ML20199J439

Proposed Tech Specs Re Radiological Doses for Locked Rotor Accident

ML20199J439
Person / Time
Site:	Beaver Valley
Issue date:	01/29/1998
From:	DUQUESNE LIGHT CO.
To:
Shared Package
ML20199J414	List: L-98-016, Proposed LAR 120 to License NPF-73,allowing Increased Calculated Doses as Described in UFSAR & Request NRC Approval of Small Break Loss of Coolant Analysis Radiological Consequences Evaluation ML20199J439
References
NUDOCS 9802050314
Download: ML20199J439 (59)

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ATTACHMENT A-1 Beaver Valley Power Station, Unit No. 2 Proposed License Amendment Request No. 120 LOCKED ROTOR DOSE REVISION The following is a list of the affected pages:

Affected UFSAR Pages: 6.4-8a 15.0-12 15.0-15 15.3-10a 15.3-11 Table 15.0-7 Table 15.0-11 Table 15.0-12 Table 15.0-13 Table 15.3-3 Table 15.3-4 9802050314 980129 PDR ADOCK 05000412 p PDR

. - ~ - - - - ---. ----. --. .~.....-.-. - . - - - - -. - . . - - . --_ ...

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s BVPS-2 UFSAR 6 $Nfth isolation and an as used 10- cfm unfiltered inleakage are the main contributors to the thyroid dose. . The maximum normal ventilation intake rate of 500 fa (for both SVPS-1 and BVPS-2 intakes) prior to isolation and an - '-- -clean up rate ;f 500 ;;_ post-isolation are I used to maximise the dose estimate. Ti; ;::: i::1;;i:: ;i:;; ;p ;;;;

1.

L; L;;;f  ; . "J7; '

:;;;;' ;;;_ e.; :_;i;;;i;; ;;; d;;; ;dj_;;;d 0;.

12in:f ;. 't;;i ;: ::_ T;1

. The analysis also assumes coincident loss of offsite power.

For Condition IV DBAs which do not initiate a CIB signal, the accident duration is 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (except for the fuel handling accident (FRA) 'h; irr'-f  ::::: ;: i' ;  ;"Pf; for which releases are assumed to continue over a 30-day period).

The information and data required to develop the radiological, consequences for the main control room are presented in the respective sections describing the design basis accident analysis.

The main control room dose presented 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.

6.4.4.2 Toxic Gas Protection 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 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, after which, the main control room will remain isolated for the duration of the accident.

Redundant, sensitive, and automatic Saismic 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 emergency supply filtration units, their fans, l

6.4-8a a

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BVPS-2 UFSAR I 15.0.9.3 Primary and Secondary Side Coolant Activities The equilibrium concentrations in the RCS and the secondary coolant '

system have been calculated assuming full power operation for the following cases: 1) one percent fuel defects, 2) normal operations using the guidelines of NUREG 0017 (l'SNRC 1976), and 3) plant Technical Specification iodine concentrations. The Technical Specification activities are used in the analysis of the main steam line break (MSLB), the locked rotor accident, the rod ejection accident, the failure of small lines carrying primary coolant outside containment, and the steam generator tube rupture. The Technical Specifications for BVPS-2 restrict the concentration in the primary and secondary systems to 1.0 and 0.1 yCi/gm I-131 dose equivalent, respectively. The 1.0 pCi/gm limit set for the primary system is equivalent to operating the plant with fuel defects in that fraction of the reactor core which produces approximately 0.26 percent of the power. The resulting iodine and noble gas concentrations in the primary coolant and secondary liquid and steam phases are presented in Table 15.0-8.

For the waste gas system rupture analys is , primary coolant concentrations with 1 percent fuel defects are assumed. These RCS concentrations are given in Table 11.1-2. The calculation of releases due to a liquid-containing tank failure uses expected normal operation concentrations of 0.12 percent fuel defects. These concentrations are also presented in Table 11.1-2, 15.0.9.4 Iodine Spiking Concentrations The analysis of an MSLB,1:r'ed ::t:: :::if:nt, steam generator tube rupture, and the failure of small lines carrying primary coolant outside containment include equilibrium coolant iodine concentrations augmented by iodine spiking. Both pre-accident and concurrent iodine spiking models are considered.

The pre-accident icdine spiking concentrations are determined by increasing the primary coolant iodine concentrations to the maximum value described in the Technical Specifications. The resulting primary coolant iodine concentrations are given in Table 15.0-9.

The concurrent iodine spike is modelled by increasing the iodine release rates from fuel rods into the primary coolant to a value which exceeds 500 times the equilibrium iodine concentration release rates, Table 15.0-10 presents the iodine release rates for concurrent iodine spiking.

15.0.10 Residual Decay Heat 15.0.10.1 Total Residual' Heat Residual heat in a suberitical core is calculated for the LOCA per the requirements of Appendix K, 10 CFR 50.46, as described by 15.0-12

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BVPS-2 UFSAR Rev. 5 the environment resulting from each accident are presented in the respective sections. j 1

Accident atmospheric dispersion coefficients (X/Q) for the exclusion area. boundary and low population zone were used to calculate the potential offsite doses. The 0.5 percent sector-dependent X/Q values, presented in Table 15.0-11, were determined as described in section -2.3.4. Main control room X/Q valuer for the LOCA are also given in Table 15.0-11.

M C M8T N The atmospheric releases ;i;; in each accident section are used in conjunction with the appropriate X/Q values of Table 15.0-11 to calculate the potential of fsite doses for the corresponding accidents and the potential control room dose due to a LOCA. The methodology for determining the doses is discussed in Appendix 15A. The resulting EAR and LPZ doses are presented in Table 15.0-12 for all postulated accidents. The potential dose to main control room personnel due to a LOCA is presented in Table 15.0-13.

2 For all cases the potential offsite doses are within the limits of 10 CFR 100, while the potential doses for the main control room due to a LOCA are within the limits of GDC 19 of Appendix A to 10 CFR 50.

15.0.13 References for Section 15.0 Bordelon F.M. et al 1974a. SATAN-VI Program Comprehensive space Time Dependent Analysis of Loss-of-Coolant. WCAP-8302 (Proprietary) and WCAP-8306.

/'

Bordelon ,F.M. et al 1974b. LOCTA-IV Program Loss-of-Coolant Transient Analysis. WCAP-8305.

Burnett, T.W.T. et al 1972. LOFTRAN Code Description. WCAP-7907, June 1972. (Also supplementary information in letter from T.M.

Anderson, NS-TMS-1802, May 26, 1978 and NS-TMS-1824, June 16, 1978.)

I

__ _.t;, " 7. 1001. Tff;it; ;;; 2;;;;;_;;;;; ;f ; '_;;h;d n;;;; I a...l_... ;; J;i; _i;h 10^. 7.11.4 T. 1,  ;;; .. ;

1 0::.

E & n;n-v-atr

..._-_ , " r.1001. ,Shseessment of the Deses in the Unit 2 control Room Due to a Locked Rotor Accident at Unit 2 Assuming 18% Failed Fuelf Hunin C. 1972. FACTRAN, A FORTRAN IV Code Thermal Transients in a UO2 Fuel Rod. WCAP-7908.

Risher, Jr. D.H. and Barry R.F. 1975. TWINKLE - A Multi-Dimensional Neutron Kinetics Computer Code. WCAP-7979-P-A (Proprietary) and WCAP-8028-A, (Non-Proprietary).

15.0-15 e- , . ,

i-BVPs-2 UrsiJL Rev. 4 For the locked rotor accic nt, it is also postulated that iodine spiking occurs concurrently with the locked rotor event. The concurrent iodine spiking contribution to the atmospheric releases is based on the iodine appoa.ance rates into the reactor coolant given in Table 15.0-10. The iodine spiking phenomenon is assumed to (continueforfourhoursfromthestarte.,!theaccident._.

The radiological dose consequences from a locked rotor accident are evaluated based on the assumptione presented in Table 15.3-3. offsite power 1r assumed lost, thereby making the condenser unavailable for steam dump. Primary system leakage o the secondary side continues until the primary siae pressure and secondary side pressure are equalised. Activity is released to the environment at 15,3-10a

-BVPS*2 UTSAR i .

ground level until secondary side pressure decreases below the relief valve actuation value.

_The environmental rt, leases l

i shown in Table 15.0-4 are combined withfor a postulated locked rotor accident, l

the atmospheric dispersion values presented in Table 15.0-11 to determine the exclusion ares boundary _and low population zone doses given in Table 1$.0 12.

The methodology' used in calculating -the offsite doses is discussed in Anoendix 15A. The radiological consequences for a locked rotor event with concurrent iocine spikiNT) do not exceed a small fraction of

-10 CFR 100 guidelines.

15.3.3.4 Conclusions Since the peak RCS pressure reached during any-of the transients is less than that which would condition stress limits, thecause integrity stresses to exceed the faulted of the primary coolant system is not-endangered.

