPS to date, was approximately 1 uCi/cc. 7. Conservative Meteorology data was also utilized fn the activity release calculations. A X/Q value of 7.8 x 10-4 sec/m was 3 used in the FSAR, whegeas thg actual annual average X/Q value for BVPS is 7.1 x 10' sec/m. This would substantially reduce the offsite release rate below the projected FSAR assumed release. 8. In the Model 51 steam generators utilized at Beaver Valley, approx-3 imately 2638 ft of volume is available above the tap of the steam generator level span to the main steam isolation valve in the short-3 est run (438 ft ) of steam pipe in the 18 Steam Generator. 1 The break flow rate through a double ended steam generator tube at a pressure of 2250 psia, which conservatively bounds the possible mass addition to the faulted steam generator is approximately 35 lbm/sec. assuming a nominal density of b eak f1 w f 50 lbm/ft, 5 I the volumetric flow rate is about 1.5 ft /sec which indicates that I the operator has more than 20 minutes to fill a faulted steam generator to the MSIV after the indicated level has gone off-scale high. l If a more realistic break flow rate were used at the equilibrium pressure where SI flow matches break flow and consideration given to the " shrink" in the steam generator level post trip, substantial i time beyond 20 minutes could be realized. The Unit 1 FSAR analyses for SGTR was based on a mass transfer of 132,000 lbm to the secondary I which is conservative with respect to the actual flows that would be anticipated for the expected duration. { l

ATTACFJiENT 5 s Legend / Notes: "or" gate, requires any input signal to produce an output ums "and" gate, requires all input signals to produce an output - i sources of offsite power required for component operation For systems inside containment, air is supplied by 1 of 2 air compressors [IA-C-1A, 1B] which are powered from emergency power sources. Cooling to these units is normally supplied by a non-IE source (chilled water) but has a lE backup source (river water). For systems outside containment, air is supplied by 1 of 3 non-IE power air compressors (SA-C-1A, 1B, IC) which are operable if one offsite power source is available to the 480 volt busses. A diesel-driven backup compressor is also installed and verified operable on a weekly basis. A permanent modification to power 1 of 3 motor-driven air compressors and cooling supplies from a diesel generator power source is being designed for projected completion during the 5th refueling outage. The air system inside containment can be cross-connected with the outside air system consistent with containment integrity technical specifications. References f Prawings RE21DN, RM155B, OM 36, OM 37, OM 38, OM 39 Manuals OM 6, 7, 29, 34, 36, 37, 38, 39 Sections 1-5 j OM 34 Procedures J, L j DCP 295 File IE/ Circular 80-15 File s NOG Letter 82-155, 83-200 Westinghouse letter NS-PL-11697 6/30/83 NUREG 0737 Item II.B.1 File j EDS Report NUREG 0737 i MED 00177 5/13/82 Schneider Summary Report EQ Status 6/24/82 l

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ENCLOSURE II i PSRB Request for Additional Information Item 1 The BVPS-1 Startup Report (Initial Startup Cycle) indicates that N-1 flow coastdown measurements were perfonned during Initial Startup. Confinn that the data is still valid (i.e., no changes to RCS piping, reactor internals, fuel design, S/G's or RCP's have occurred which would affect the original test results in a less conservative direction).

Response

An evaluation was conducted to review changes to the Reactor Coolant System and internals. The following is a summarized listing of minor changes made or scheduled since initial startup and the affects on flow coastdown: Change Affect Reactor Internals: Insignificant Guide tube spit pin replace-ment modifications (Westinghouse design) Fuel Design: Insignificant 1. Two optimized fuel assemblies 2. Pre-pressurization of fuel reduced from 500 to 450 psig Steam Generator Tubes: Insignificant cycle (Cycle 3) gged during last One tube was plu RCS Hot and Cold Leg Piping: No affect l-No changes l l ' Attached is a copy of the Test Results Report for the initial flow l coastdown test (BVT 1.1-7.6.1) performed in 1976. The report identifies the case for two loop operation with a loop isolated. The test demon-strated measured core flows in excess of their corresponding FSAR curves (Reference Figure 3.8-6) for the N-1 loop case. 1 1: Based on our evaluation, it can be concluded there have been no signifi-cant changes that would affect the original test results of the N-1 flow coastdown measurements performed during the initial startup in a less conservative direction. i l ~ '~ l '.~

Attachment DUQUESNE LIGHT COMPANY Pag 2 1 Besvar Vallay Powsr Station - Unit 1 i Test Results Report Date: 10/14/76 7 Reactor Coolant Flow Coastdown BVT 1.1-7.6.1 Title 0/40'I Issue 5/2/76 Revision 7/2/13 JTG Approval Date Test Date: Start 5/5/76 End 5/12/76 Partial Test No Complete Test Yes Test Results: SatisfactoryVII A,B,C,D,E,F. Unsatisfactory None Retesting Recommended No. Unreviewed Safety Question . Attachments: FIGS. 3.6-1 THIU Involved / Evaluated No / NA 3.8-6, FIG. 2A Purpose-Scope: To measure the rate at which reactor coolant flow changes following various j reactor coolant loop loss of flow incidents, and to measure the protective system l time delays associated with a loss of flow incident to determine the values . assumed in the accident analysis are conservative. Test Summary (Conclusion): All values measured / calculated for this test satisfied the applicable acceptance-criteria as specified in the procedure (shown on page 4 and Fi'g's. 3.8-1 thru 6). Individual loop flow data for this test was obtained from visicorder traces of one loop flow transmitter output voltage for each loop. These traces were reduced in accordance with the V startup procedure, DLW-SU-5.1.8, flow coastdown measurements. As backup' data, RCS pressure, T ve and flow in each a loop were recorded (three indications per loop) before and after each test run. All cases were analyzed for the first ten seconds of the flow coastdown transient and compared to the W revised FSAR curves and calculated minimum DNBR points for each case, which were also furnished. Appropriate allowances were made for the flow sensor time delay (0.608 seconds) which was extracted from'the data (Fig. 2A). s t l Recommendation (if any):

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.A.l. ^ ~~ Attachynt DUQUESNE LIGHT COMPANY Page' Beaver Valley Power Statiion . m., Test Results Report Continuation Sheet Test Summary (Conclusion): (Continued) Initial attempts to compare the flow coastdown data obtained during this test with the curves furnished in the BVPS FSAR section 14 were not successful

