ML20100P958

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Rev 0 to Assessment of Doses in Unit 2 Control Room Due to Locked Rotor Accident at Unit 2 Assuming 18% Failed Fuel
ML20100P958
Person / Time
Site: Beaver Valley
Issue date: 01/27/1992
From:
DUQUESNE LIGHT CO.
To:
Shared Package
ML20100P955 List:
References
ERS-MPD-91-035, ERS-MPD-91-035-R00, ERS-MPD-91-35, ERS-MPD-91-35-R, NUDOCS 9202040291
Download: ML20100P958 (62)
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{{#Wiki_filter:- hY A9. 621 O 1 2 3 Health Physics Department REVISION Subject Assessment Of The Doses in The Unit 2 PAGE1 OF Control Room Due To A 1,0cked Rat or Acc i dent At ERS MPD 41-035 M Unit 2 Assuming 18*4 Failed Fuel Reference RCM RP EPP T/S EM DCP Other RIP Review Category 10 CFR 50.59 X RSC Reg'd RSC Not Reg'd b Required Purpose To calculate the dosed t o t he control room personnel resulting from the Locked Rotor Accident at Unit 2. Assumption Of Is', fuel damage and the use of the new fuel is made. i by ante chk date wo dets by date 3 date wo dere by date CM dere w dete by I de,tey ] cM ~ i _m i n a Get*, /4 /j / Checklist Attachments O . [ /ojd f. O Purpose D input Data O Data Sheets e D Assumptions O Results

G litustrations 14 4

Q Methodology Q References

G Printouts

,,, g g g y / /f4,, O Code Listings / E BV RECORDS CENTER E DIR, RadOps 1 E Tmg Dept. E CALCULATION FILE E DIR, RadOps 2 E Author: 9. nn e n., t n 5 MGR, Health Physics O DIR, RadHealth E T2 T <v( m L E DIR, RadEng 5 DIR, EnvSves. O RE 1.1031 99o l 9202040291 920127 PDR ADOCK 05000412 PDR

p N4 DUC UCSOG Of t% F.RS-M PD-91 -03 $ page 2 Hea'tn Physics Department DISCUSSION This analysis is made to document the Control Room (CR) dose estimate ft.t a Locked Rotor Accident (LRA) at Unit 2. This analysis is similar to that done for the Unit 1 LRA, documented in Reference [1], except that the offsite doses are not included in this analysis. They were documented in Calculation ERS-MPD-91-022, Reference [2]. References [1] and [2] should be consulted for further information concerning the background for this calculation. METUODOLDGT This analysis vill be performed using the TRAILS code documented in Reference j [31 TRAILS is a simple FORTRAN program that mechanizes the solution of first order linear equations and dose calculations. Although the code is documented in detail in Reference [3), the more significant methodologies, constants, and assumptions incorporated into the code are listed below: s The code implements a model of the release that incorporates one or two compartments (or nodes) prior to the release to the environment. The transport of radioactive material through the system is governed by simple first order linear equations based on the postulated flow rates and the radionuclide decay constants. those provided in Regulatory Guide The offsite dose calculational methods are 1.4 [ Reft 4). Thyroid dose conversion factors and breathing rates are those provided in TID 14844 [Ref: 5] and Regulatory Guide 1.4. The average energy per disintegration for included isotopes (Kr. Xe, I) vere calculated from the spectra data provided in the DRALIST data library [Ref: 61, which is a subset of i the ORNh Evaluated Nuclear Structure Data File, and which is available in hardcopy as DOE / TIC-11026 [Ref 7). This data source was used in lieu of the suggested 6th Edition, Table Of Isotopes, which is no longer in print. l The Control Room is treated as a compartment with variable (by time step) intake, filtration, X/0, breathing rate, and exhaust rates. ( For this analysis the duration Control Room doses (30 days) vill be analyzed. l The release modeling that vill be utilfred is similar to that used by SVEC in the evaluation of the combined Control Room habitability in 1987 [Ref: 8,9). In that analysis, SVEC performed an evaluation of Control Room doses due to a locked rotor accident at Unit 1 in order to meet Unit 2 licensing commitments, assuming no fuel failures. Since the Control Room doses are sensitive to the timing of Con'rol Room isolation initiated by area radiation monitors, an evaluation of the Control Room dose rate at various times post-accident is assessed first. Based on these results, an isolation time is postulated and included as an analysis assumption during evaluation of Control Room dose over the duration of i l the accident. RE 1.1013 l