Since the peak cladding surface temperature calculated for the hot spot during the worst transient remains considerably less than 2,700*F, ,the core will remain in place and intact with no loss of core cooling capability. No fuel failures are predicted for the locked rotor accident (Van Houten 1979).

15.3.4 Reactor Coolant Pump Shaft Break 15.3.4.1 Identification of Causes and Accident Description The accident is postulated as an instantaneous failure of an RCP shaft, such as discussed in Section 5.4. Flow through the affected reactor coolant loop is rapidly reduced, though the initial rate of

reduction of coolant flow is greater for the RCP rotor seizure event.

Reactor trip _ in initiated on a low flow signal in the affected loop.

Following initiation of the reactor trip, heat stored in the 1.e1 rads continues to be transferred to the coolant causing the coolt.nt to expand. At _the same time, heat transfer to the- shell side of the

' asteam generator-is reduced, first because the reduced flow results in decreased tube side film coefficient and then because the reactot coolant la the tubes cools down while the shell side temperature increases (turbine steam flow is reduced to zero upon plant trip).

The rapid expansion of the coolant in the reactor core, combined with reduced heat transfer in the steam generators causes an insurge into

-the pressuriser and a pressure increase throughout the RCS. The insurge' into the pressurizer compresses the steam volume, actuates the automatic spray system, opens the PORVs, and_ opens the pressurizer safety valves, in that sequence. The PORVs are designed for reliable operation and would be expected to function properly during the accident.. However, for conservatism, reducing effect as well .s the pressu e reducing effecttheir pressure of the spray is not included in the analysis.

15.3 11

4 BVPS 3 UFSAR Rev. 3 TABLE 15.0-7 IODINE AND NOBLE GAS INVENTORY IN REACTOR CORE AND FUEL ROD GAPS

• Fraction of Activity Core Core Activity in Fuel Rod Gaps Nuclide (Ci) in Gap (C1)

I-131 6.9x10' O.1+ 6.9x10' I-132 9.9x10' O.1 .0' I 133 1.6x10' O.1 10' I-134 1.8x10' O.1 1.dx10' I-135 1.4x10' O.1 1.4x10' Kr 83m 1.2x10' 0.1 1.2x10' Kr 85m 3.0x10' O.1 3.0x10' Kr &S 6.8x10' -0.1** 6.8x10" Kr-87 5.9x10' O.1 5.9x10' Kr 88 8.3x10' O.1 8.3x10' Kr-89 1.1x10' O.1 1.1x10' Xe 131m 4.2x10' O.1 4.2x10" Xe 133m 3.7x10' O.1 3.7x10' Xe 133 1.6x10' O.1 1.6x10' Xe 135m 4.2x10' O1 4. 410' Xe 135 4.1x10' O.1 4.1x10' Xe-137 1.4x10' O.1 1.4x10' Xe 138 1.4x10' O.1 1.4x10' NOTES:

• Based on 650 days of operation at 2,766 Wt

• • The fuel hand 11 acciden/ analyzed in :ccordance with Regulatory Guid 1.25, usin .3 as the fractional releast of Kr-85. W4M  :

+The fuel han in acciden analyzed in accordance with Regulatory C ide .25 and NUREG/CR 5009, using 0.12 as the frac 1 1on release of I-131.

\sa a i. a a a 1 of 1 ,

e BVPS-2 UFSAR R *,;v . 5 TABLE 15.0-11 ACCIDENT METEOROLOGICAL PARAMETERS Exclusion area boundary - 547 m, NW 0-2 hr 1.44 l, S Low population zone - 5,794 m, NW 0-0 hr 0.0707 0.060Vb 8-24 hr 0.0516 deOV77 1-4 days 0.0259 d.#Al d 4-30 days 0.00963 d.00?YVb Control roome*

0-8

• 8-24 1-4 4-30 Relaana Point ggggg ggggg h g containment Bldg.

- top 3.65 2.47 0.956 0.139

- edge 4.16 2.81 1.09 0.158 Auxiliary Bldg. 13 3 9.30 3.94 1.20 Main Steam Valve House 7.75 5.63 1.99 0.326 Service Bldg. 5.42 3.62 1.43 0.331 Turbine aldg. 6.46 4.90 2.00 0.575 Caseous Waste Storage Vault 123 104 50.8 20.0 NOTE:

A*.1 values are in X/r) (x10~3 sec/m3 )

Control X/Q values recalculated in 1991. See Table 15.0-14.

The values shown above were used in analysis performed prior to 1/92, and are shown here for historical purposeo.

J (o rAne values wen use,/ See na.<tyst., ,+ .4 g,4,)g heeNe4, Ms - en ont, 1

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CVPS m .FSAR dev. 7 TABLE 15.0-12 POTENTIAL DOSES DUE TO POSTULATED ACCIDENTS (Rem)

Exclusion Area Boundary Low Population Zone *- l FSAR Whole Body Beta whole Body Beta l Postulated Accident Section LhYroid Camuna Skin Thyroid Canana Skin Main steam line treak 15.1.5  ;

Pro-accident Iodine spike 10.5 1.2x104 4. 6 x10-3 1.5 . 4 x 10-3 6.1x104 Concurrent Iodine spike 9.1 2.2x10 4 6. 7x 10-3 3.2 6. 8 x 10-3 2. 2x 10-3 4 d Loss of nonemergency 15.2.6 1.5x10 4 5.2x10 4.1x10 2.1x10-2 6.5x10 4 6.8x10 4 cc power to the station  ;

auxiliaries f

~3,1 XN' 3*$ dR /*$KM 'T.$X5 2.~3 FAF'l Locked rotor 15.3.3 2. _ ;0' O.;; -eree-  :.::_;;' _: _;;' O ;._;;'

Rod ojection 15.4.8 containment leakage 4.1x10 8 1.9x10 4 6.5x10 4 2.0 9.2x104 3.2x10-3 Secondary side 2. 2x10-8 5.1x10 4 3.7x10 4 1.1x104 2.5x104 1.8x104 Small line L eak - loss-of- 15.6.2 1.6x10' 7.Dx104 2.4x104 8.2x104 3.4x10 4 1.2x104 coolant Eteam generator ts.be rupture 15.6.3 Pre-accident iodine spike 71.6 2.Ox104 1.0x104 3.6 7.0x10 4 5.0x10 4 concurrent iodine spike 13.4 2.Ox10-8 2.0x104 8.0x10 4 9.0x10 4 7.0x104 Loss-of-coolant 15.6.5 containment leakage 2.7410 3 5.3 2.5 1.3x10' 2.6x104 1.2x104  ;

ECCS leakage 8.3x104 1.3x104 5.1x10-3 6.3x10 4 1.2x104 1.1:t104 ECCS backleakage to RWST 0.0 0.0 0.0 6.9 7.0x10 4 3.4x104 Wasto gas system rupture 15.7.1 Line rupture' 3.1x104 1.9x104 Tank rupture 1.6x104 1.5 Fuel handling 15.7.4 2.9x10 1 2.33 6.58 1.4 1.1x10 4 3.2x10 4 Thyroid CDW ggg skin DE Thyroid CDE g skin DE Small Break LOCA 15.6.5.5 2.9x102 2.2 1.3 2.0x10' 1.3x104 7.7x10 4 l

NOTE:

• For duration of accident i I of 1  !

i SVPS-2 UFSAR Rev. 7 TABLE 15.0-13

cont rol Roorn Dosas, ram, From Desian Basis Accidents 1

Accident Thyroid f33 sag Egla Notes Main steam Line Break Co-incident Spike 11.0 6.8E-4 4.8E-3 3ff Pre-incident spike 5.4 1.8E-4 1.7E-3 3,7 l Small Line Break 8.1 8.0E-4 7.7E-3 3,7 steam Generator Tube Rupture Co-incident spike 1.9 3.0E-4 6.1E-3 3, f Pre-incident spike 8.7 5.0E-4 7.9E-3 3, f Rod Ejection Accident 4.9 4.9E-4 3.8E-3 3f f Fuel Handling Accident 2,3 9.3E-3 5.3E-1 3j s*

/e7 f. $ f*1

2. 3 5 */

Locked Rotor Recident 'ev4== + vet-t== 1,4f 7 Loss of Auxiliary AC Power 2.1 1.8E-4 1.2E-2 3[

Waste Gas System Rupture Line preak --

5.8E-2 1.3 3ff Tank Rupture -- 3.5E-2 9.7 4f DBA LOCA 1.3 3.2E-1 1.2E-1 2f j Thyroid CDE EDE Skin DE Small Break LOCA0 11.0 3.2E-3 3.2E-2 Notes 1: Isolation by manual operator action at T=30 minutes post-accident.

2: Control Isolation actuated by CIB signal.

3: Nc action requ!. red.

4: Purge of Control Room atmosphere for +-ed, menus > ed 30 minutes at

/$, $M ecfm at no later than T=8 he post-accident initiation.