}

due to the difficulty in extracting exact points from the FSAR curves for data comparison. Upon request, W furnished coordinate values used to plot the curves shown in the FSAR. The initial plotting of the test data and the data furnished ~by H revealed numerous instances where the test data fell below the values for the FSAR curves by observable amounts. W personnel visited the site to review the raw test data reduction which resolved some of the deviations; however, in most cases', the test data' indicated slightly i quicker flow coastdown than the curves furnished in the FSAR for the This problem.was discussed initial few seconds of the transient. with H personnel and it was agreed by all parties that the method used to correct for sensor time delay,'although performed in accordance with W guidelines for data reduction, made comparison of the flow values determined during the onset of the transient to the FSAR values appear pe'ssimistically This did not resolve the reason for the lower measured values during low. After further discussion E indicated the later portions of the transients. they would reanalyze the particular loss of flow transients with some changes to the input parameters of the code to more closely address the ' l d at BVPS. , dynamic characteristics.of the reactor coolant pumps instal e T This new analysis was performed and new curves were furnished by W for use in the analysis of the results of this test and for subsequent revision of the curves contained in the BVPS FSAR. E also furnished the coordinate values used to generate these curves to allow more precise comparison with references to In the following sections of this report, the test data. the FSAR curves and values are based on the reanalysis of these transients While the rather than the analysis performed initially for the FSAR. quoted for the time interval from the initiation of the exact numbers event until minimum DNBR is reached, as well as the specific values for the DNBR associated with the various transients have changed, the changes are i relatively small in value and the original conclusions stated in the FSAR concerning these accidents are not affected by the results of the reanalysis. The. case of three reactor coolant loops operating with one reactor coolant pump coasting down (RC-P-1A) revealed an absence of the loop C input flow signal (From TP-436-1 via FT-RC-436) and failed to trace on the visicorder i However, as flow was verified in loop C by backup indication in Since flow in loop C was 2 100% pap er. the control room, this run was not repeated. in subsequent cases under similar conditions, the conservative assumption of 100% loop C flow was used to arrive at a core flow figure for this case. With the above assumption, the measured value for core flow was found With the acceptable in comparison to its FSAR curve (refer to Fig. 3.8-1). full length ' control rod withdrawn to 228 steps (K-14 of CB-A), and slowest RC-F-1A was tripped and the all three reactor coolant pumps, operating, following monitored variable successfully met their acceptance criteria: sJ s BVT 1.1-7.6.1 i i L_ ~. r-- -me.. w=-,e,--- wr r- ,,-g- - - - w--,---m,,r-- ,47-e,,.- -r--~ g w, -,-w m- -+-r-r---e---, w-v y ~- -w,,,

l Attachment DUQUESNE LIGHT C0FTA20f Page' 3 Beaver Valley Power Station 3 Test Results Report Continuation Sheet Test Summary (Conclusion): (Continued) Low flow trip time delay (1.70 second), undervoltage trip time delay (1.17 second) and pump underfrequency trip time delay (0.53 second refer to data sheet on page 4 ). The case of three reactor coolant loops operating with three reactor coolant pumps coasting down, proved that the measured minimum core flow is above the FSAR values for flow coastdown (Refer to Fig. 3.8-2). The slope of the measured inverse core flow was less than the FSAR curve slope for this l case, as required (Refer to Fig. 2A). 'The test requirement that all three j ' RCPs trip within 100 milliseconds of each other was proven in the above case (two RCPs trip within 100 milliseconds of each other was demonstrated l in subsequent cases that followed), m i y The cases of two out of three loops operating with one or.two reactor coolant age - j pump (s) coasting down - one isolated loop, demonstrated measured core flows in excess of their corresponding FSAR curves (Refer to Figs. 3.8-6 and 3.8-4). The cases of three~ reactor coolant loops operating with one or two reactor l coolant pump (s) coasting down - no isolated loop, also demonstrated measured core flows above the values shown in the corresponding FSAR curves (Refer to i, 3 1 J Figs. 3.8-5 & 3.8-3)., 111.casesexaminedabovedemonstratedthatthemeasuredcoreflows,through 7,gy - the flow coastdown transient (10 seconds), including the time identified as j the point of minimum DNBR, were above the FSAR assumed core flows (Refer l l g,to Figs. 3.8-1 thru 3.8-6). ,p MWR #762226 was issued to trouble shoot TP-436-1 and FT-RC-436.co determine l the cause of the signal failure. Since backup data for loop flow was available and traces taken later indicated sufficient flow in loop C, lack of this signal trace did not renstitute a setback to core flow calculations. Also (FI-RC-414], RCS loop A flow' indicator, failed to zero when all flow had ceased in the core, 3 i as indicated by remaining flow indicators in loops B & C (including second { flow indicator in lo.o'p A). MWR #762227 was issued to request calibration of } the flow indicator. MWR #762226 was closed on 6/8/76 while MWR #762227 is open. I ~ i o ,)' BVT 1.1-7.6.1 e w=w-- y r-m,.- rv -aw* e c-y.e-, .-,--w y e s--

___-l_.____.__.__._..~~. i.i_ _ _ _.. _.. _.. __ __._____[,_. ~ G v J. - l; REACTOR COOLANT SYSTEM FLOW COASTDOWN TEST DATA Low Flow Low Flow Trip Under Volt Under Volt Under Freq Under Freq Inverse Core Inverse Core Trip Time Time Delay Trip Time Trip Time Trip Time Trip Time Flow Slope Flow Slope l; Delay (MEAS) (FSAR) Delay Heasured Delay (FSAR) Delay Heasured Delay (FSAR) Heasured (FSAR) (sec) (sec) (sec) (sec) (sec) (sec) Flow / Flow.sec "*/ Flow.sec o 1.70 s,2.42 1.17 5 1.20 0.53 s 0.60 0.0985 s 0.1043 t' All Above obta.ined From Fig. 3.8-1 Case From Fig. 3.8-2 Case i I H $n (D E 1 2". i M TC a F aL a i

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c,** 0 ~) C3 L .j 1 LOOP AND CORE FLOW CALCULATIONS 1 2 Loops Operating, 1 Loop Coasting Down - 1 Isolated Loop U T PERCENT OF INITIAL FLOW T l Z Time (t) 0 SEC. 2 SEC. 4 SEC. 6 SEC. 8 SEC. 10 SEC. Q 4 a <m Ox Ox 0x Oz Oz o2 1i l O O i O $n" 16 0 % lol x 161 x lb2.x I 02. x 162.2 L i O 2 z o 'EE' 100 z $5 x 73 x 6bz sb~z 47 2 0 R H loo x 93 z 17x 23 x 79 x 75 z E 5 tr e 2 Loops Operating, 2 Loops Coasting Down - 1 Isolated Loop l G e s a j d i ]U m u J7 l PERCENT OF INITIAL FLOW

• \\

n o i 's Time (t) 0 SEC. 2 SEC. 4 SEC. 6 SEC. 8 SEC. 10 SEC. i 0 L P Q Q 6 O Q y, IA z g g E 2.

• K3 H S,s

'?P JD Q f7 x 7g x 68 gQ ,3 3 1 a ? I z x x x nu wn W % N y. H , k_, N ~ u, '$C IOO bl 75 x 00 x OEz EL 1 ~ hG z 1 n -s O jgg 37 75 44 gyx g,3 g g g; M Core x x x 2 2 +. 9 >W 3 u I T y) i .4

? Y (q,

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I oo

) S TABLE 3.2-1 m DNB PARAMETERS. M E LIMITS e E 2 Loops In Opera-2 Loops In Opera-i l: M 3 Loops In tion & Loop Stop tion & Isolated Loop PARAMETER Operation Valves Open Stop Valves Closed

• {

'I ~ Reactor Coolant System T,yg < 581*F. < 570*F < 570*F Pressurizer Pressure > 2220 psia * > 2220 psia * > 2220 psia

• Reactor Coolant System 265,500 gpm

> 189,000 gpm > 187,800 gpm w 2 Total Flow Rate i ) uy ,i

l

• Limit not applicable during either a THERMAL POWER ramp increase in excess of 5% RATED THERMAL POWER

][ per minute or a THERMAL POWER step increase in excess of 10% RATED' THERMAL POWER. il;j i' i I e I 1 .i l I i .!i;

Enclosure II Page 2 Item 2 It is not clear, from the BVPS-1 Startup Report for the Initial Start-up, that the "RTD Bypass Loop Flow Verification" test data and accept-ance criteria are adequate to support N-1 operation in view of the fact that N-1 operation may reduce cold leg RTD bypass flow. Confirm that RTD response times will be acceptable for N-1 loop operation.