s' i Nd.DUCU0S00 s h ERS-MPD m 03S Pa90 3 OfL H en Pucs occartment The analyses address the following telea w components: 1. Release of iodine activity from the thtee steam generators due to technical specification primary to-secondaty leakage. Indine partitioning is assumed. The activity is based on 18% fuel clad tallute, with the fuel clad activity based on St Jatd Peview Plan assumptions (i.e., 30% Kr-85, 10% all others), with the exception at 1-131, which is assumed to be 12% in keeping with the conclusions of NUREc/cR-5009 [ Ret: 10]. 2. Same as #1, but for noble gaset with no credit tot partitioning. 3. Release of activity teleased into the RCS and hence to the three steam generators by an iodine spike that o c c u t <: concurrent with the locked rotor. RCS leak rate based on technical specifications. 4. Release of the initial activity contained in the steam in all three steam generators at time =0, This activity based on technttal specifications. 5. Release of the initial activit', of iodine contained in the secondary liquid in all three steam generatots at time, O. (Noble gases ate assumed to be in the steam space). This activity based on technical specifications. Releases from steam generators are based on the steam mass releases described in Reference 11. INPUT DATA / ASSUMPTIONS 1. General Methodology Based on SRP 15.3.3-15.3.4 [10l 2. Core Inventory From Table 11.1-1 Based on 2766 MVt [11] 3. Core-Gap Fractions Kr-85 0.30 [11) I-131 0.12 [12] Other 0.10 [11) The assumptions of References 11 and I? are included, even though they are not specifically applicable to a locked rotor accident. This is conservative. 4 Fraction Of Rods In DNB - 18% [13) 5. Fraction of Rods Assumed failed = 18% (Section 15.3.3) [10) 6. Concurrent Spike Appearance Rates. uti/sec (Tbl. 15.0-10) [11] I-131 1.36E6 I-132 2.52E6 I-133 3.08E6 I-134 3.68E6 I-135 2.81E6 7. Concurrent Spike Duration 0-4 Hours (Tbl. 15.3-3) [111 Assume appearance rates are 1/500 for 4-8 hours 8. RCS Volume 4.2E5 lbs. (Tbl. 11.1-3) [11] 9. RCS Primary-To-Seconaary Leakage = 1 gpm Total (Tbl. 15.3-3) [11) 10. Steam Generator Mass (Tbl. 15.3-3) [11] Liquid - 99300 lbs. Steam - 8700 lbs. v RE 1.1013

dYDUC UOSne b %k b ERS-MPD-91-035 Page of t to 4 n.smeny.c p.p.m m 11. Steam Generator Mass Release (Tbl. 15.3-3) [11] 0-2 hours = 443,878 lbs. 2-8 hours - 793.644 lbs. 12. Iodine Partitioning In Steam Generators. 0,01 (Tbl.15.3-3) [11] 13. Duration Of Plant Cooldovn Dy Secondary System (Tbl. 15.3-3) [11) 1.e., duration of release - 8 hours 14 X/0 Values For The Control Room (Sec/M ) (14l 0-8 hr.- 1.59E-4 24-96 hr - 5.96E-5 8-24 hr - 7.86E-5 96-720 hr - 3.76E-5 15. Control Room Volume 1.73E5 cu.ft. (Sec. 6.4) [11] 16. Control Room Normal Intake 500 cu.ft./ min. (Sec. 6.4) [11l 300 cu.ft./ min. From Unit 1 And 200 cu.ft./ min. From Unit 2 17. Control Room Pressuritation Rate 690 cu.ft./ min. (Sec. 6.4) [11] e 18. Control Room Infiltration: 10 cu.ft./ min. (Sec. 6.4) [11] 19. Control Room Filter Efficiency: 95% (15] j To account for the filter bypass indicated in Item 18, this is modified by: 690(.95) + 10(0] .936 690 +10 20. Control Room Purge Flow: 19800 cu.ft./ min. (based on the (11] recirculation fan flow rate which can be used to purge the CR) 21. Purge Initiation At 8 Hours [11] 22. Purge Duration.5 Hours (9] CONTROL ROOM MODELING, DOSE RATE EVALUATION For the dose rate evaluation, the Control Room intake and exhaust are set to 500 cu.ft./ min., vith no filtration. Time steps selected vere based on trial calculations which identified the most likely times for the dose rate to reach the safety limit of 1 mR/hr. Note that this value is different than the alarm setpoint of the Control Room monitors. This is explained in Reference [17]. A chart shoving the dose rate versus time was prepared. It is included as. The print outs from the TRAILS runs are included as Attachment 2. The chart shows that the safety limit of 1 mrem / hour is reached at about 2320 seconds ('40 minutes) after the start of the accident. Since it is reasonable to assume operator manual actions after 30 minutes, it vill be assumed that the operators vill manually initiate isolation at T = 30 minutes. CONTROL ROOH MODGLING, DOSE EVALUATION In order to be consistent with the results above, the Control Room isolaticn was assumed to occur at 1800 seconds or 30 minutes. For the one hour period after isolation the only intake is the unfiltered 10 cu.ft./ min. infiltration. The exhaust is equal to the pressurization rate of 690 cu.ft./ min, plus the infiltration for a total of 700 cu ft./ min. RE 1.103-3