5: References ERS-SFL-93-004 l 6: References ER8-8FL-94-014 7 AebMt ' Ed$- M96~ fl 035" l

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BVPS-2 UFSAR Roe. $

TABLE 15.3-3 PARAMETERS USED TOR THE LOCKED ROTOR RCCIDENT Jed..is.1 Parameter .ro-- & 2 - e-- >>>-.. ..

Power (HWt) 2,766 4 2,766 Fraction of fuel with defects 0.0012 .0026 pr).or to the accident l

Primary coolant concentrations able 11.1- 2 Table 15.0-8  ;

prior to the accident i 1 Secondary coolant. concentrations l T ble 11. -6 Table 15.0-8 g l prior to the aceirtent l l Primary to secondary leak rate (gpm) 0. 09 1.0 Zodine partition factor in all steam 0.01 generators prior to and during the f 0.0 accident ,

Duration of plant cooldown by 8 8 secondary system after accident (hr)

Steam telease from steam generators (1b) 0-2 hr 443,8 443,878 2-8 hr 793,6 793,664 Feedwater flow to steam generators (1b) 0-2 hr 527, 65 527,065 2-8 hr 874 470 874,470 Sceam generator fluid content /SG (1b)

Liquid 9 ,300 ^

-;,0^0 M2 coo Steam 8 700 4rMG.*1,tio Cw.e-...nw bdin er! ;

.ici;;ee .;t:: inte pr'-- y_ able 15.0 10 =Tebi; 15 0-10 c cc12:.

03:ttle- (ts; < 4 &

Ctv NOI sm pmst vit*hcwbn flow ; c.fm iOJO C4M N #M f W y b6'W g C.h ((p f 00 CRES APS flow cost 4 mad, c M C

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1

+

• . SVPS-2 UrsAR R:v. 5 TABLE 13.3-4 LOCKED ROTOR ACCIDENT RELEASES TO THC ENVIRONMENT Total Relaamas fCi Nucli 0-2 Mr 0-[ Hr Kr-83m 3.60x10 2 6.45x10 2 Kr-85m 1.11x10 3 2.94x10 3 Kr-85 8.88x10 1 3.54x10 2 Kr-87 1.54 0 3 2.28x10 3 Kr-88 2 0x10 3 6.22x10 3 Kr-89 1.79x10 2 1.79x10 2 Xe-131m 1.80x10 1 7.11x10 1

• Xe-133m 1.57x10 2 6.01x10 2 Xe-133 6.82x10 3 2.67x10 4 Xe-135m 3.31x10 2 3.33x10 2 i

Xe-135 1.63x10 3 5.25x10 3 Xe-137 2. 6x10 2 2.77x10 2 Xe-138 1.02 0 3 1.02x10 3 I-131 2.64x10 2.56x10 2 1-132 2.13x10 1 8.08x10 1 1-133 4.89x10 1 4.26x10 2 I-1 4 2.16x10 1 3. 9x10 1

-135 3.89x10 1 2.64x 0 3 4

kleje tbd kWe, & 6 l etdad by aC-uenee.

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, ATTACHMENT A-3 Beaver Valley Power Station, Unit No. 2 Proposed License Amendment Request No. 120 REVISED SMALL BREAK LOCA CALCULATION PARA 11ETERS s

The following is a list of the affected pages:

Affected UFSAR Pages 15.6-22 15.6-23 Table 15.0-13 Table 15.6-15 Table 15.6-16 l

l

BVPS-2 UFSAR Proposed Changes to UFSAR 15.6.5.5,

" Radiological Consequences of a Small Break LOCA"

. (Pages 15.6 22,23) ne radiological consequences adbessed in Section 15.6.5.4 are based on the desien basis LOCA analysis which l Amingsa4arge"' ' ^" that-tesulte4n the immediate release of 100 percent of the core inventory of noble l gases, 50 percent of the core inventory of halogens, and I percent of the core invene ry of other nuclides, to the containment atmosphere. ne containment pressure exceeds the CIB setpoint within seconds and the containment sprays actuate at this time to lower pressure and to scavenge radioiodine from the containment atmosphere.

In the event of a small break LOCA, the increase in containment pressure is not as rapid, and t: e actuation of j sprays is delayed. He delay in spray response reduces the overall efficiency of the spray for removing j radiciodine from the containment atmosphere. Westanchouse Nuclear Sar ctv Advisorv Letter 93-016 succested thc20ssibililylt i: pezible that there could beare combinations of break size, fuel damage, and spray actuation .

delay for which the offsite dose consequences might be greatermor: !%iting than those calculated innostulated-by I l

the large break LOCA analysis-provi&d in S= tic 15.6.5 A.-his4ssuewas-*nitially-raised 4>y-Westinghouse 4n Nuclear 4afety-Admory-Letter 4M16. An evaluation of the issue concluded that a substantial safety hazard per 10 CFR 21 criteria did not exist._ Calculated PCT for small break LOCA is much less than for larce breal t i,0CA and indicates that small break LOCA is less likely to result in fuel failures. Althouch Reculatory Guide 1.70_qaly frgnests thalanalyses of radiolocical conscauences be provided for limitine events. an analysis has been pgrformed for the purpose of comoarin.g_small breaklOCA conseauences (with delayed containment IEray effects) with the desien basis analysis rssults. This analysis is not intended to be ne desien basis acr.h sis as dgfined in Reculatory Guide 130 but is a more realistic analysis for comoarison eurooses only.

Analy=: cf1: dc= :==q=== cf postub::d LOGA: re peric=d in acecrd==witlHhe-tuidanee4n USNRG4egulatory-Ga!& ! 4 =d Chap:= !5.6 5 of S St=&s! P: view P!= (SRP) 3: SRP+equitet4he radiological dc= c= qu== analysis 4c be b=ed on 1: ==t =vn: LOGA---This4erm49-not-dermed4n4he SRP, hcw=r, it4*-und h : = 6: d=ign b=i: LOGA4 hat yi:!61 gr=:=t i;=c/hydraulie challenge 4e4he containment-and4GCS. n; radiologieal,~..y...-analy:i: i: b=d on 6 me:: xv=: LOCA : a :aeans-of showmg-compliance vie de dc= gui&lhe of 10 CFR 100. In ec=:rastrthe+ mall-break-LOGA-analyses-are perkrmed40 evaluate 4GCS p=fc=== f= Se mest4tmiting LOCA, ehkh i: :cken-here4ebe4he42OGA-that generates 4he4tigh=: p=k chd4emperat= for ; given =: of-analy:!: input values-and-assumptions--Becauseof this<hfrerencein4ntent;4RP Chap:= 15.6.5 i: act considered 40-be :ppli=bk :o 1: =allbrealebOGA4ssue-Although4heemal! b=h LOCA ir i: act dx=d 10 be =bj=:4e SRP RS.5, an :=lysitefahe-postulated radiological consequences *as4aitittd T =ppe:1 As cart olthis analyris, a containment pressure transient analysis was performed.

For the anahredmost4imiting small break LOCA, the containment pressure exceeds atmospheric pressure at 268 seconds following the b cak. CIB (and contaisunent sprays) actuate at 1630 seconds (less than 30 minuteRaner the break and the containment pressure is subatmospheric at 3390 seconds after the break. The radiological consequences analysis was performed assuming the release from the core to be limited to gap activity (10 percent core inventory noble gases and radioiodines) for the first 30 minutes of the event. This assumption is consistent with that used for a desien basis rod control cluster eiection accident and is believed to be more realistic than assumine an instan.taneous release of 100 pitEtnL At the end of this period, additional core inventory is assumed to be released, as a step increase, to result in a total release to the containment of 100 percent noble gases and 50 percent halogens. This staged release of core inventory is d.more reahstic mode. ling _of==i =' uith a break of this size than would be the traditional T=0 instantaneous release, l and is consistent with the results of severe accident research, Dose quantities deternuned were EDE, skin DE, and thyroid CDE. Other analysis inputs and assumptions are consistent with UFSAR Appendix 15A, SRP 15.6.5, and Regulatory Guide I A. The analysis inputs are tabulated in Table 15.615. The postulated doses are tabulated in Table 15.6-16. The postulated nuclide releases due to containment leakage and leakage from ESF equipment outside of containment are tabulated in Table 15.617.

- . ~ - - - . _ _ _ _ _ _ - _ - ~ . _ - - - _ . . . _ _ - - . _ . . - _ . , - . . _ - . - - _ _

I BVPs-2 UFSAR Rev. 7 '

TABLE 15.0-13 control Room Domes, rash FromDamianBasisAccidentm5h Accident Thyroid Gamma 3314 Notes Main Steam Line Break co-incident Spike 11.0 6.8E-4 4.8E-3 3 Pre-incident Spike 5.4 1.8E-4 1.7E-3 3 small Line areak 8.1 8.0E-4 7.7E-3 3 steam Generator Tube Rupture co-incident spiko 1.9 3.08-4 6.1E-3 3 Pre-incident spike 8.7 5.0E-4 7.9E-3 3 Rod Ejection Accident 4.9 4.9E-4 3.8E-3 3 Fuel Handling Accident 2.3 9.3E-3 5.3E-1 3 Locked Rotor Accident 1.1 1.1E-2 1.5E-1 1,4 Loss of Auxiliary Ac Power 2.1 1.8E-4 1.2E-2 3 waste Gas System Rupture Line Break --

5.3F-2 1.3 3 Tank Rupture --

3.5E-2 9.7 3 CBA LOCA 1.3 3.2E-1 1.2E-1 2 7hyroid CDE EQR Skin DE Small Break LOCA6 11.0 3.2E-3 3.2E-2 Notes 1: Isolation by manual operator action at T=30 minutes post-accident.