Response

Based on a review of the test data and Test Results Report for BVT 1.1-4.6.7, titled "RTD Bypass Loop Flow Verification", which is attached for your information (Attachment 1), the actual flow recorded in the RTD cold legs exceed the minimum required by at least 300% for each loop. This indicates that the flow would have to decrease to less than 1/3 the original value before it reached the minimum required value during N-1 loop operation. These minimum required values were calculated as part of the BVT 1.1.4.6.7 Test Results Report (Attachment 2), and were based on the RCS piping measurements and a one second transport time. The Ap across the RCS piping is proportional to the flow rate. The total RCS flow rates are verified every 18 months per Tech. Spec. 4.2.5.2, Table 3.2-1, which gives minimum flow rate values for 3 loop, 2 loop w/stop valves open and 2 loop w/stop valves closed conditions. The limits for flow rates with stop valves closed and stop valves open for 2 loop operation are within 0.6% of each other. In the most unlikely event of a low flow condition, a low flow alarm set at approximately 250 gpm would annunciate in the control room and would immediately alert operations personnel of the abnormal condition. The DNB related parameters for RCS Tavg, Pressurizer Pressure, and RCS flow rate are presently covered under Technical Specification 3.2.5, which requires the parameters are verified to be within limits every 12 hours for both N loop and N-l loop conditions. ,i I l> !I' I I -o-- + - - ...~.

f r) Attachment i j DUQUESNE LICllT COMPANY Page 1 Beaver Valley Power Station - Unit 1 Test Results Report Date: 5/3/76 Title RTD Bypass Loop Flow Verification Issue 1 Revision Rev.0 FR 3 JTG Approval Date 12/11/74 Test Date: Start 4/1R/7s End 4/25/76 Partial Test Yes Complete Test No 6 Test Results: Satisfactory Yes Unsatisfactory

• No i

Retesting Recommended No Unreviewed Safety Question Attachments: None Involved / Evaluated No / NA Purpose-Scope: The purpose of this section of the test was to measure the actual flow rate in each RTD bypass loop to ensure the transport times are acceptable. The low flow alarm setpoint in the combined RTD bypass loop flow on each RCS loop was verified. Test Summary (Conclusion): I Flow Rate (GPM) Transoort Time (Sec) Actual Min. Rea'd Actual Max. Allowed i Hot Leg 113 95 .84 1.0. lA Lcop Cold Leg 175 49 .28 1.0

• E 1B Loop Cold Leg 162 48

.30 1.0 "8 IC Loop Cold Leg 165 49 .30 1.0 l Recommendation (if any): i l None I Review 0.S.C. Anproval JTG Approva_1 Test Engr. h m d8-S&W Test Supvr. ./ [ M a3 E(OthertbM/r h8[/ /08/2b Sta. Supt. Oi A D.L.Co. _~ ~

s' Attachment i 1 r DUQUESNE LIGHT COMPANY Page 2 fN Beaver Valley Power Station .\\ Test Results Report Continuation Sheet Combined Hot and Cold Leg Actual Total Low Flow Alarm Low Flow Alarm (?.' of total flow) Flow (GPM) (GPM) Actual Acceotance 1A Loop 288 263 91.3 90 1 2 1B Loop 275 248 90.2 9012 IC Loop 275 246 89.5 9012 All hot and cold leg RTD bypass loops met teceptance for transport time. l The low flow alcrms for the combined hot and cold leg RTD bypass loops of all three RCS loops were reset to bring the setpoints within tolerance. The as left values are shown in the above tabulation. 4 l. lC ie l / l l I l BVT 1.1 - 4.6.7 i< ww-ws. - - _ -e-w--,.. ms ... m m

3

... d

• DUQUESNE LIG11T COMPANY Pag 3 1 Beaver Valley Power Station - Unit 1 t

Test Results Report Date: 1-27-76 Title RTD Bvoass Looo Flow Verifinneinn Issue 1 ~ Revision n JTG Approval Date 12_11,7s_ Test Date: Start 4-18-75 End 1-26-76 Partial Test Yes Complete Test No Test Results: Satisfactory Yes Unsatisfactory No Retesting Recommended No Unreviewed Safety Question Attachments: Attachmen* 1 Involved / Evaluated No / NA DLW-SU-5. 9 Purpose-Scope: This is a partial test, Section VII.A. The purpose of the test is to measure the length of the installed piping from reactor coolant loop connections to the last downstream RTD on the RTD manifold for both cold and hot leg RTD bypass loops. This measurement and pipe size is used to determine the flow rate required to obtain a transport time of 1.0 second. Test Su==ary (Conclusion): The length of of the various schedule piping was measured in the hot and cold bypass loops. The calculated flows required to obtain transport times of 1.0;second are listed in attachment 1. i No exact time has been specified as acceptance criteria for the bypass loop coolant transport time.. The time has been increased from 0.5 seconds to 1.0 seconds. (Ref. DLW-SU-5.1.9, 6.0 - acceptance criteria). Due to the locatiori of the piping tap off on the coolant loops, the bypass loop driving heads of the hot leg is considerably less than the cold leg thereby causing difficulty in obtaining available flow to achieve a 0.5 second transport time. The 1.0 second transport time reduces required flow by 50%. Recommendations (if any): ,l None i i ~ l 'i E h Review 0.S.C. Approval JTG Approval bf S&W W .,, /A Test Engr. j h h w }w ( /t t[7,(. N ..o) 8/ A W (Other) Test Supvr. Wdd )[ 7j 4 D.L.Co. Sta. Supt. [A M i ( ~ 7 V /

4 BVT 1.1 - 4.6.7 6*dd 0 X. DATA SHEET NO. 1 DATE: /[27/76 INITIALS: fB4

i. o 18 A.

Hot leg RTD bypass flow rate necessary to achieve a Ord second t yeg transport time. Measured total Measured total, Calculated Calcult 1" pipe length 2" pipe length t otal volume require (ft) (3 paths) (ft) L (2) (ft)3 VEL flow rc Loop RTD L (1) (gpm) FF 1A TRB-4113, 412B, 411D / 2 3 k,, / 3 / , 2f/ ff IB TRB-421B, 422B, 421D /3 7 " /99 .222 /00 I IC TRB-4313, 432B, I 431D /2/, /%3 e2/$ 9 f* b

i. o

\\ y g,g B. Cold leg RID bypass flow rate necessary to achieve a eri second transp. time. Measured total Measured total Calculated Calcult i{ Loop RTD 1.5" pipe length 2" pipe length total vol. require 3 (f t ) VCL flow r: (ft) L(1.5) (ft) L(2) gpm FCI l 1A TRB-412C, 412D,411C (f S f' , /d 9 +'9 1B TRB-422C, 422D, ftf " 53" . /o 7 ff 421C 1C TRB-432C, 432D, 431C f'7 f8,, ,//O I9 iliIlIl\\ I l l I /!'a A3 3 ROVD CO 3Y 1Ssut 1. to/7/7c i 7 7 y-- ^

.. :L.. - t Attachment L i (ceab BVT 1.1 - 4.6.7 e P( XI. APPENDIX 2 A. Calculated Transport Time To calculate the flow rate necessary to achieve a (Hfr second l.7% transport time, utilize the formula F= V x D/T where: F = Flow rate in hot or cold leg bypass loop V = Total volume of pipa (hot or cold leg) from the Reactor l-Coolant System loop pipe connection to the last downstream RTD. D = Volumetric conversion factor T = Transport time To determine the volume of the pipe, multiply the length of each pipe (1", 1 1/2", or 2") by the cross sectional area of each pipe. The i l( cross-sectional areas for the pipes are as follows: Area 1" schedule 160 pipe - 0.00362 s.quare feet Area 1.1/2" schedule 160 pipe - 0.00976 square feet i j Area 2" schedule 160 pipe - 0.01556 square feet s i Let L (1), L (1.5) and L (2) be the lengths of 1",1.5" and 2" pipe respectively in feet. l' For the cold leg bypass loop, there is only 1 1/2" and 2" piping. V Cold Leg = VCL = L (1.5) x (.00976) + L (2) x (.01556) 'l D = 7.48 gallons / cubic foot i l; 09667 T = 4.Guoaa miputes, the required transport time [ i l Then - FCL = VCL x 7.48 gpm 706 ...x-ss 0lG G 1 ,} The calculation of the required flow for the hot leg bypass is done i