n. N4,DUCUOSne 5 d5b? $ip L8ft tRS-MrD-91-035 PaOe of j Heath Physra Depavent At one hour following isolation, the Control Room ventilation realigns for filtered intake up to 1000 cu.ft./ min. In accordance with the U2 UFSAR, the clean-up rate is 690 cu.ft./ min. This is added to the 10 cu.ft./ min. infiltration to yield 700 cu.ft./ min exfiltration. The ait intake necessary to maintain this is the same 700 cu.ft./ min. This ventilation continues until eight hours after the start of the accident. At tight hours a Control Room purge is initiated. It continues until 8.5 hours at which time the normal system arrangement is restored. CALCUl.ATION Source term development, determination of transfer lam'odas, and telease modeling is the same as in Reference [1] and are not repeated here. The addition of the CR to the calculation is accomplished by insertion of suitable parameters for the flows, volumes, filtration rates, and X/0 values. The X/0 parameters vere determined by NUS Corporation using the methodology in Reference [14). The print outs from the cases are included as Attachment 3. RESULTS The results and total from the five cases are as follows: Gamma Dose Beta Dose Thyroid Dose (mrem) (mrem (mrem) Case 1 6.23E-2 4.31E-1 1.07E+3 2 1.07E+1 1.4BE+2 0 3 5.16E-4 3.41E-3 7.39E0 4 1.74E-5 1.23E-4 4.44E-1 5 1.34E-4 1.08E-3 3.58E0 Total (mrem) 1.08E+1 1.48E+2 1.08E+3 CONCLUSION The dose limits in GDC 19. Reference [16] and the Standard Reviev Plan. Reference (10] are : Vhole Body gamet - 5 rem Thyroid - 30 rem Beta Skin dose - 30 rem l l I RE 1.103-3 i ?

W A 4 DOC 00$00 4b @ rRS-MPD-91-035 Pa9e of{4 7 5 6 sea,n % c o.navnem The doses developed in this package ate: Vhole Body gamma - 1.08E-2 rem Thyiold - 1.08 em Beta SIin dose - 1,48E-1 tem The analysis of the 1.ocked Potor Accident shovs that the Conttol Room doses are acceptable and the new fuel can be used, but tequites that manual isolation of the Control Room occur at 30 minutes aftet the stat t of the accident. RE 1,103-3

= _. _. NdDUC00S0G hil$ bkNE ERS-MPD-91-035 7 o ge of a s. e m.c. o.n tEFERENCE_S 1. ERS-SFL-89-021 Safety Analysis of The Dose Consequences of A Locked Rotor j Arcident At BVPS-1 Vith 18% Fuel fallute -- EAB, LPZ, Control Room, 1989 j ~~ ~~ quences of A Locked Rotor Accident At 2. ERS-MPD-91-022, Of f si te Dose Conse Unit 2 Vith 18% FAITe3 ruel 3. DLC, Calculation ERS-SFL~089-020 TRAIL L Ttansport of Radioactive Ma.terial In Linear Systems, 1989 4.

NRC, Assumptions Used for Evaluatint _ The Potential Radiological Consequences Of A Loss 01 Coolant Accident fot Pressutized Vater Reactors, Regulatory Guide 1.4, 19 fi~~

5. USAEC, Calculation of Distance Factors _, Tot Power And Test Reactor Sites, TID-14844, 1962 6. ORNL, DRALIST: Radioactive Decay Data for Application To Radiation Dosimetry And Radiological Assessments, DLC-80, 1981 7. D. C. Kocher, Radioactive Decay Data Tables, DOE / TIC-11026, 1981 8. SVEC, Calculation 12241-UR(B)-449, Doses In The BV1 And BV2 Combined Control Room Due To A Locked Rotor Accident And A Loss of Non-Emergency AC Power To Station Auxiliaties Accident At Unit i Vith 500 cfm Normal Ventilation, 1987 9. SVEC, Calculation 12241-UR(VB)-456 Combined BVI-BV2 Control Room Habitability Due To Design __ Basis Accidents (Except LOCA) At BV1, lI87 10, NRC, 'tandard Reviev Plan For The Reviev Of Safety Analysis Reports For Nuclear Power Plants, LVR Edition. NUREG-0800, 1484 11. DLC, Unit 2 Updated Final Safety Analysis Report

12. NRC. Assessment Of The Use Of Extended Burnup Fuel In Light Vater Power Reactors, NUREG/CR-5009, 1988 13.

Vestinghouse, ltr 89DL*-G-0055, dtd 7/13/89, Rods In DNB For Locked Rotor Event 14. Halliburton NUS Environmental Corporation ltr ARP-91 444, dtd 10/8/91, Control Room X/0 Values for BVPS (Attachment 4) 15. USNRC, Safety Evaluation Report, related to the operation of BVPS, Unit 2 RE 1.1013 l