2 control Isolation actuated by CIS signal.

3: No action required.

4: Purge of Control Room atmospisere fc'r a minimum nf 30 minutes at 20,000 cfm at no later than T=8 hr post-accident initiation.

5: References ERS-sFL-93-004 g x l 6: References ERS-SFL-94-014 ~ Values -6>c swll j,4 L,p gg Qg

'b 100 7. -Q re, &ck gg

%asad ons ag .oesa, u kaaa.,p,,,,4 4 ,e.,g _ _ ,

V

& Liska d** valas repriscid -fA &,ugey ,,to,3, g;g ,Y g g3p kn enrrsA analysis nsults. -

1 of 1

.. SVPS-2 UFSAR Rev. 1 TABLE 15.6-15 (Cont)

Noteorology EAB, LPE Table 15.0-11 Control Room Table 15.0-14 ESSF Leakage Doses

• Table 15.0-12 Nuclide Release Reference 28 RWST Backloakage Doses Table 15.0-12 Dose Conversion Factors EDE (substit'ete for whole body photon) Reference 29 Skin DS (su'setitute for whole body-beta) Reference 29 Thyroid CDE (substitute for thyroid) Reference 30 control Room Volume (f t3) 1.73E5 control Room Normal Intake (cfa) 500 Control' Room Intake Filter Efficiency 95 (percent)

Control Room Air Pressurisation Rate (cfm) i Control Room Infiltration (cfa) 10 Control Room Isolation Time from start of 1630 event (seconds)

Control Room occupancy 0 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.0 1 - 4 days 0.6 4 - 30 days 0.4 Breathing Rate (m3/sec) offsite gpntrol Room 0 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 3.47E-4 3.47E-4 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.75E-4 3.47E-4 1 - 30 days 2.32E-4 3.47E-4 i

1 2 of 3

.- . - . . . . _ _ _ . . - . - - _ _ - _ . -. _ - _ .-. . - . - . . . . . . . = . . . . . _ . - . -

s BVPS-2 UFSAM Rey, 7 TABLE 15.6-16 RADIOLOGICAL CONSEQUENCES OF A SMALL BREAK LOCA. REN (DLC, 1994 123 Skin DE Thyroid cDE EAS O-2 Hour Dose 2.2E0 1.3E0 2.9E2 LPE Duration Dose 1.3E-1 7.7E-2 2.0E1 Control Room 3.2E-3 3.2E-2 1.1El i

N.

1:

Lided dose valaes repruent 14. Lou,,d,.*3, valuu, whiekhmay 7 e .

W kr th. urred angst, mul6 j 1 of 1 l

O i ATTACHMENT B-1 Beaver Valley Tower Station, Unit No. 2 License Amendment Request No. 120 LOCKED ROTOR DOSE REVISION A. DESCRIPTION OF AMEN') MENT REQUEST The proposed amendment would revise the UFSAR calculated doses to addresc a non-conservative assumption regarding control room emergency pressurization fan flow during the Locked Rotor accident and include new x/Q values in calculating the Exclusion Area Boundcry (EAB) and Low Population Zone (LPZ) doses.

This change is not the result of hardware changes to the plant or a change in op6 rating practices. It is intended to reflect corrected analysis results only and is necessary to allow correction of the licensing basis to reflect conservative assumptions used in the revised dose analysis for a Locked Rotor event.

The UFSAR changes are identified in Attachment A-1.

This amendment is being requested in accordance with 10 CFR 50.59(c) because it has been identified that prior NRC approval is needed to increase the analyzed dose from a Design Basis Accident (DBA).

Table 1 (refer to page B-1-8) is provided showing changes in input parameters used in the analysis for the Locked Rotor Accident. Inputs reflect revised plant parameters as well as l

I revised x/Qs for calculating offsite doses, revised control room occupancy factors to be consistent with the Standard Review Plan (SRP), and model improvements. The X/Qs used are consistent with those applied in supporting analyses for Technical Specification Change No. 115 submitted by [[letter::L-97-045, Application for Amend to License NPF-73,consisting of Change Request 115,modifying TS 3.4.8 by Reducing RCS Specific Activity Limits IAW GL 95-05|letter dated October 22, 1997]].

B. DESIGN BASES The control room habitability system, in accordance with General Design Criterion (GDC) 19 in Appendix A of the Code of Federal Regulations Title 10 Part 50, maintains an environment to ensure that actions can be taken to operate the plant safely under accident conditions, including loss of coolant accidents.

Adequate radiation protection is provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident.

The control room habitability system is designed such that upon receipt of a Phase B containment isolation signal, or high radiation signal from the control room radiation monitors, the outside air dampers are automatically closed and the control room envelope is isolated for 60 minutes. During this period, the

= . _ _ - . - _ . - . - . -. - . _ _ - - . _ _ _ . _ . _ _ . - - - . - - . - - - _ - -

. ATTACHMENT B-1, centinuCd Proposed Licer.se Amendment Request No. 120

, Page 2 control room is pressurized by the compressed air storage bottle system. Aftfr 60 minutes of isolation, one of the two automatic pressurization fans starts and draws outside air through its associated charcoal filtration unit to maintain the control room pressurized for the duration of the event. Unit No. 1 also has a fan pressurization system consisting of two fans and one charcoal filtration unit which can be manually started following the depletion of the compressed dir storage bottle system should the Unit No. 2 automatic system fail to operate. Unit No. 2 does not credit the Unit No. 1 fan pressurization system for maintaining control room habitability.

The control room habitability dose calculations are based on the control room being maintained at a positive pressure by the bottled air system for 60 minutes following a DBA requiring control room isolation. This 60 minute period was based on the time requirement for the containment building to return to a subatmo3pheric condition following a Large Break Loss of Coolant Accident (LBLOCA).

For non-LOCA events, the locked rotor accident is the most severe undercooling event that is analyzed. This event is analyzed in UFSAR Stiction 15.3.3, which notes that the peak cladding-temperature is well below the 2700*F design limit. However, the radiological consequences of a postulated locked rotor accident are analyzed assuming ist failed fuel. The percentage of failed fuel is based on the assumption that all fuel rods that exceed Departure from Nucleate Boiling fail. The primary to recondary system leakage rate is at the technical specification value of 1 gpm. The environmental releases for the locked rotor accident are combined with the atmospheric dispersion values to determine the EAB - and LPZ doses. The methodology used in calculating the offsite doses is discussed in Appendix 15A. The radiological consequences for the locked rotor event do not exceed 10 CFR 100 guidelines.

C. JUSTIFICATION The-proposed revision to.the Control Room Doses are a result of an identified non-conservatism in the radiological dose methodology. Previously, assumption of minimum control 2 emergency pressurization system fan flow was expected to maxicize the calculated control room doses for all DBAs. Recently it was identified that maximizing the system flow assumptionu could increase the calculated control room dose. A maximum emergency pressurization fan flow of 1030 cfm has been selected as an input to the calculations for co.itrol room dose. This flow value is slightly above the technical specification maximum flow of 1000 cfm to include instrumentation uncertainty.

The DBAs which credit isolation of the control room in their calculations of control room dose were re-evaluated to determine whether maximizing or minimizing the control room emergency B-1-2

i

, ATTACHMENT B-1, ctntinu d Prcp;ccd License Amenfmant Request No. 120 Page 3 pressurization system fan flow assumptions after a DBA would increase the calculated control room doses. The most limiting case for _each DBA was citablished for the emergency pressurization flow rates. As a result, the following UFSAR control room calculated doses will be changed:

Locked Rotor Accident

' Control Room f Remi--

Thyroid Gamma Beta Current UFSAR 1.1 1.1E-2 1.5E-1 Revised UFSAR 1.7 1.6E 2.3E-1 As shown, . the results increase slightly; however, they remain l within the limits of GDC 19 for the control room. Since Unit No.

1 and Unit No. 2 share a combined control room, a DBA at either l-Unit will affect the common control room, and hence either Unit's cont.ol room personnel.