(

in the same manner except that there is only 1" and 2" piping in the !U l hot leg bypass loop. " ~ ~ A? MOVEJ CO?Y e m m -m ,p_-,, . ~., .g -g 4..,,.-,, y

f Attachment i. BVT 1.1 - 4.6.7 (C** D V XI. APPENDIX 2 (Continued) B. Actual Transport Time To detamine the actual transport time, transpose the equation: act o.a.t flow rAtc. as calcula.ted A T = V x D/F where F is now the ab ou t.

1. 80 i

-eeee. This will give the actual transport time which can be compaced to the required transport time. t j

c..a c i c E c the o.c. t w t cold Ic3 a.e a het ic3 RTo 6 psn lo a g (low nle >

hy w song the f.,Ilowy 7 ep o_ t... r :. : F+ g F[ i (l+ #'/Fi., ) F c.' = F Fw' i v w h er e. t F coAined hot and told le3 RTO 6f pus /.og //.a r<Ae. l t 6pcn loop //ow r c.+ e F m e a.s u.e e d e o ld le3RTo 7 c b po.s s I.,g flow rat e, f w - me c.s w< e d hot. Ie3 RTo y l F, '

o. c 4. w o.1 cold le3 b po.3 3 loop (Iow c <~ t e,

Rro y e f g' -

o. c E w o.1 Lot le3 RTo bypo.33 loo p (Io -

r te. l Record the s e f iw s . D t o. SAce4

• 2.

an l l i L> i, ISSUE 1, 10/7/74 A)') ROV ] CO?Y i l

- ~ - -... i 1 j ENCLOSURE III N-1 Loop LOCA Reanalysis j Attached is the results of the LOCA reanalysis performed for Beaver Valley Unit I using the 1981 Westinghouse Evaluation Model. I l The large Break N-1 LOCA ECCS analysis was performed at a power i level of 1724 MWt. Other pertinent analysis assumptions include 17X17 standard fuel design. Also, the accumulator water volume remained 1025 cubic feet per accumulator. The analysis was performed with the NRC approved 1981 Westinghouse Evaluation Model as described in WCAP-i 9220-P-A Rev. 1. The results of this analysis are presented in tuo attachments. Attachment A contains the analysis results, including tables and figures. Attachment B contains LOCA related technical specification values. The CD = 0.4, 0.6 active loop break and CD = 0.4 inactive loop break sizes were analyzed in this study. The worst break size is the CD = 0.4 DECLG active loop break, and resulted in a peak clad temperature (PCT) of 1882 F at a total peaking factor (FQT) of 3.03. This analysis demonstrates conformance with the 10CFR50.46 requirements for Large Break ECCS LOCA Analyses. / 1 O .----+we . 9

.,x,, '. v,.-. ] ATTACHMENT A MAJOR REACTOR COOLANT SYSTEM PIPE RUPTURES (LOSS OF COOLANT ACCIDENT) WITH ONE COOLANT LOOP OUT OF SERVICE An analysis specified by 10CFR50.46EI3,"AcceptanceCriteriaforEmergency Core Cooling Systems for Light Water Nuclear Power Reactors", for 2-loop operation of the Beaver Valley Station is presented in this section. The results of the loss of coolant accident analyses are shown in Table 15.4-2 and show compliance with the Acceptance Criteria. The analytical techniques used are in compliance with Appendix K of 10CFR50, and are described in Refer-ence[2]. The boundary considered for loss of coolant accidents as related to connec-ting piping is defined in Section 3.6. Should a major break occur, depressurization of the Reactor Coolant System results in a pressure de. crease in the pressurizer. Reactor trip signal occurs when the pressurizer low pressure trip setpoint is reached. A Safety Injec-tion System signal is actuated when the appropriate setpoint is reached. These countermeasure will limit the consequences of the accident in two ways: 1. Reactor trip and barated water injection complement void formation in causing rapid reduction of power to a residual level corresponding to fission product decay heat. 1 2. Injection of borated water provides heat transfer from the core and pre-l l vents excessive clad temperatures. i At the beginning of the blowdown phase, the entire Reactor Coolant System contains subcooled liquid which transfers heat from the core by forced convec-tion with some fully developed nucleate boiling. After the break develops, the time to departure from nucleate boiling is calculated, consistant with l Appendix K of 10CFR50. Thereafter, the core heat tranfer is bcsed on local conditions with transition boiling and forced convection to steam as the major heat tranfer mechanisms. During the refill period rod-to-rod radiation is the only heat transfer mechanism. 3 I 4702Q:l/071383

TT 7 :' ~-~ ~ ~ ~

1 When the Reactor Coolant System pressure falls below 600 psia the accumulators begin to inject borated water. The conservative assumption is made that accu-mulator water injected bypasses the core and goes out through the break until the termination of bypass. This conservatism is again consistent with Appen-dix K of 10CFR50. 15.4.1.1 Thermal Analysis 15.4.1.1.1 Westinghouse Performance Criteria for Emergency Core Cooling System The reactor is designed to withstand thermal effects caused by a loss of cool-ant accident including the double-ended severance of the largest Reactor Cool-ant System pipe. The reactor core and internals together with the Emergency Core Cooling System (ECCS) are designed so that the reacter can be safely shutdown and the essential heat transfer geometry of the core preserved fol-lowing the accident. Emergency safeguards systems present at the Beaver Valley station will be available during 2-loop operation. l The ECCS, even when operating during the injection mode with the most severe single active failure loss of a low-head SI pump, is designed to meet the AcceptanceCriteriaEl3 / 15.4.1.1.2 Method of Thermal Analysis The description of the various aspects of the loss of coolant accident analy-sis is given in Reference [2]. This document describes the major phenomena modeled, the interfaces among the computer codes and features of the codes which maintain compliance with the Acceptance Criteria. The individual codes are described in detail in References [3] through [6]. The analyses presented l were performed using the 1981 version of the Westinghouse Evaluation Model. This version includes the modifications to the models referenced above as specified by tne Nuclear Regulatory Commission (NRC) in Reference [7] and j complies with Appendix K of 10CFR50. The 1981 Westinghouse Evaluation Model is documented in References [8], [11] and [12]. Containment data used to cal-culate ECCS backpressure is presented in Table 15.4-3. l 4702Q: 1/071383 f _._.__m__. _.. _ _ _ _, - _.,,..,.,,. _ -.,. -,.., m.