}VM we ens g _ m, y 7Ar5 LkjT; e/ ^ .,+,m N*m ATT ACWENT l Page of O r) i CONTROL ROOM AIARM RESPONSE mrem / hr n. wronds 120 IJO a /p 'V l.00 ,X 0.90 / ,/ Oh0 / 0.70 O.60 050 / c O.40 / ".30 ,f / 020 / / n2320 030 / sea. 0 ODO 60 300 600 9tK) 1200 1500 1800 2l00 2400 2700 Total from all cases for times showrt CONTROL ROOM ALARM RESPONSE TABLE Doses obtained from the calculations shown in Attachment 2. Units are mrem /hr and seconds. l TIME CASE 1 CASE 2 CASE 3 CASE 4 CASE 5 TOTAL 60 7.97E-06 7.19E-02 3.12E-10 7.04E-06 1.17E-06 7.19E-J? 300 1.92E-04 2.74E-01 2.31E-08 6.86E-06 5.79E-06 2.74E-01 600 7.39E-04 4.36E-01 1.78E-07 6.64E-06 1.14E-05 4.37E-01 900 1.60E-03 5.63E-01 5.82E-07 6.44E-06 1.68E-05 5.65E-01 1200 2.73E-03 6.74E-01 1.34E-06 6.24E-06 2.20E-05 6.77E-01 1500 4.11E-03 7.72E-01 2.54E-06 6.05E-06 2.70E-05 7.76E-01 1800 5.69E-03 8.60E-01 4.27E-06 5.88E-06 3.19E-05 8.66E-01 y 2100 7.47E-03 9.40E-01 6.59E-06 5.70E-06 3.66E-05 9.48E-01 l 2400 9.40E-03 1.01E+00 9.57E-06 5.54E-06 4.11E-05 1.02E+00 2700 1.15E-02 1.08E+00 1.33E-05 5.38E-06 4.55E-05 1.09E+00

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m Ryy E R S /W()J1l-OK' Nf W Page T' p of $(', s.. m c tweav=1 ATTACHMENT tihiL4LLIBURTON NUS Qif innronmens.nl Corporatwn f 9( - I h g L\\ ARP-91-444 October 8, 1991 Mr. Steve La Vie Duquesne Light Company Beaver Valley Power Station P. O. Box 4 S h ipp ine,oort, PA 15077

Subject:

Control Room X/Q Value; for BVPS

References:

1) HALLIBURTON NUS Proposal No. PS808127 2) Duquesne Light Company Purchase Order No. D068239 3) Telecon From S, La Vie to M. Septoff August 9, 4 1991 4) Letter from J. Ramsdell to M. Septoff August 15, 1991

Dear Steve:

this letter presents the X/Q values suitable for design basis accident releases from selected plant vonts to the BVps common control room intake. The methodology used is presented by Ramsdell in NUREG/CR-5055 as modified by NUREG/CP-0116, by discussions with Ramsdell in September 1991 (Reference 4) and a paper presented in Atmospheric Environment, Volume 248, No. 3, pp 377-388, 1990. The work was performed in accordance with *he descriptlon contained in Reference 3 as approved by reference 2. It is noted that all work was performed using approved project prc:edures in accordance with the HALLIBURTON NUS ESD Quality Assuranca Mar :sl, Revision 6. Tables 1 and 2 presents the X/0 values for the selected plant vents for Unit 1 and Unit 2, respectively, for averaging periods of 0-8 hours, 8-24 hours, 1-4 days, and 4-30 days. The release points, distances and directior.s to the coi. anon control room intake, and building cross sectional areas were those provided in an analysis performed by Stone and Webster Corporation using the Murphy and Cr.npa methodology. Duquesne Light Company provided this information to us on September 6, 1991. Five years of meteorological data for the period January 1986 through December 1990 were used. Results of thm Murphy and Campe tiethodology are also presented in the tables for comparison. It is seen that, for the 0-8 hr averaging period reductions in X/O values as compared to those obtained using the Murphy and Campe methodology range f rom a factor of 2 to over an order of magnitude, r ie,1, w,,,m. t u r, n a s p., a, ;m.,,, u i u,n,

N DUC M E R S-pFD- @ l,Q35 Ak D Q Page N of 66 . coa,x: o.am ATTACHMENT Mr. Steve L Vie ARp-91-444 October 8, 1991 Pagn 2 A report describing the methodology and results with a comparison with the Murphy and Campe methodology suitable f or presentation to the NRC will be prepared. The computer code and its documentation will be provided with the report. Further, as we discussed, a scientific paper could also be prepared suitable for presentation at a DOE /NRC Nuclear Air Cleaning Conference. As part of HALLIBURTON NUS Quality Assurance procedures a Division Quality Assurance Review Board (DQARB) was convened to review tne technical and quality aspects of the analysis. The results of the DQARB indicate that (1) the documentation for the computer code generated for the analysin should be revised to provide additional information related to code identification and its use, and (2) the need to clarify the relationship between the test cases and the hand calculations used to verif y the correct operation of the code. Therefore, the results provided in the tables should be considered as preliminary until the DQARB approves the documentation, receives the clarification, and completes its review. It is expected that this will be completed early next week. If you have any questions please call me. Sincerely, (_- o i h -k Michael Septoff Project Manager Enclosures-cc: A. Toblin W. McIntire itALLlitUR10N M 5