The offsite dose calculation assumptions were revised to be consistent with the calculation. assumptions-for the control room

-doses. The following UFSAR offsite calculated doses will also be changed:

Exclusion Area Boundarv (Rem)

Whole Body Beta Thyroid fegama Skin '

Current UFSAR 3.25x1p l

3.41 2.09 Revised UFSAR 3.7 x 10 3.6 2.2 Low Pooulation Zone (Rem)

Whole Body Beta Thyroid faampA Skin Current UFSAR l_

Revised UFSAR 1.44xig 3.48 x 10~1 3.6 x 10-1 2.17 x 10~1 1.6 x 10 2.3 x 10~1 As shown, the EAB and LPZ doses increase slightly. The proposed EAB'and LPZ donas remain less than 10 CFR 100 guidelines.

D. SAFETY ANALYSIS The tenit 1 and Unit 2 - control rooms are located in- a common

-control area and are pressurized by the Control Room Bottled Air Pressurization System _ (CREBAPS) to prevent radioactive inleakage during the first - hour following detection. Following depletion of the-compressed air-aupply, the emergency ventilation subsystem receives a time delayed initiation signal to operate. The emergency ventilation f ans draw air from ' an outside air intake through heaters to - limit the humidity in the . air, and . then through a' filter bank to reduce inhalation dose to the control B-1-3

ATTACHMENT B-1, c:ntinutd Proposed License Amendment Request No. 120 Page 4 room area. This subsystem is capable of maintaining the control area pressurized indefinitely.

The proposed amendment does not change the function, method of operation, or- the technical specification requirements for maintaining the control room habitability system operable. The same level of habitability will be provided. The revision to the dose calculations involves maximizing the fan pressurization system flow rate instead of using the minimum intake flow of 690 cfm. Minimum flow was found to bo a non-conservative assumption in previous analyses. The result of revising this input is a slight increase in the calculated control room dose.

This changa is considered safe because control room habitability system analyzed flow rates have been conservatively chosen and will continue to maintain the control room dose within the limits of GDC 19 during the Locked Rotor acciaent. The dose calculation assum;3 tion af maximum fan flow is conservative since it results

, in hngher control room doses and bounds the range actually permitted by technical specifications. The new analysis results indicate a slight increase in control room dose; however,- the dose remains less than the limits of GDC 19. By continuing to meet the requirements of GDC 19, the control room habitability system will maintain the control room environment habitable to ensure actions can be taken to operate the plant safely under accident conditions.

With respect to calculated changes in offsite dose consequences, increases have not resulted from changes to the plant or its operation. Calculated increases are the result of changes in the calculation inputs, parameters and methodology. No fuel damage is actually predicted for the Locked Rotor accident, but 18% fuel failure was assumed in the ' dose calculations. The increased doses remain less than 10 CFR 100 guidelines.

E. NO SIGNIFICANT HAZARDS EVALUATION The no significant hazard considerations involved with the proposed amendment have been evaluated, focusing on the three standards set forth in 10 CFR 50.92(c) as quoted below:

The Commission may make -

tinal determination, pursuant to the procedures in paragrhph 50.91, that a proposed amendment to an operating licenr.e for a facility licensed under paragraph 50.21(b) or paragraph 50.22 or for a testing facility involves no significant hazards consideration, if operation of the facility in accordanc e with the proposed amendment would not:

(1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or B-1-4

- - -_ -_ _ _ - - - -. _~ . ___ _-

• ATTACHMENT 11el, c0ntinutd Proposed Licenaa Amendment Request No. 120 Page 5 (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

The following evaluation is provided for the no significant hazards consideration standards.

, 1. Does the change involve a significant increase in the probability or conseqvnnees of an accident previously evaluated?

The proposed amendment would revise the calculated control room doses for a Locked tor accident to address a non-conservative assumption for the fan pressurization cystem flow rate. The proposed amendment does not affect the capability of the control room habitability system to maintain control room dose within the limits of General Design Criterion (GDC) 19 in Appendix A of the Code of Federal Regulations Title 10 Part 50. The control room habitability system is an accident mitigation system and will continue to operate as designed. The system has no accident prevention function nor does it interact with systems that have such a function. The proposed change does not alter plant systems, structures or components.

The proposed amendment would also revise calculated offsite doses resulting from a locked rotor accident. This change in doses is not due to physical plant changes, but results mainly from use of more conservative assumptions used in calculating doses.

The proposed change does not affect the manner in which the plant is operated. The physical plant equipment and operating practices are not changed; therefore, the probability or an accident previously evaluated remains unchanged.

The ?arformance requirements of the plant systems which are requnrad to minimize the radiological consequences of a Locked Rotor accident remain unchanged. The proposed change slightly increases calculated control room doses due to an analysis input change for filtration fan flow rate. This slight increase remains below the limits required by GDC 19.

The proposed change does not involve a significant increase in the consequences of an accident previously evaluated since adequate control room radiation protection continues to be provided to ensure actions can be taken to operate the plant safely under accident conditions. The radiological consequences to the environment from a Locked Rotor accident remain unchanged since the performance of plant systems remains unchangcd. Although slightly increased, revised calculated offsite doses remain less than 10 CFR 100 limits. 1 B-1-5

4

, .- ATTACHMENT B-1, cantinued Proposed License Amandas t Request Po. 120 l Page 6 Based on the above discussion, it is concluded that this proposed change does not involve a significant increase in the probability or consequences of an accident previously

' evaluated.

2. Does the change create the possibility of a new or different kind of accident from any accident-previously evaluated?

The proposed change does not alter the method of operating the plant nor does it pose additional challenges to the

i. design or function of the control room habitability system.

The control room habitability system will continue to operate as designed.- -The control -room habitability system will continue to- maintain the control room dose consequences within the limits specified in GDC 19. Adequate control room radiation protection will continue to be provided to ensure

-actions can be taken to operate the plants safely under accident conditions. The proposed change to the control room dose is - only the result of a change in analysis input parameters. Plant performance has not been modified in any way which affects doses to the-public. '

Therefore, the. proposed change does not create the possibility-of a new or different kind of accident from any accident previously evaluated.

3. Does thu' change involve a significant reduction in a margin of safety?

-The slight increase -in calculated control room dose as a restit. of assuming increased. fan flow- does not result in exceeding the' limits prescribed in GDC 19. Calculated doses-to.the public-are slightly increased, but-not as a result of physical changes. The-proposed change will not result-in any additional challenges to plant equipment including - the fuel

-and reactor coolant system pressure boundary since adequate control room radiation' protection will continue to be provided. The control room habitability system will continue-to provide adequate - radiation . protection to ensure actions can be taken to operate the- plant. safely under accident l conditions. The offsite doses increase slightly; however, the calculated dose results remain less than 10 CFR 100 limits. .Therefore, the proposed change does not involve a significant reduction in a margin of safety.

4 B-1-6

. ATTACHMENT B-1, c ntinutd Proposed License Amendment Request No. 120 Page 7 F. No SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Base 6 on the considerations expressed above, it is concluded that l the activities associated with this license amendment request '

satisfy the requirements of 10 CFR 50.92(c) and, accordingly, a no significant hazards consideration finding is justified.

G. ENVIRONMENTAL CONSIDERATION This license amendment request changes the calculated DBA control room dose. It has been determined that this license amendment risquest involves no significant increase in the amounts, and no significant change in the types of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. This license amendment request may change requirements with respect to installation or use of a facility component located within the restricted area; however, the category of this licensing action does not individually or cumulatively have a significant effect on the human environment. Accordingly, this license amendment has no potential environmental impact. The amendment is I

I necessary to allow correction of the licensing basis to reflect conservative assumptions used in the control room dose analysis for design basis accidents. Therefore, possible alternative solutions would have greater potential for environmental effects and have not received serious consideration.

{

B-1-7

. ._- -. . . - - . ~ . - . . - - _ . . . . . ~ . _. - --- . -- - . . - . .

3 4

, .- ATTACHMENT B-1, c0ntinutd Proposed License Amendment Request No. 120 Page 8 Tabla 1 l 4 sed Parameters for

-Unit 2 L ked Rotor Accident Analysis Input / Parameter tinits Previous Revised Reactor Coolant Mass 1bm 4.20E+05 3.51E+05 Steam Generator 1.Lauid Mass- lbm 99300 92000 I Steam Generator Steam Mase ibm 0700 7190 3

CR Pressurisation Rate (Fana) ft / min 690 1030 -

3 CR Pressurisation Rate (Bottles) ft / min 690 0 CR Purge Flow ft /3 min 19800 16900 CR occupancy Factor 1 - 4 day --

1 0.6 CR Occupancy Factor 4 - 30 day --

1 0.4 3

EAR 0 - 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> y/Q sec/m 1.44E-03 1.25:1-03 3

LPE O - 8 hour-x/Q- -sec/m 7.07E-05 6.04E-05 LPE 8 - 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> x/Q sec/m 5.16E-05 4.33E-05 LPE 1 - 4 day y/Q sec/m 2.59E-05 .2.10E-05 3

LPE 4 - 30 day x/Q sec/m 9.36E-06 7.44E-06 Progeny Ingrowth not used. used a

i 4

B-1-8

ATTACHMENT B-2 Beaver Valley Powar Station, Unit No. 2 License Amen 6aelt Request No. 120 REVISED SMALL BREAX Lot A CALCULATION PARAMETERS EM A. DESCRIPTION OF AMENDMENT REQUEST The proposed amendment would revise USFAR Tables 15.0-13, 15.6-15 and 15.6-16 to modify calculation parameters and Section 15.6.5.5 to include editorial changes to ensure that our intentions in describing the Small Break Loss of Coolant Accident (SBLOCA) radiological consequences are clear. The following items in the UFSAR description of the SBLOCA radiological consequences analysis were changed: 1) a new lower minimum control room emergency pressurization fan flow rate and 2) a new lower minimum air bottle discharge rate. Table 1 (refer to page B-2-6) provides a list of specific parameters changed in this analysis.