fThe methods used to model ECCS performance for (N-1) loop operation are de-scribed in Reference [10]. Two distinct reactor coolant pipe cold leg break locations are possible during operation with a loop out of service; the break may occur either in an active loop or in the inactive loop. The SATAN nodali-zation scheme described in Reference [3] has been expanded in order to include the portion of the reactor coolant pipes in the isolated loop between the reactor vessel and the loop isolation valves in the blowdown calculation. In the active loop break case scheme, element 52 is the inactive loop hot leg pipe length, element 54 is the cold leg pipe length, and element 33 is the inactive loop accumulator which feeds element 54, with the break location unchanged. For an inective loop break as described in Reference [13], ele-ment 52 remains the hot leg, while elements 54 and 55 constitute the vessel side of the broken pipe. Element 56 represents the valved off pipe segment side of the break. i f The WREFLOOD Code 19 element model is presented in Reference [5]. Figure 4.1 remains unchanged for active loop br~eak analyses. To model the inactive loop i break location a 14-element loop model was devised and reported in Refer-ence [10]. 'i I The ECCS calculations were performed based on a core power peaking factor envelope calculated for 2-loop operation of Beaver Valley Unit 1. A design ,j Fg of 2.32 for N-loop operation results in a Fg of 3.03 for N-1 loop operation. The normalized power versus core iieight [K(Z)] curve for N-1 loop operation is presented in Figure 15.4-17. i l Figures 15.4-1 through 15.4-16 present the transients for the principal parameters for the break sizes analyzed. The following items are noted: 4 Figures 15.4-la The following quantities are presented at the clad burst through 15.4-3c location and at the hot spot (location of maximum clad tem-perature) both on the hottest fuel rod (hot rod): (1) fluid quality 1 (2) mass velocity (3) heat transfer coefficient. 4702Q:l/071383 ~ ) z :: ru - ----:.

i The heat transfsr coefficient shown is calculated by the LOCTA-IV code. Figures 15.4-4a The system pressure shown is the calculated pressure in the through 15.4-6c core. The flow rate out the break is plotted as the sum of both ends for the guillotine break cases. This core pres-sure drop shown is from the lower plenum, near the core, to the upper plenum at the core outlet. Figures 15.4-7a These figures show the hot spot clad temperature transient through 15.4-9c and the clad temperature transient at the burst location. l The fluid temperature shown is also for the hot spot and burst location. The core ficw (top and bottom) is also shown. Figures 15.4-10a These figures show the core reflood transient. through 15.4-10f Figures 15.4-11a These figures show the Emergency Core Cooling System flow for through 15.4-12c all cases analyzed. As described earlier, the accumulator delivery during blowdown is discarded until the end of by-pass is calculated. Accumulator flow, however, is estab-11shed in refill reflood calculations. The accumulator flow assumed is the sum of that injected in the intact cold legs. Figures 15.4-13a, b, c The containment pressure transient is also provided. l lt Figures 15.4-14a, b, c These figures show the core power transient. l Figure 15.4-15 This figure shows the break energy released to the contain-ment during blowdown for the limiting case break. l ' i Figure 15.4-16 This figure provides the containment wall condensing heat transfer coefficient for the limiting case break. I l 47020: 1/071383 l- = __ z ::

( i t i . Figure 15.4-17 This figure provides the operating power peaking factor envelope. In addition to the above, Tables 15.4-4 and 15.4-5 present the reflood mass and energy release to the containment and the broken loop accumulator mass and energy flowrate to the contanment, respectively. The clad temperature analysis is based on a total peaking factor of 3.03. The hot spot metal water reaction reached is 3.0 percent, which is well below the embrittlement limit of 17 percent, as required by 10CFR50.46. In addition, the total core metal-water reaction is less than 0.3 percent for all breaks as compared with the 1 percent criterion of 10CFR50.46. ,Coriciusiens - Thermal Analysis L; i For breaks up to and including the doubled-ended severance of a reactor cool-ent pipe during 2-loop operation, the Emergency Core Cooling System will meet I the Acceptance Criteria as presented in 10CFR50.46. That is: L 1. The calculated peak fuel element clad temperature provides margin to the requirement of 2200*F, based on an F value of 3.03. q 2. The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1 percent of the total amount of Zircaloy in the reactor. I 3. The clad tempcrature transient is terminated at a time when the core geo-metry is still amenable to cooling. The clad oxidation limits of 17 per-cent are not exceeded during or after quenching. 4. The core temperature is reduced and decay heat is removed for an extended l period of time, as required by the long-lived radioactivity remaining in i the core. 4 i 4702Q: 1/071283

L

L.-.

L

1 s ,15.4.1.1.3 Results The sequence of events for each case analyzed is shown in Table 15.4-1. Table 15.4-2 presents the peak clad temperatures and hot spot metal reaction for a range of break sizes and locations. This range of break sizes was determined to include the limiting case for peak clad temperature from sensi-tivity studies reported in References [9] and [10]. The SATAN-VI analysis of the loss of ecolant accident is performed at 102 per-cent of the (N-1) License Power Rating. The peak linear power and core power used in the analyses are given in Table 15.4-2. Since there is margin betweert the value of the peak linear power density used in this analysis and the value expected in 2-loop operation, a lower peak clad temperature would be obtained by using the peak lineer power density expected during operation. For the results discussed below, the hot spot is defined to be the location of maximum peak clad temperature. This location is given in Table 15.4-2 for each break size analyzed. 4702Q:l/071383 T. L r' - ~ T T::. :,: -.-

_ _. ~ _ _ _ _ _ _ Y. )5.4.7 References 1. " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors," 10CFR50.46 and Appendix K of 10CFR50. Federal Register, Volume 39, Number 3, January 4, 1974. 2. Bordelon, F. M., Massie, H. W. and Zordon, T. A., " Westinghouse ECCS Evaluation Model - Sumary," WCAP-8339, July 1974. 3. Bordelon, F. M., et al., " SATAN-VI Program: Comprehensive Spacetime Dependent Analysis of Loss of Coolant," WCAP-8302, June 1974 (Proprietary) and WCAP-8306, June 1974 (Non-Proprietary). 4. Bordalon, F. M., et al., "LOCA-IV Program: Loss of Coolant Transient Analysis," WCAP-8301, June 1974 (Proprietary) and WCAP-8305, June 1974 (Non-Proprietary). 5. Kelly, R. D., et al., " Calculation Model for Core Reflooding After a Loss of Coolant Accident (WREFLOOD Code)." WCAP-8170, June 1974 (Proprietary) and WCAP-8771, June 1974 (Non-Proprietary). 6. Bordelon, F. M. and Murphy, E. T., " Containment Pressure Analysis Code (C0CO)," WCAP-8327, June 1974 (Proprietary) and WCAP-8326, June 1974 (Non-Proprietary). 7. " Supplement to the Status Report by the Directorate of Licensing in the Matter of Westinghouse Electric Company ECCS Evaluation Model Conformance to 10CFR50 Appendix K," Federal Register, November 1974. 8. Bordelon, F. M., et al., " Westinghouse ECCS Evaluation Model - Supple-mentary Information," WCAP-8471, April 1975 (Proprietary) and WCAP-8472, April 1975 (Non-Proprietary). lI 9. Salvatori, R., " Westinghouse ECCS - Plant Sensitivity Studies," WCAP-8340, f July 1974 (Proprietary) and WCAP-8356, July 1974 (Non-Proprietary). 4702Q: 1/071383 ,.~_T . : :: ~ ~::. T -

L - - : =.

I, 6 /10. Kemper, R. N., " Westinghouse Emergency Core Cooling System Evaluation Model for Analyzing (N-1) Loop Operation of Plants With Loop Isolation Valves," WCAP-8904, December 1976.

11. Eiche1dinger, C., " Westinghouse ECCS Evaluation Model,1981 Version,"

WCAP-9220-P-A (Proprietary), WCAP-9221-P-A (Non-Proprietary), Revision 1.