i N4.Duc uesne ERS g4 g_g3{ Ab L Page $ of 5 6 w.e Pmcs oemv' ATTACHMENT TABLE 1 DESIGN DASIS ACCIDENT X/Q '!ALUES (sec/m3) FOR CELECTED DVPS PLANT VENTS TO THE COMMON CONTROL ROCM INTME - UNIT 1 R e l e a r.e Point Methodology 0-8 hr 8-24 hr 1-4 day 4-30 day Containment Ramsdell 4.33E-4 2.04E-4 1.46E-4 8.84E-5 Edge Murphy-Campe 2.88E-3 1.90E-3 6.62E-4 1.41E-4 Containment Ramsdell 2.73E-4 1.2BE-4 9.17E-5 5.57E-5 ??r Murphy-Crepe ?.51E-3 1.6dE-3 5.77E-4 1.23E-4 Auailiary Ramsdell 4.30E-3 2.01E-3' 1.49E-3 9.25E-4 Building Murphy-Campe 8.24E-3 6.01E-3 2.32E-3 6.18E-4 Main Ramsdell 7.60E-4 3.51E-4 2.59E-4 1.5BE-4 Steamvalve Murphy-Campe 2.97E-3 1.94E-3 6.92E-4

1. 5 4 E-4 Service Ramsdell 6.25E-4 3.04E-4 2.36E-4 1.57E-4 Building Murphy-Campe 7.47E-3 5.09E-3 1.88E-3 4.93E-4 Turbine Ramsdell 2.43E-3 1.22E-3 8.90E-4 6.26E-4 Building Murphy-Campe 7.81E-3 5.55E-3 2.11E-3 7.11E-4 Cas Waste R3msdell 5.11E-4 2.15E-4 1.65E-4 1.14E-4 l Storage Murphy-Carpe 2.03E-2 1.51E-2 5.99E-3 1.32E-3 Vault H ALLlBL R I v'. M N

4 +.. A.oucuesne E F1S-jp O-9 l- 035 d 9P[5lhjT; A ap [h of $- .e es,< 0 0w,' ATTACHMENT TABLE 2 DESIGN BASIS ACCIDENT X/Q V A LL'ES (sec/m3) FOR SELECTED BVPS PLANT VENTS TO THE COMMON CONTROL ROCM INTAKE - UNIT 2 Roloase Point Methodology 3-8 hr 8-24 hr 1-4 day 4-30 day Containment Ramsdell 1.88E-4 9.32E-5 7.06E-5 4.18E-5 Edge Murphy-Campe 4.16E-3 2.81E-3 1.09E-3 1.58E-4 Containment Ramsdell 1.20E-4 5.91E-5 4.45E-5 2.64E-5 Top Murphy-Campe 3.65E-3 2.47E-3 9.56E-4 1.39E-4 Auxiliary Ramsdell 1.04E-3 5.15E-4 4.04E-4 2.46E-4 Building Murphy-Campe 1.33E-2 9.30E-3 3.94E-3 1.20E-3 Main Ramsdell 1.59E-4 7.d6E-5 5.96E-5 3.76E-5 Steamvalve Murphy-Campe 7.75E-3 5.63E-3 1.99E-3 3.26E-4 Service Ramsdell 2.21E-4 1.11E-4 8.51E-5 5.17E-5 Building Murphy-Campe S.42E-3 3.62E-3 1.43E-3 3.31E-4 Turbine Ramsdell 2.72E-4 1.43E-4 1.10E-4 6.30E-5 Building Murphy-Campe 6.46E-3 4.90E-3 2.00E-3 5.75E-4 Gas Waste Ramsdell

1. 7 4 E-3 9.^6E-4 7.69E-4 5.55E-4 Storage Murphy-Campe 1.23E-1 1.04E-1 5.08E-2 2.00E-2 Vault HALLiliUR :( a M s

s ATTACHMENT B Beaver Valley Power Station, Unit No.2 Additional information - Proposed Technical Speelfication Change No. 57 Information copy of the UFSAR changes to address the radiological consequences of a locked rotor accident l

BVpS-2 UFSAR of a 1[ epa fi'.;er and carbon adsorber with e f f ect ive iodine removal elficiency of 95 percent. These emergency supply filtration nits and associatec air handling equipment are designed to Sei;mic Category 1 and Safety Class 3 requir ement s. A cc.n t ro l roem air manifold system which consists of flexible hose connections to air storage bottles is ptovided to ensure chlorine-free air for up to six (6) hours. The manifold system is located within the common control room and is provided with sufficient lengths of ficxible hoses to reach all vital areas of the common control room. In

addition, a sufficient qu'ntity of portable self-contained breathing apparatus and protective clothing are provided for the operators who are located in the control room.