B. DESIGN BASIS In Rev'ision 7 (April 1995) of the BVPS-2 UFSAR, DLC included the results of an analysis of radiological consequences of a SBLOCA (UFSAR 15.6.5.5). This new information was provided because Westinghouse Nuclear Safety Advisory Letter 93-016 had suggested the possibility that there ceuld be combinations of break size, i fuel damage, and spray actuation delay for which the offsite dose

)

consequences might be greater than those calculated in the large break LOCA analysis. The design basis analysis as defined in Regulatory Guide 1.70 had historically been performed with respect to the large break LOCA for BVPS-2. Even though a SBLOCA was calculated to have a much lower peak fuel clad temperature (PCT) than large break LOCA and Regulatory Guide 1.70 only requests that analyses of radiological consequences be provided for limiting events, an analysis was performed for the purpose of comparing SBLOCA with the design basis analysis results. This J

' analysis was not intended to be the design basis analysis but was a more realistic analysis for comparison purposes only. UFSAR 15.6.5.5 describes key differences between the analysis methodology used for SBLOCA and the design basis analysis. DLC wishes to emphasize that the small break analysis relied to some extent on parameters presented in NUREG 1465. Specific parameters used in evaluating the consequences of a SBLOCA are described in table 15.6-15. Although we do not believe that 10 CFR 50.59 is intended to require prior NRC review for changes to analyses of this nature, the attached amendment request includes revisions to UFSAR 15.6.5.5 and related tables for URC consideration. This is being provided because the methodology used for calculating radiological consequences has not been previously approved by the NRC.

1 l

1 i

p

• ATTACHMENT B-2, continu0d (

,' Proposed License Amendment Request No. 120 i t Page 2 i i  !

. C. "

JUSTIFICATION i

Even though - a SBLOCA was calculated to have a much lower peak j

, ' fuel-clad temperature (PCT) than large_ break LOCA and-Regulatory ,

i Guide 1.70 _ only requested that analyses of radiological i j consequences be provided for limiting events, an analysis has l been provided for the purpose of comparing SBLOCA with the design '

basis analysis. results. The analysis is not the = design basis i analysis.but is a more realistic analysis for comparison purposes  ;

i only. Current PCT analyses show that no fuel damage is-  :

[ postulated.-

fl D. SAFETY ANALYSIS-  ;

, The proposed amendment does not change the function, method of j the _ technical specification requirements for

^

operation, or maintaining the control room habitability system operable. _ The same level of habitability will be:provided. . The revision to the dose calculations--involves reducing the assumed bottle discharge  !'

.- rate and the- fan pressurization system flow. Minimum flow is l' found to be a conservative assumption. The result _of revising the analysis is-a= slight decrease in_the calculatedicontrol' room >

, dose,'however, the current stated . doses in UFSAR - Table _15. 0-L  !

- will be maintained.- Therefore, :the control room dose can be l assured to be within the limits of GDC 19.

• i

, This change is considered safe because control room ~ habitability  ;

system analyzed flow rates have been conservatively chosen and F_ will continue to maintain the control room dose within the limits I of GDC 19 during1the-SBLOCA., The dose calculation assumption:of ,

minimum fan flow is conservative -since- it results in higher control room doses 1 and bounds the range actually permitted by ,

technical specifications. The new analysis results indicate a-  ;

slight ' decrease in control room dose which remains "below the- i l limits-of GDC 19. By continuing-to-meet-the requirements of GDC I l .- ' 19, .the' control room -habitability system will maintain the l control room environment habitable to ensure actions can be taken '

= to operate the plant safely under accident conditions.

E. - NO SIGNIFICANT HAZARDS-EVALUATION ,

, The - no - significunt -hazard considerations involved: with the proposed amendment have- been evaluated, focusing on the tnree standards set forth in 10 CFR_50.92(c)-as quoted belows c

. i The Commission may make a final determination, pursuant to the procedures in paragraph:50.91, that a proposed amendment' to'~an operating license for a facility licensed under  ;

._ paragraph __._50. 21(b) or- paragraph 50.22 or- for. a testing 0 f acility involves no significant hazards consideration, if

operation of - the facility in accordance with the proposed amendment would not

l 4

B-2-2

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~ _ . _ _ - _ . _ _ .

-_._____.7.__ .

ATTACIOtENT B-2, continu0d Proposed License Amendment Request No. 120-Page 3 (1) Involve a signifPant increase in - the probability or consequences et an accident previously evaluated; or

{

(2) Create the possibility cf a new or different kind of i accident from any accident previously-_ evaluated; or

.(3) Involve a significant reduction in a margin of safety.- l The following evaluation is provided for the no significant hasards consideration' standards.

1. Does the change involve a significant increase in the '!

probability or consequences of an accident previously '

evaluated? '

i The proposed amendment would revise the control room dose analysis parameters for a' Small Break Loss of coolant.

Accident (SBLOCA) to include more conservative- assumptions- '

for the pressurization system flow rate. The proposed amendment does not affect the capability of the control room i habitability system to maintain control room dose within the  ;

limits of General.-Design Criterion (GDC) 19 in~ Appendix A of ,

the ' Code of Federal Regulations Title 10 Part . 50. The control room - nabitability ' system is an accident - mitigation i system and will continue to operate as designed. The system' .

has no accident prevention function nor does it interact with systems that.have such a function. The proposed change does not alter plant' systems,. structures or components.

The_ proposed change does not. affect the manner-in which the plant is operated. The physical plant equipment and operating practices are not changed; .therefore, the probability of an accident previously evaluated remains

. unchanged.

The performance requirements of the plant systems which are -

L required to minia:,se the radiological consequences of a

, SBLOCA remain unchanged. The proposed change slightly L decreases calculated control room doses due to analysis input -

L changes. Calculated doses remain below the limits required -

.by GDC 19.

l Based on the above -- discussion, it is ' concluded that this proposed change does . not involve a significant increase in the probability or consequences of an accident previously L evaluated.-

2. Does the change create the possibility of a new or different

_ -kind of accident from'any_ accident previously_ evaluated?

i l'

The proposed' change does not alter the method of operating the plant- nor does it pose additional challenges to the design or f mction of the control room habitability system.

B 2-3.

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_ . . _ . . _ . . . _ . ~ . _ . _ _ . , _ _ . _ . _ _ . . _ . _ _ _ _ _ _ _ _ _ _ _ _ . - - - -.

5 i ,

i.

.. ATTACHMENT B-2, continutd l Proposed License Amendment Request No. 120 i c Page 4 i

'Ihe control room habitability system will continue to operate ,

as designed. The control room - habitability system will >

continue _to maintain the control room dose consequences i within the limits specified in GDC 19. Adequate control room '

. radiation protection will continue to be provided to ensure actions can be taken to operate the plants safely- under i accident conditions. The proposed change to the control room- .

l dose is only a result _ of an analysis being revised. Plant i performance hss not been modified in any way . which af facts-doses to the public.

[

1 i Therefore, the proposed change does not -create the i: possibility of a new or different kind of accident from any ,

l accident previously . evaluated. Although - no new types of-l accidents are created, the analysis represents a new

[ methodology different than any evaluated previously _ by the

NRC.

r

! 3. Does the change involve a significant reduction 'in a margin

! of safety?  ;

i The slight Ldecrease in . calculated --control room dose - as a

result of the revised enalysis does not - result in exceeding ,

the limits prescribed -. in GDC 19. The proposed change 'will not result in any additional challenges to plant equipment f

including the fuel and reactor. coolant system pressure boundary - since adequate control room radiation protection

will continue to be provided.- The - contro.1 room habitability-

system will continue to provide adequate radiation protection j,

to ensure actions can be taken to -operate the plant sately .

under accident conditions.

t

[il F. No SIGNIFICANT HAZARDS CONSIDERATION' DETERMINATION  !

Basodi on-the considerations expressed-above, it-is concluded that .

the - activities associated with this license amendment - request satisfy the requirements of 10 CFR' 50.92 (c) and, accordingly, a

no significant hasards consideration "inding is' justified.