12. Rahe, E. P. (Westinghouse). Letter dated November 8,1982 to James R. Miller (USNRC), letter number NS-EPRS-2679.

13. Eicheldinger, C. (Westinghouse). Letter dated September 7, 1977 to J. F. Stolz (USNRC), letter number NS-CE-1540.

/ f I i S 47020:1/071483 T -. - : :z

I TABLE 15.4-1 ) LARGE BREAK i l TIE SEQUENCE OF EVENTS i Inactive Active Active Loop Break, Loop Break, Loop Break, C =0.4 DECLG C =0.6 DECLG C =0.4 DECLG D D D ) (Sec) (Sec) (Sec) START 0.0 0.0 0.0 Rx Trip Signal 0.48 O.46 0.46 S. I. Signal 4.41 2.11 2.66 Acc. Injection 23.60 9.58 12.40 End of Blowdown 41.07 23.25 26.33 Bottom of Core Recovery 56.55 36.57 39.11 ( Acc. Empty 61.96 46.43 51.03 l jl Pump Injection 29.41 27.11 27.66 f End of Bypass 41.07 23.25 26.27 l 1 i i

)

t / TABLE 15.4-2 LARGE BREAK Inactive Ac tive Active Loop Break, Loop Break, Loop Break, C =0.4 DECLG C =0.6 DECLG C =0.4 DECLG D D D Results l Peak Clad Temp., *F 1882 1775 1882 Local Zr/H O Rxn, (max)%

2. 7

2. 3 3.0 2

Peat Clad Location, ft 9.0 7.25 9.0 Local Zr/H 0 Location, ft 9.0 7.25

7. 0 j

2 1 Total Zr/H O Rxn, % <0.3 <0.3 <0.3 2 Hot Rod Burst Time, sec 118.5 73.6 53.8 Hot Rod Burst Location, ft 6.5 6.25 6.0 i Calculation NSSS Power Mwt 102% of 1724 Peak Linear Power kw/ft 102% of 10.2 5 f Peaking Factor (At License Rating) ' 3.03 3 Accumulator Water Volume, ft 1025

'i i

4702Q:1/071583

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~ ~ / TABLE 15.4-3 CONTAINENT DATA (DRY CONTANENT) Free Volume 1.89 x 10 ft3 6 I Initial Conditions Pressure 9.5 psia { Temperature 90*F FWST Temperature 40*F Service Water Temperature 35*F l Outside Temperature 35'F i l Qt.anch Spray System Number of Pumps Operating 2 i Runout Flow Rate (each) 2200 gpm Actuation Time 55 sec

1 i

Recirculation Spray System Number of Pumps Operating 4 Runout Flow Rate (each) 3300 gpm Actuation Time 300 sec i STRUCTURAL HEAT SINKSIII i Wall Number Material Thickness (ft) Surface Area (sq ft) 1 Concrete 0.5 6,972 2 Concrete 1.0 77,446 3 Concrete 1.5 36,848 4 Concrete 2.0 17,010 l ,i 5 Concrete 3.0 ' 8,632 l' 6 Carbon Steel 0.03125 18,270 Concrete 4.5 i 1 47020:1/071483 --w- -e w e +

i 4 TABLE 15.4-3 (Continued) CONTAINENT DATA STRUCTURAL HEAT SINKS (I) Wall Number Material Thickness (ft) Surface Area (sq ft) 7 Carbon Steel 0.03125 32,445 Concrete 4.5 8 Carbon Steel 0.04167 26,250 Concrete 2.5 9 Concrete 2.0 13,125 i Carbon Steel 0.03125 Concrete 10.0 10 Stainless Steel 0.06875 3,270 11 Carbon Steel 0.02202 10,750 12 Carbon Steel 0.06242 748 13 Carbon Steel 0.1532 2,132 14 Carbon Steel 0.1833 5,479 15 Carbon Steel 0.0893 3,770 16 Carbon Steel 0.1041 10,938 17 Carbon Steel 1.020 600 18 Carbon Steel 0.0119 118,091 j 19(2) Stainless Steel 0.0833 2,932.5 (1) All walls are painted with the exception of Walls 9 and 10. The thickness of paint is 5.0 mils for all painted walls with the exception cf Wall 11, which has a paint thickness of 3.75 mils. (2) Inactive Loop Pump Metal. I t 4702Q: 1/071383

TABLE 15.4-4 MASS AND ENERGY RELEASE FOR LIMITING BREAK REFLOOD TRANSIENT (ACTIVE LOOP BREAK, C =0.4 DECLG) D Break Mass Break Energy 4 Time (sec) Flow (lbm/sec) Flow (10 BTU /sec) 39.1 0.0 0.0 39.7 .0242 .00311 40.0 .0243 .00313 40.7 .0245 .00316 ,l 41.9 21.98 2.8353 i 51.6 204.50 8.6411 69.2 305.46 9.9087 91.5 317.78 9.7194 116.7 325.22 9.4086 173.9 337.12 8.7204 241.3 348.95 7.9802 326.4 354.77 7.2195 b /

I l

l. l! !? l! i <j Accumulator nitrogen was released from the accumulators between 58.5 and 78.5 seconds at a flow rate of 181.6 lbm/sec.

i

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i TABLE 15.4-5 BROKEN LOOP ACCUMULATOR FLOW RATE TO CONTAINMENT FOR THE LIMITING BREAK ( ACTIVE LOOP BREAK, C =0.4 DECLG) D Time (sec) Mass Flow Rate (1bm/sec) 0.000 4121.948 1.010 3716.991 2.010 3417.334 3.010 3181.414 4.010 2989.343 I 5.010 2828.876 lt 6.010 2691.456 f 7.010 2572.130 8.010 2466.572 9.010 2372.179 I 10.010 2287.153 11.010 2210.235 12.010 2140.266 13.010 2076.303 \\ 14.010 2017.505 15.010 1963.205 16.010 1912.901 17.010 1866.001 4 18.010 1822.347 l 19.010 1781.747 20.010 1744.203 21.010 1709.189 j 22.010 1676.418 i 23.010 1645.808 l 24.010 1617.065 l 25.010 1589.752 I 29.299 0.0 i i 4702Q:1/071283

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i d

m e e 2 2 j TIME ISEC1 l Figure 15.4-5B. B egk Flow Rgte - DECLG IC U =0.6) i fclave l_oop reak. i 'i 1 1 1

l 't t 1: j o 4 i; I 4 3.00E+05 i J ; i' i l l S.00E*04 + j l U td j-Q E.00E+04 l m 2 i d \\ 4 3oa 4.00E +04 ta. i s x< td .) E m - I i 2.00E +04 i l, Ii l 1, l 1 0.0 1 I M i g 4 +} o 9 0 9 O O r: d 2 2 2 S g i: i TIME ISECl i }j! Figure 15.4-5C. Breqk F10w R te

j Active LOOP break. DECLG (C D =0.41 J '

~ ~ ' ~ ' ~ ' ' ' I i i l 2, 70.0 {- i i ij ,j 50.0 1 i i i I. = 25.0 mb s n. i O i, t '1 E O.0 i W 3 V _N'- m (l I m M i j W K 4 g, -25.0 4 W [- e t-1 O 1 U -50.0 i-s i f 4 il 4 .i -70.0 I I l I E 9 e o o o a I. 9 o d e 2 2 8 3 i i

1 TIME ISEC)

Figure 15.4-6A. C re inac 'ive $_ure %op - DECL G (CD =0.41 res 00p reak. I \\ }.;

i i i o j 70.0 t ? 50.0 l 1

25.0 en i!

b 1 a. -t o i 0.0 hs g ~ ,t 3 4 sn in l bl E l a. i i, -25.0 ha E i o j u t 'i -50.0 '1-li t <l . I