This equipment is located in the locker rooms near the control room. The evaluation of radiation exposure to personnel in the main control room envelope examined the contribution from the four LOCA sources de fined in Section 6.4.2.5. In addition, the inhalation dose from inleakage into the main control room of radionuclides in the external cloud was also examined for each of the DBAs considered in Chapter 15. Dose calculations are based on the source terms and pertinent paraceters defined in Chapter 15 for each DBA, the flux-to-dose conversion factors given in Table 6.4-2 and the appropriate inhalation dose conversion factors described in TID 14844 by Di Nunno (et al 1962). The limiting design basis accident for the main control room personnel whole body gamma and beta skin dose is the LOCA. The small line break outside containment (SLB) is the limiting DBA for the thyroid dose, Exposure from inhalatica is principally attributable to airborne radioactivity in the main control room envelope due to: 1. Intake prior to main control rrom isolation, m/% m. 2. Inleakage during main control room isolation, or pg pq /, c C R E 6 Ser d g' Act s o E W (L eta) ~' 3. Post-isolation ventilation intake. % f.- The CIB signal isolates the control room almc st immediately after a LOCA. For DBAs that do not cause a CIB

signal, control room isolation is initiated by a high radiatior signal from redundant Category 1 area monitors centrally located 'n he BVPE-2 control room except for the main steamline break (MSLB). hich do# not initiate a high radiation signal.

For the MSLBk manual. operator action by t=30 min post accident is needed to maintfain habitabilityr^ " A O THC L ock E b fo rca A u.').w L selech*nTparanieteWtE' The analyses considers a conservative calculate the thyroid dose. Ventilation intake prior to control room

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'D. i7 BVPS-2 UFSAR !c, 'L isol at ion and an assumed 10 cfm unfiltered inleakage are the main contributors to the thyroid dose. The maximum normal ventilation l R) intake re e of 500 cfm (for both BVPS-1 and BVPS-2 intakes) prior to %Il isol-S on and an minimum clean up rate of 690 cfm post-isolation are 0 f m ed to maximize the dose estimate. The post-isolation clean-up rate { is based on BVPS-1 control room pressurization test data adjusted for j the combined control room volume. The analysis also assumes l coincident loss of offsite power. \\ 9l For Condition IV DBAs which do not initiate a C1B sigral, the a dent duration is 8 hours (except for the fuel handling accident ( 9 \\ (FHA) for which releases are assumed to continue over a 30-day initiate a CIB f, perio ). Control room doses for those DBAs that do not {' c signal (except for the FHA) are therefore based on purging the G control room 8 hours after accident initiation. 49 The information and data required to develop the radiological k consequences for the main control room are presented in the \\ respective sections describing the design basis accident ana lys is. N; The main control room dose ptasented in Table 15.0-13 has been ,/ calculated to be less than the limit specified in General Design Criterion 19 and the main control room may, therefore, be safely occupied during any condition of operation. 5. 4. 4. *, Toxic Gas Protection i The main control room design provides protection of the personnel in the main control room from any toxic effects from spills of chemicals stored onsite. The effects of spills of chemicals along transportation routes are evaluated in Section 2.2.3. In the event of a toxic gas release, main control room habitability is maintained by isolating the air intake, recirculating air conditioned air, and by maintaining a positive pressure using compressed air for 1 hour, after which, the main control room will remain isolated for the duration of the accident. Redundant, sensitive, and automatic Seismic Category I detection and isolation equipment is provided for the detection of chlorine gas. The storage areas of toxic gases and chemicals that could produce toxic gases are shown in Table 6.4-3 and on Figure 6.4-5. 6.4.5 Inspection and Testing Requirements The major items of equipment that maintain the habitability of the main control room are the amergency supply filtration units, their

1ans, 6.'-Ee 6

R " 5' - 2 IWps.2-py3Ag 15,0.9.2 Activ: tes-in the Fuel Pellet Glad Gap for accident a n a l yr, is, the core gap activities are bat.ed on the guidance provided in Regulatory Guides 1.?5 and 1.77, The noble gas and iodine inventory in the fuel gap segion is assumed to be 10 percent (30 percent for Kr-85 and 12 percent for 1-131 for the fuel handling accident of the core inventory. The values are l presented in Table 15.0-1, ' ~ ~ ~ ~,., f h( ~ ^ ^ ' m - %.- -"~ L / -- x ~ I s ) - -) d ^ Cf05$b ~ ~~

i. pypg 4 m-iont 1ro presented in :Fe

  • 1-eneitonmer*

tesult. f om e.n o 1 respective sec. ms 'Ccident a t 'T oh : T;e t i t d1SpelSton C o' 1 1 1 ( I*. ' Q ) for the PAclubIOn me uo used to calculate the nea 1;oundarv n: low populat:on cotential otisite doses. The O pe tent sector-dependent X, Q ,termined as descr: bed in values, presented in Table 1;.0-11, ute < Section 2.3.4. Main cont rol room X/Q s alues fcr the LOCA are also given in Table 15.0-11. The atmospheric releases given in each accident section are used in 1 conjunction with the appropriate X/Q values of Table 15.0 11 to calculate the potential offsite doses for the corresponding accidents ar.d the potential control. room dose due to a LOCA. The methodology fer determining the doses is discussed in Appendix 15A. The ~ resulting EAB and LPZ doses are presented ;n Table 15.0-12 for all postulated acc2 dents. The potential cow to main control room personnel due to a LOCA is presented in Taele 15.0-13. For all cases the potential offsite doses ate within the limits of 10 CFR 100, wLile the potential doses for the main control room due to a LOCA are within the limits of GDC l'1 of Appendix A to 10 CFR 50. 15.0.13 References for Section 15.0 Ecrdelon F.M. et al 1974a. SATAN VI ?rogram: Comprehensive Space Time Dependent Analysis of Loss-of Coolant. VCAP-8302 (Proprietary) and VCAP-8306. Bordelon F.M. et al 1974b. LOCTA IV Program: Loss-of-Coolant Transient Analysis. VCAP-8305.