2- t I-

~

i B-2-4

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i ATTACHMENT B-2, 0:ntinued-

-Prcpored Lionn30 Am:ndment Requ Ot N3. 120 Page 5-G. ENVIRONMENTAL CONSIDERATION This license- amendment request modifies calculation parameters and methodology, and includes editorial changes describing the SBLoCA radiological _ consequences. It has been determined that this license amendment request-involves no significant increase in the amounts, and no significant change in the types of any offluonts ' that may be released- offsite, and that there -is no

- significant increase in individual or cumulative occupational--

radiation exposure. This license - amendment request may change requirements component with respect to installation or use of a facility located -within the- restricted area; however, the category of this licensing action does not individually or curulatively have a significant effect on the human environment.

Accordingly, this license amendment- has no -potential environmental impact. The amendment is necessary to - allow correction of the licensing basis ta reflect conservative assumptions used in the. control room dose analysis- for design l basis accidents. Therefore, possible alternative solutions would have greater. potential for environmental effects and have not received serious consideration.

l 4

I

--B-2-5

l

.- ATTACHMENT B-2, c ntinuCd .

Proposed License Amendment Request No. 120 I

'Page 6 Table i Revised Parameters f: 1 Unit 2 Small Break LOCA Aa,41ysis Input / Parameter Unite Previous Revised CR Pressurisation Bottle Flow Rate ft / min 690 I 600--

CR Pressurisation Fan Flow Rate ft / min 690 600 RWST Location X/Q (0 - 8 hr) sec/m 1.20E-4 8.25E-5 8 - 24 hr 5.91E-5 4.07E-3 1 - 4 day 4.45E-5 3.13E-5 4 - 30 day 2.64E-5 1.89E-5 Unfiltered Inflitration Rate efm 100 10 Filter Deposit Radiation flource from RWST mrom Not 0.548 Leakage

• considered Filter Deposit Radiation Source from ECCS mrom 0.88 0.817 Leakage-B-2-6

4 ATTACHMENT C-1 Beaver Valley Power Station, Unit No. 2 License Amendment Roquest No. 120 LOCKED ROTOR DOSE REVISION Attached is the 10 CFR 50.59 evaluation that documunts the unreviewed safety question for the radiological consequences of the Locked Rotor Accident.

i-4

10 CFR 50.59 EVALUATION WORKSHEET (ctntinu;d)

Plant Change er Procedure No.: UFSAR Changes Revision flo.: -

Plant Change or Procedure

Title:

Rev'ew of Cose Calculation Results for the Locked Rotor An~alysis Against UFSAR Listed Bounding Dose Values i

Unit Number: 2 l l

Note: Personnel who initiate changes to the f acility or procedures as described ir the UFSAR SHALL also initiate a UFSAR 7.ange per NPSAP 7.3.

.fWA 11 , .

Mark Duranko Preparer (Print) _ /2 [

Preparer's Signature date '

Steve Nass Independent Rev d'rint)

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22 Si INTRODUCTION

1. Dascribe the plant or procedure change (test or experiment) being evaluated and its expected effects below.

2. Also describe below why the plant or procedure is being changed.

CHANGE DESCRIPTION:

As a result ofissues in Condition Reports 972304 and 972390 involving controlroom habitability,

) BVPS control room dose calculations were re-evaluated. Tine re-evaluation initially involved an identified non-conservative fan flow rate assurrption used in control rocm radiation dose calculations (i.e., maximum versus minimum controlrocm emergency pressurization fan flow rate).

An extent of condition re-review of input assumptions for Control Room doses then identified several other conservative changes which were also factored into the BVPS Unit 2 Control Room dose calculations.

Yhe control room re-evaluation only addressed any DBA and its associated control room dose which credited isolation of the control room during the DBA since the issue is limited to control roon e ,ergency pressurization and filtration system fan flow rate. There are three DBAs in the Unit .* iJFSAR which were recently re-evaluated, since they credit control room isolation. This 10CFR50.59 evaluation only addresses the re-evaluation conducted for the Locked Rotor DBA.

The re-evaluation for the Small Break LOCA and DBA LOCA arc addressed in a separate 10CFR50.59 and th;s 10CFR50.59 draws no conclusions regarding the two above listed DBAs.

. . 10 CFR 50.59 EVALUATION WORKSHEET (continued)

CHANGE DESCRIPTION: (Continued)

UFSAR pages which describe any analysis assumptions will be revised where that input to current BVPS Unit 2 dose calculations was revised. UFSAR T?bles 15.0-12 and 15.013 willbe revised to show the results of the revised Locked Rotor Accident Dose Calculation. t

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I l

1

3. Identify the operating parameters, design parameters, and systems affected by the change.

The Control Room Emergency Air Pressurization and Filtration System us described within BVPS Unit 1 UFSAR section 9.13.4 and within BVPS -1 Technical Specification 3/4. 7. 7.1 and bases.

The Control Room Emergency Air Pressurization and Filtration System as described within BVPS Unit 2 UFSAR section 6,4 and within BVPS 2 Technical Specification 3/4.7. 7 and bases.

The design calculations used to support that the above systems fulfill the General Design Criteria (GDC 19) criteria to provide adequate radiation protection to control room personnel.

4. Identify the credible failure modes associated with the change.

' The credible failure mode considered is the potential failure of a controlroom dose calculatron to adequately meet regulatory calculation criteria and meet GDC 19 limits, using certain assumptions and approved methodology.

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- 10 CFR 50.59 EVALUATl N WORKSHEET (continued)

5. Provide references to the location of information us1d for the safety evaluation.

Unit 2 UFSAR sections 5.4, 6.4, 15.0, 15.3 Unit 2 UFSAR Tables 6.4 1, 15.0-7, 15.0 11, 15.0-12, 15.0-13,15.3 3, 15.3 4 Unit 1 TechnicalFiscifications 3/4.7.7.1 Unit 2 TechnicalSpecifications 3/4.7.7 ERS SFL 06 021, RG 1.145 Short Term Accident X/Q Values for EA8 and LPZ, Unit 1 and ?,

based on 19861995 Data,1996 EM 115712, dated January 5,1998, titled " Minimum ControlRoom Ventilation Fan Flow

' Ra tes ".

NED Letter NND1MNE:7954, dated January 12,1998, " Inputs for CRE9APS (DCP 2306)

RadCon Scoping Analysis" NDINEM:1144, dated April 7,1987 l ERS MPD 91022 "Offsite Dose Consequences Of A Locked Rotor Accident At U* tit 2 With 18% Failed Fuel" ERS MPD 91035 " Assessment of the Doses in the Unit 2 ControlRoom Due To a Lccked Rotor accident At Unit 2 Assuming 18% Failed Fuel" BVPS Unit 2 Technical Specification Amendments No. 46 and 57 EM 115766, dated January 15,1998, titled " Dilution Flow for Health Physics Control Room Dose Calc."

EM 115783, dated January 20,1998, titled." Steam Generator Liquid and Steam Masses" NRC Letter, dated 1/9/98, titled, ."NEl 96-07 Guidelines For 10CFR 50.59 Safety

-- Evaluations" NUREG/CR 2858,1982, titled "PA VAN: An Atmospheric Dispersion Program for Evaluating Design Basis Accidental Releases of Radioactive Materials from Nuclear PGwer Stations" l The following references were reviewed, but determined to not be applicable or have a direct affect on the conclusion drawn within this 50.59 evaluation:

ERS-SFL 93-004, Safety Analysis of Consequences of CR Damper Response Delay at Unit 2

- Accidents.

Unit 1 UFSAR sections 9.13.4,11.3.5 L

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6. Identify specific UFSAR parts (i.e. pages, tables and/or figure numbers) modified or potentially modified by this facility change, procedure change, test or experiment.

Unit 2 UFSAR Pages: 6. 4 8a, 15. 0- 12, 15. 0- 15, 15. 3- 10s, 15. 3- 1 1 Unit 2 UFSAR Tables: 15.0-7, 15.0-11, 15.0-12, 15.0-13, 15.3 3, & 15.3-4 NPDAP 818 Rev 4, Form Rev.: 0701/97 Page3 fo54%s

10 CFR 50.59 EVALUATION WORKSHEET (continued)

PART 1: EFFECT ON DESIGN BASIS (UFSAR) ACCIDENTS 1.A Supporting Information i 1.A.1 Identify the safety SYSTEMS and/or SYSTEMS important to safety affected by the change.

The proposed changes to the BVPS Unit 2 UFSAR sections and tables either address or list the controlroom doses calculated from the Locked Rotor Accident. The proposed:

UFSAR text and table revisions are changes resulting from revisions to the dose calculation which involve the control room pressurization and filtratica systems for (Jnit 1-and Unit 2 as described in the UFSAR and within technicalspecification 3/4. 7. 7 and -

' bases. The proposed UFSAR changes originated from identifying a non-conservative assumption usea in the controlroom dose calculations forlicensing both units regarding the possible control room pressurization fan flow rate - Condition Report 972390. An extent of condition re review ofinput assumptions for Control Room doses then identified several other conservative changes whicti were also factored into the BVPS )

Unit 2 dose calculations. This 10CFR$0.59 evaluation only addresses the re-evaluation - l conducted for the Locked Rotor DBA. :The re-evaluation for the SmallBreak LOCA and DBA LOCA are addressedin a separate 10CFR$0.59 and this 10CFR50.59 draws no conclusions regarding the two above listed DBAs.'