1 i

-70.0 t I g g 9 o 9 o o o o m e m o . i, ~ ~ ,j TIME ISEC1 i 1 l Figure 15.4-6B. Coke Pre $_sure O're$k. DECLG (C D =0.6) ctive oop i 4 k

l i t ir

+j

,!i:i i' l j 1iIt i .t e-02 M. 9S 0 i D C ( G L C ED l 92 C E k S pa I o e E r r M DB I T ep r o u o y ssL ere Pv i 92 i t e c r 4 oA C C 6 A 5 W-1 i O2 e rug i F 4 ( i - f od o. o. o. o. o. o. s o o D o o s z s 7 7 s 2 D .-mE a.OEo wKammwEA mEo0 1: { l i, ,,iI' j ijt i! .i 11 .I 4l 1 E ,i <l1j si

seo ,l ii> lill. ' !I ' <i, la e,Iti i:,. i

1

>l, tl '1l1 -JI: 'll!ijI: 4 i: 1i1 'i l l jl o_ 4a <>d Fw2.o,*oF aoo awOa wm t_ a J t 2 o s s 1 o o 5 o 5 .o o o c o 0 o o .o 0 .o o o o o &O - i [ F igu re 9Oq i i .s 7 .A P Ine ca ck t ivC ?O9 I el a d L .s s oT s o c e p m T Bp I e M r r e E a c t I ku SE re C ROa I l DE C L G ( C SOa D I =0 7 4 1 t 3O9 r iIi!. i,!I t ! li'

li i4!i l i ~ ~ ~ a*l ) 6 0= D g o*I C ( h. l E ~ G ~ L C D '5 52 2 E ~ 7 e r g oO2 u t ak r c ) e C e p E r mB S I e E Tp M O IT dO laL C e l

• O2 v

k i at ec s PA 8 7 4 5 1 e l oO r ug iF Od 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 5 0 5 1 5 0 2 2 1 _l . UWtEo (lO-oOR bOI,n.2 yF $(>4 a<JU n l a a A s. l Il 1 I,' t ,l!I l i:1 - j

i,t i:

t 3i 4 l1 l, ' i' ..i!!

i, l i l! l I t! !i,,iI.i l i ~ ooE ) 4 0 ~ = I ooS D C ^ ( G ~ L ~ C E ~ l D ) o C I e E r S u I t Ea Mrkec '0 T pe I 9 mr eB Tp 0 O E d O a L l I oa0 C e kv ait e c PA C ~_ 7 a Z s I od9 I e rug iF T. b od 0 0 0 0 0 0 0 0 0 0 0 0 0 0 5 0 5 0 0 0 5 0 2 2 1 i ,.2WF d>< o4JU eWWEoWO~ OoE F - - n _l S a i' l i);*' 1jlIi ') ,. ' i i,2 l

ii

1lp) 1 I

i!

l,

!l l i-

I i

.1 2 4I' l!1 l

0*005 T. O l / ~ 0*00& g / / 9 e / / d O lil a O i .xo / / 0*00C ~ I 3C U U o. u g g C

c. O 9

w EJ 88 2 e b >e ~ c -u 3 0 / 0*00Z [c l W I e 1 W j l e u 'j 0*001 [n ~ LL. I F l ) C - l I I i i . I i l 0*C 4 9 9 O O O o. o. o. E o o o g g a e o o I e o E E 2 g m D j (J S33W030 3Wn.l.YW3dr13.L O!n13 1 6 m-m.......,

o00s 7 .l f l / i W O 0*00k t / Q o ~ O / J .t o (d i O = 1 lj 0*00C y ~ x

l 9

E* k 3 i ? 2 a C g E e' o et c. o

1 E o e

_J ll k W a e = 0*002 '3 % ~ u l :. L.s. < CD 1 s/ c,o T li I.D l 1 b G ~ 0*0 03 uo o. Lt. I II I ll u f I I 1 I 0*0 9 9 9 9 o o q h k E 9 e ~ o ll 'l (.3 533WO30 3WO1YW3dFt 310!n7.3 L: .I,' - ~ ~ - - - - - -, -, - -, " ~ ' ' - - ~ ~ ' ', - - " ~ ~. - " -,.., ~, ~ " - -,..,~,--,r i, ' ~ ~ ' ' .e ---.,-,,_---,y

g 0*005 r T 0*00> o' l ~ u O O O J 9 O f.Ll O 0*00E oW l m o.at W wc 2.s uW i o. UQ ~ t w Lo { Q. E Eo e >g 0*002 .c,_ s.- ~u4 Il d. ir i e s t v ID e r 0*0 01 6 3 4' w LL., N ~ I I I I I t 0'0 i o e a 9 9 9 e o i g 3 8 g g 3 o e ~ s n = ~ o IJ 5338030 3Hn1YW3drt 31010'1J I l ,7-_;__;__; ,,,77-------_____

,7 - - - - - ;

.7

y 4 i l{ i! j 'l 7000.0 i 5000.0 2500.0 U 8 til /~ TOP OF CORE 0.0 s l U 7 4 O v\\ j'

d. -2500.0

• j N

\\ ll BOTTOM OF CORE l it l -5000.0 d / -7000.0 I I I I o O 9 0 0 a o 9 2 E S E TIME ISECl Figure is.4-sA. Core Flow - Top and Bot tom - DECL G (C =0.4) Inactive Loop Break D 'e

.I 1 7000.0 5000.0 i I li 2s00.0 w R TOP OF CORE M =! O.0 s-m F-I 4 \\ 1 o - l a -2s00.0 }

t ts..

l !l N !i BOTTOM OF CORE -5000.0 l' t -7000.0 g i g i ii Ii O O 9 9 i o o ^ d e e 2 lQ 4 TIME ISECl Figure 15 A-98. C0re Flow - Top and B0ttOm - DECL G (C D =0,,61

j Active LOOP Break l-

.I

j 1

j 7000.0 I l 5000.0 i 1 2500.0 U id TOP OF CORE o <0 / J ii j! 0.0 s i! l; td ii F 4 i k BOTTOM OF CORE O .i J -2500.0 f i ta. 8 ..:l ey I i -5000.0 1 1 I I I I

1

-7000.0 l! f 'l o o o o o o d e 9 e 2 2 TIME ISECl l Figure 15.4-9C. Core Flow, T0p and Bottom - DECLG (CD =0.4) Active Loop Break 4

]

45 } 20.0 i t 17.5 I DOWNCOMER 15.0 12.5 P h J 'l td> 10.0 1 ll W J l iI tr (

I W 7.5 4l F

l ,i 4 3 } i CORE 1 5.0 t t i i I 1 d. e l! 2.5 i. I i D.0 1 1 I I I o o o o 0( g 8 8 9 o 8 o u n m 'l s 4 TIME ISECl-FruE 15.4-loa. Reflood Transient - DECLG (C D:OA) Inactive Loop Break DomcmER AND Core Water Levels i t, '

ht iI6.lt I' {t j, 9$ kae rB poo I o$ L ev i tcA ER ) O 6 C y 0=s l I o$ C l E De S Cv I e ( E L R M G I E r T Le M O Ct C Ea N W DW OD e r o I o! tnC e ids n n a a r r Tem do oC o n l w f eo oN RD I B o i 4 s 1 i e r u g i F 0 5 0 5 0 5 0 5 0 0 7 5 2 0 7 5 2 O 2 1 1 1 1 FbIu>oJ G p<E -t t ? .i;! 11iji*l jt

)!