Eurnett, T.V.T.

et al 1972. LOFTRAS Code Description. WCAP-7907, June 1972. (Also supplementary informa* ion in letter from T.M. Anderson, NS-TMS-1802, May 26, 1978 and :,-rMS-IS24, June 16, 1978.) D Hunin C. 1972. FACTRAN, A FORTRAN IV Code Thermal Transients in a UO: Fuel Rod. VCAP-7908.

Risher, Jr. D.H. and Barry R.F.

1975. TalS1'LE - A Multi-Dimens ional Neutron Kinetics Computer Code. VCD-7979-P-A (Proprietary) and VCAP-8028-A, (Son-Proprietary). U.S. Nuclear Regulatory Commission (CSNRC) 1972. Assumptions Used for Evaluating the Potentia' Radiological Consequence of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Vater Reactors. Safety Guide 25. USNRC 1974 Assumption Used for Evaluating a Control Rod Ejection Accident for Pressurized Vater Reactors. Regulatory Guide 1.77. 3Gvffer mn (Arrwwo /Pojh)J E 0 -

y D eno~ ll$t],4 $(O f, lCl0 l, ?!$ W 0$b [ Ai[b 0 U $ 5 L e n t,o c c o L.c Acri oe s r s c Wc 2 wi n, 'rL <x ) ' HAIL ( f d i t., Y d[ - All't) - 9/- 02), / r {;kh N h$C,, fYl f) , Y {$ $ $ $A' T (W ?l!{ 0 lY 0xm, n /cem As 78 /f low.= o Awa /kcen(),vir me 2 ae U~,r 2 /fu-a mn-as fauso FL n.. f d

Rev. 2 BVPS-2 UFSAR ~ TABLE 15.0-12 POTENTIAL DOSES DUE TO POSTULATED ACCl DENTS (Rem) Exclusion Area Bounaary low Popu t a don _for e* FSAR WhoIe Body Beta WhoIe Body BetJ,. Postulated Accident Section Thyroid Caema~ ikh Thyroid Camma . Rin Main steam line break 15.1.5 Pre-accident fodine spike 10.5

1. 2x 10 - 2 4.6x10-3 1.5
1. na n 10- 3

.ixio-4 Concurrent Iodine spike 9.1 2.2x10-2

6. 7 x10 - 3 3.2
6. e x 10- 3 m2.2xio-3

-I -0 4.1x10-0 2.1x 10- 2 ac power to the station --6. 5x 10 - 5 6 ex to-5 Loss of nonemergency 15.2.6 1.5x10 5.2x10 ~ -m M ~ ~3 4g g_ ; g_,yg_j auxilleries 3 p5 gs; 3,4; y,gq y,,yg+ 1 1.7x10-0 I Locked ro to r 15.3.3 TP-O-Rod ejection 15.4.8 Containment teskage 4.1x10l 1.9x10-I 6.5x10-2 2.0

9. 2x 10 - 3

'3. 2x i o-3 g Secondary side 2.2x10 5.1x10-1

3. 7x 10- 1 1.1 x 10- 2
2. 5 x 10 - 2 1.e iG-2 Small f i ne b rea k - loss-of-15.6.2 1.6x101 7.0x10-2 2.4x10-2 8.2=10-1 3.4x 0- 3 1.2 10-3 coolant Steam generator tube f ui.ture 15.6.3 Pre-accident iodine spike 11.7 6.8x10-2 5,2,io-2 1.0
1. 6 = 10 - 3 2.7x10-3 Concurrent iodine spike 6.0 7.3x10-2 5.2x10-2 1.0 4.6-10-3 3.oxio-3 Loss-of-cootent 15.6.5 Containment teskage 2.7x102 5.3 2.5 1.3=101
2. 6 x 10- 1
1. 2x i o - 1 ECCS leakge
8. 3x10 - 1 1.3x10-2 5.1xio-3 6.3x10-1 1.2x10-2 1.1 x io - 2 i

14ast.e gas system rupture 15.7.1 3.1x10-1 1.9x10-1 Line rupture 1.6x10-1 1.5 Tank rup tu re 1 2.33 6.58 1.4 1.1x10-1 3.2x10-1 Fuel handling 15.7.4 2.)x10 NOTE:

  • For duration of accident 1 of 1

.m

-- -. - ~. - -.. a HVPS-2 UFSAR ,4 -2. Locked Rotor with Two Loops Operating The transient results for this case are shown on Figure 15.3-21 ( t hrough 15.3 4. The results of these calculations are also surmarized in Table 15.3-2. The peak RCS pressure is slightly higher than for the previous case, but is still less than that which would cause stresses to exceed the faulted condition stress limits. The cladding temperature transient is still well below the 2,700'F limit. 3. Locke.d Rotor with Three Loops Operating. Loss of Power to the Femaining Pumps The transient results for this case are shown on Figures 15.3-17 through 1543-20. The results of these calculations are summarized in Table 15.3-2b, The peak RCS pressure reached during the transient is less than that which would cause stresses to exceed the faulted condition stress limits. Also, the peak cladding surface t(mperature is considerably less than 2,700*F. Both -the peak RCS pressure and the peak cladding surface temperature for this-case are similar to the 3-loop transient with power available as discussed on the previous page. -The calculated sequence of events for the three cases analyzed is shown in Table 15.3-1. Figures 15.3-17 and 15.3-21 show that the core flow _ reaches a new equilibrium value by 10 seconds. With the reactor tripped, a stable plant condition will eventually be attained. Normal plant shutdown may then proceed. Following reactor trip, Beaver Valley Power Station - Unit 2 (BVPS-2) will approach a stabilized condition at hot standby; normal plant operating procedures may.then be.followed to maintain a hot condition or to cool the plant to cold shutdown. The operating procedures would call for operator action to control RCS ooron concentration and pressuriser level using the CVCS, and to maintain steam generator level through control of the main feedwar.er system or-AFWS. Any action required of the operator to maintain BVPS-2 in a stabilized condition will be in a time frame in excess of ten minutes following ^ ~'V reactor trip. & town li'l. FAU.Eo F0EL ) 15.3.3.3 Radiological Consequences _h The radiolo k 1. consequences of a postulated locked rotor accident are analyze th the prix:;; and s a c;- hr; ceci nt

e n i;r.s 4

be = Te $'_Mc 2L.- S p ; c i f i c e t i cr. d i m i r c - The primary to ="a +- secondary systes leakage rata is at the Technical' Specification value of 1 gpm. L pp --eivity_in aesn=ad tr i. c ! m vi 4 n e ^ W phy c;;1-6 u..;: te i; nerfuel f a '.lm e pr_tule M. ..The, primary coolant and secondary' side iodine and noble gas concentrationsgare presented in Table 15.0-8. uN _~ ~ w ~ {g-m ~ y,g g mc SYM r e= i.W g /\\0ct02 6 m ieo ozr&& )

1' 's 1 ~ .i BVPS-2 LTSAR TABLE 15.3-3 PARAMETERS USED FOR THE LOCFID ROTOR ACCIDENT Technical Parameter Expected Sy cification Power (MWt) 2,766 2,766 Fraction of fueb wit efects 0.0012 .0026 Q P Rt9 9 > lLMccr o s n r j Primary se ant-concentrations Table 11.1-2 Table 15.0-8 < (f@cR W CC30 g G Secondary-lent ennetntrations Table 11.1-o Table 15.0-8 _,% Cc s O G my,_ QQR_ 0.009 1.0 Primary to secondary leak rate (gpm) Todine partition factor in all steam 0.01 0.01 generators prior to and during the accident Duration of plant cooldown by 8 8 secondary system after accident thr) Steam release from steam generators (lh) 0-2 hr 443,878 443,878 2-8 hr 7 9 'a,6 e a 793,664 feedwater flow to steam generators (Ib) 0-2 hr 527,063 527,065 2-8 hr 874,470 874,470 Steam generator fluid content /SG (1b) Liquid 99,300 99,300 Steam 8,700 8,700 Concurrent iodine spiking Release rates into primary Table 15.0-10 Table 15.0-10 coolant Duration (hr) 4 4 i v' depoMO -

e v., ,t* el BVPS-2 UFSAR TABLE 15,3-4 LOCKED ROTOR ACCIDENT RELEASES TO THE ENVIRDNMEKr x f [ ~ Total Releases (C1) 0-2 Hr 0-8 Hr \\ Nuclide / s !340E4 7x10"* 6 7x'O'8 iM5 E +2 \\ Kr-83m ~ 1 h.'i I E #3 Xr-85m

1. I t E 4 3 2.3x1h
5. x 0 Kr-85 (ME + 1 1.

5. t 3. 6 4 E + 2. f~8 10'5f 2.2 6 E t3 Kr-87 l.W E +3 8. xf0 1. 1 1 1 Kr*88 J.%eE *3 f 3. ,0 6.6a10 (o. 2 2 E +.3 Kr-89 1 '/9 E +2,'. 8 LO 4.8 10t I E D t l' ! 2 '/. l i E t-8 Xe-131m l,goE+1 1.3 10 5, 10 Xe-133m l $7G +2.!.5x 0'1 1.i \\ { p[p.ol 6 f A 3 1.px!0 . 67 E +V 8 Xe-133 gg(3.1 3.'53E+2. Xe-135m 33 eE 4 2 7.9> 1p'*

1. )

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