I!I' l a ' i i' j i'! .)!J i1 ,!:i !4 i 1

~ ^ ~ ' ' ' .3 i h i 20.0 l I i 17.5 f' DOWNCOMER j r 15.0 12.5 i ,,'i I l-ij to J 1 W l > 10.0 il W 1 a 1 e j g 7.s i

j j

i! E CORE I s.o l' .i 1 a 2.s !} li tj o.o I I I I ji Ii o o o l' a o o o d 2 2 2 S l j TIME ISEC) Figure 15.4-10C. Reflood Transient - DECLGIC D =0.4) Active Loop Break Downcomer and Core Water Levels l 1-

l z.oo l.75 l 5 !j i.so i.zs a 1 3 w 4 z i.oo s. W F 4 it a j i o o.7s i ; o il O i J J: w I a i .o.so-i ; 2 o.as 4 i j + o.oo I I I I 4 1: o o o o. i ; d 8 8 h 2 o o 1! g e a a I i f TIME ISECl fi DECLG (C D 0.4) Inactive Loop Break F laum 15.4-100. thfi..d Transient ,ty Core Inlet Veloci il i f l' li 1

,i t a [ z.oo )! 1.7s i 1.5o I.25 n u 4-w $z U I.oo w p ! }! K l1 O o.75 -l 0 O _J l k

i o.so

]. 4. \\- c.25 l !, o.00 i i i o 9 9 o od 4 .o o o o d 2 2 l g 1 TIME ISECl Figure is.4-ioE. Reflood Transient - DECLG (CD =0.6) Active Loop Break 'l Core Inlet Velocity !i

i t 2.0 \\ i 1.75 l.5 l? f.25 -o W I 2 i} O 1.0 i W l-4 i Er ,l O O.75 i< o O J' J 4r 0.5

I I

r e i; 0.25 4 3i I j 0.0 I i i I o O O O 9 i o o o o o o o o, o Q N d'l WD 4 TIME ISEC) 1 Figure 15.4-10F. Reflood Transient - DECLGIC D =0.4) Active Loop Break Core Inlet Velocity i 9

1 il i I 2500.o 2000.0 't 1' 1] 11 o W 1500.o ! 'i s l

l m

i J i s e I E O .J i, A 1000.0 l,' i 2 i j-a

t u

a l' I o t -i : 1 l 1' 500.o J i ; - t, - r ., t

i

'j o.o i I

i e

i i f* o o o o o E 3 s 2 R S 2 4. TIME iSECl j - Figure is.4-: A. Accumulator Flow iBlowdowns - DECLG (C D :0.41 Inactive Loop break 1. Oi J

.,I 1 5000.o t 4' 4000.0

J

'l >i esij u .i W 3000.o

i R

t in d i; 3: o .J 8a-2000.o i 2 .i i. 3oo< ii i ji 1000.0

i iI 1 i i,

it !l [l o.o I I i +i ). o o 9 o o o o d 9 e o g TIME ISECl l Figure 1s.4-118. Accumulator Flow IBlowdowns - DECLG (CD =0.6) Active Loop Break I

i.,

f

.1

t: l' lL l \\ I 1 j i ;, e Om 1 4 0= I D C I G L C ED l Om C I ) E n S w I o E d M w I T o lB ( k w ae o r lF B I ON r p 0 O t O a L lu m e u v i c t c c A A ~ C I So-I a I s 4 5 ~ 1 e ~ r ug i F O*O 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0, 0 0 0 0 0 0 5 4 3 2 i _ol'R 8-R o I_8 33ou4 8 8 a a x. s ii !;i

!i J

1!'.t itt ..o, $ i l ,1 2 i':' i13 i. k

.l; 3il ' kl- .i' I ,i t(' a 00 4 os 3 G L C ED o ) z l a 4 0 D C I o o l z tn ) S e D is N n O a 0 C r l 4 E T 2 S I e E r M u k s I a T s e e o r o rB l P z p tn o eo mL n o e l s o v l e t i nt oc CA C o l z 3 a 1 4 5 1 l o e s rug iF 0 l 4 o O 0 s t 5 0 o s 2 i -g mG wmammwmG li!!! ',{;ih41 4 s a ;, l. 4I 1! l l , ;i)

,' [

iIt 1!i! 'f,:)!:l. 4

I e l I.0 !'i !'l l i 0.8 ,t 't i i a i o 0.5 k i A i i s i 0.4 a. na I O a. ,\\ i + i j 0.2 !l l 1i i l 4 i 0.0 I I I I o O 9 9 9 0 j-2 E S E o TIME ISEC) Figure 15.4-14A. Core Power Transient - DECL G (C D :O A) Inactive Loop Break .I !i

- 1 I 1.0 i i 0.9 4 l

{

5 o.s ii ( a; s I i I i, o.4 g un3On. i 4 I o.2 l 1. i 4 i I 0.0 I I I I 1 o o 9 o o o lj o m 2 0 2 M l }1 8 TIME ISECl 1 { Figure 15.4-14B. Core Power Transient - DECLG (CD =0.6) Active Loop Brenk 11

i l.0 0.8 I

1. i o 0.6 9

s ? a. i ~ 0.4 , i gr

I ta 3

i o G. i 0.2 4 l 0.0 1 I I I l i!! o 9 O O O o o 9 2 E S S TIME ISEC1 Figure is.4-14C. Core Power Transient - DECL G (C j j Active Loop Break D:0.4) I .i

' ' - - ~ _ r i 5.00E +07 I ., i

i'!

4.00E + 07 J t Mu . l W $3 3.00E

• 07 F

B I \\ >-c @ 2.00E+07 zw M l' a W c' m l.00E+07

i

. l 'd 0.0 I I I I l I o O 9 0 0 o o 9 0 E S E

{

TIME ISEC) . i i Figure 15.4-15. Break Energy Flow, (C D =0.4) DECLG Active LOOP Break 3

I

> i 4

I 2

HEAT TRANSFER COEFFICIENT fBTU/NR-FT..Fl i O E E E E 5 O O .( O O O O O O O O O . 4, o l

1 4

~~ i g i i O s !I a -4 0 l -r m W m -a o O O z it o m 1 ~ no O ng l i 1-u; II Figure 15A-16. Containment Wall Condensing Heat Transfer Coefficient, Active Loop Break (C :0.0 ECW D lt

= i y ? 0*21 0*01 l i G om l o W l C 0*8 LL) t e N w W cL ~ o k o w J H i 0*S T O I e.W Z I W b g I O 9 u 10 0*v oc*o o oo M W E"f o o 9 m u Md oo 3 08 LL. i 2 i O .o i 6 W oo i o=88 g 0z a $@g 6 si 2 !! < Woo.. n i eu ' i I I I I 00 1 o m o m o m o 9 N o N 9 N o o o o o tzix 'f y e a-y- w q--g

---

w,-w

-y-

r 4 ..w ~ ATTACHMENT B w ATTACHMENT B LOCA RELATED TECH SPECS t N-1 LOOP OPERATION Plant Name: Beaver Valley Unit 1 (DLW) Type /Date of CD = 0.4, 0.6 N-1 Active Loop Large Break i t LOCA Analysis: CD = 0.4 N-1 Inactive Loop Large Break 1981 Model June 1983 4 l Total Peaking i l Factor (FQT): 3.03 i f Cold Leg Accumulator f Water Volume: 1025 cubic feet per accumulator (nominal) J r i I 6 Cold Leg Accumulator 4 Gas Pressure: 600 psia (minimum) K(Z) Curve: See attached figure (/F4-/7) 4 l k ~ w '"w' 4 m w n -,,. _ - -}}