The safety systems are the BVPS Unit 1 and Unit 2 control mom emergency pressurization and filtration systems as describedin No. 3 in the introduction Section.

1.A.2 Discuss the effects of the cliange and/or failure modes associated with the change on th'e probability of failure of the s ? stems identified.

The proposed UFSAR changes originated from identifying a non-conservative assumption ,

used in the controlrocm dose calculation forlicensing both units regarding the possible

. control room pressurization fan flow rate. Previously all fan flow rates used in the calculations were assumed to be 690 cfm even though plant technicalspecifications .

require actual tested flow rates between 800 and 1000 cim. This flow rate was used since minimizing the fan flow was assumed to maximize the dose to personnelin the controlroom. This was describedin section 6.4.4.1 of the U2 UFSAR. Allcaice'ations which modeled the control room emergency pressurization fans were re-evaluated using.

a maximized flow rate (rather than the curren*ly used minimized flow rate) to determine if .

a more limiting (i.e., higher) control room dose would be calculated. Re-running control -

room dose calculations with a maximized control room emergency pressurization flow .

rate (i.e.,1030 cfm) rather than the minimized flow rate (i.e., 690 cfm) resulted in one BVPS Unit 2 dose result where the UFSAR listed values could have needed to be increased, to represent the most ilmiting bounding control room dose case for that DBA.

NPDAP 8.13 Rev . Form Rev.: 07/21/97 Pal

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10 CFR 50.59 EVAL.UATION WORKSHEET (continued)

(Note that the control room habitability system design assumed is the design described .

in the BVPS Unit 2 UFSAR for the Unit 2 ControlRoom Habitability Systems. BVPS Unit 2 recently shutdown due to the inability to completely meet this design. This 50.59 evaluation assumes that the original design as describedin the BVPS Unit 1 Technical Specification 3/4. 7.7.1 bases can be returned for the BVPS Unit 2 control room emergency pressurization system, via a separate 50.59 evaluation (or a different design and 50.59 is provided recognizing the results of this 50.59). The issues involving single failure and the timing / tolerances of controlroom bottle depressurization and controlroom emergency pressuritation fan initiation were resolved via DCP 2306 and DCP 2311 This evaluation makes no new conclusions regarding the effectiveness of the current BVPS Unit 1 or Unit 2 control room habitability systems to conduct their design functions, only that the design can adequately function as currently described in the respective UFSAR, -

such that the generic methodology issua can be addressed on its own merits. No changes sre assumed for the response of the contsvlroom habitability systems nor any changes to current operator responses as described orintended by the current UFSAR descriptions. This 50.59 evaluation is addressing the extent of condition for the non-conservatism of methodology assumptions only.)

The failure mode listed in No. 4 of the Introduction Section addresses the ability of a methodology to be effective. No failure modes are listed which involve the operation or existence of a system, structure or component.

The proposed UFSAR page and table changes support the January 1998 dose calculation revision. These changes ensure that the methodology and assumptions remain valid and

\

the UFSAR description involving control room dose calculation, methodology at:d i assumptions remain correct.

The proposed UFSAR table change: (coming as a result of the prevuous methodology assumption non-conservatism) do not increase the probability of occurrence of a malfunction of equipment important to safety because the changes do not involve equipment. The dose revisions more carrectly show that the design for the current Unit 1 & 2 control room pressurization and filtration system as describedin their current UFSAR continue to adequately operate to meet their design basis that the control room is provided adequate radiation protection.

s-I NPDAP 818 Rev 4, Form Reva 07/21/97 Page5 40 m s

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10 CFR 5@.59 EVALUATION WORKSHEET (continued) 1.A.3 Discuss the effect of the change on the performance of the safety SYSTEMS.

The safety systems (describsd in No. 3 in the introduction Section) involved with the .

control room dose methodology assumptions are unaffected by the proposed UFSAR table changes. This 10CFR50.59 svaluation only addresses a change in the methodology -

assumptions needed as a result of an identified previous non conservatism.

Although the methodology provides a basis for determining the overal! effectiveness of the control room emergen:y pressurization and filtration systems, th9 proposed UFSAR changes do not directly affect the operation of these systems since they do not have a physical constraint orimpact on these systems.-

The proposed UFSAR changes dn not change the performance of the controlroom emergency pressurization and filtration systems as previously described in the current BVPS Unit 2 UFSAR since they do not involve equipment modifications.

1 A.4 Identify the design basis accidents in the UPDATED FINAL SAFETY ANALYSIS REPORT

('JFSAR) to be reviewed for potential impact by the change.

( BVPS-2 UFSAR Table 15.0-13 lists the applicable design basis accidents for whlch controlroom doses are calculated.

Since BVPS Urait I and Unit 2 share a combined controlroom, a design basis accident on .

. either Unit will affect the common control room, and hence either Units' control room personnel. However, Unit 1 design basis accidents are not analyzedin this 10CFR50.59

. evaluation because 1) this evaluation only sddresses the Unit 2 license and UFSAR, and

2) these events have been analyzed under the Unit 1 license in the Unit 1 UFSAR. Unit 1 design basis accidents are not addressed in the Unit 2 UFSAR.

t NPDAp 818 Rev 4. Form Rev : 0741/97 Pqe 6 $0 M M

10 CFR 50.59 EVALUATION WORKSHEET (continued) 1.A.5 Discuss how the parameters and SYSTEMS, affected by the change, affect the assumptions and radiological consequences of the accident (s) identified in Part 1, A.4.

As described in 1.A.2, a minimum control room emergency piessurization systam fan flow rate of 690 cfm was used in all BVPS calculations used to calculate controlroom doses for all DBAs. Recently it was identified that maximizing the actualpressurization system fan flow rete may increase the calculated controlroom dose. As a resuit, all Unit 2 events which credit isolation of the controlroom in their calculation of controlroom dose were re evaluated to determine whether maximizing or minimizing the actual expected controlroom emergency pressurization system fan flow rate after a DBA would increase the calculated control room doses.

NED recently issued EM 115792 which provided new flow rates for the purge fan flow rates. These new values were developed to determine the minimum expected air flow rate for radiologicaldose calculations. These flow values are usedin the purge of the controlroom which occurs between eight and eight and one half hours after isolation of the controlroom. The flow rate for Unit 2 is now 16,900 cim, rather than the 19,800 cfm used previously.

A test was conducted involving the complete discharge of the BVPS CREBAPS bottled air. The test results suggested that the value used in the BVPS Unit 2 dose calculations for minimum possible CREBAPS air bottle discharge flow rate, which was previously \

thought to have been chosen as a conservatively low value, may not have been chosen to be low enough to bound allpossible conditions.

_(The primary intent of the CREBAPS air bottles is to maintain an air flow discharge mto the control room in order to maintain the control room at a positive pressure relative to  ;

' the outside environment (defined to be 1/8" w.c.) such that intrusion of outside environment (assumed to be contaminated by the postulated DBA) is minimized. The Stendard Review Plan requires that dose calculations assumes a constant 10 cfm unfiltered inleakage even when the control room is maintained at its design positive pressure (to model some intrusion due 1:) control room door openings). Thus the calculation always models some levelof radioactivity within the control room envelope.

An unintended but beneficial effect of CREBAPS air bottle dischargs into the control room is the continuing dilution effect of the air mass within the control room. AVith the introduction of radioactivity free airinto the controlroom, the air exiting the cot 'ol room (through the smsll existing leak paths in the controlroom, e.g. under the bottom of doors) will carry out some (small) amount of radioactivity that resides within the control room at that time. If actual CREBAPS bottle discharge flow rate is less than tha minimum discharge flow rate modeled into the dose calculation, the calculated dose rate willnot be maximized. Thus, the modeled minimum CREBAPS bottled air discharge flow should bound all actual credible minimum flow rates possible. (Note: Assuming a perfectly tight control room boundary envelope withos t any leak paths, the cor trol room could be maintained at the designedpositive pressure of 1/8" w.c. with an extremely low bottle discharge rate. The velue assumed must make some 'ssumption regarding the credible, reasonable tightness achievable for ;ne controlroom envelope.}

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10 CFR 50.59 EVALUATION WORKSHEET (continued)

NED evaluated the test data, in addition to the current tightness of the controlroom envelope. A ne v value of 600 cfm was provided by NED (EM 115766) as the minimum credible average value for CREBAPS air bottle discharge flow rate into the controlroom, considering the test data obtained on January 13,1998. (Note: the dose calculation assumes a constant (i.e., average) value throughou: the 60 minutes that the bottled air is modeled to be discharging.) The 600 cfm value represents a value wMch conservatively.

bounds a dose analysis assumption for addressing dilution effect and also a value which willprovide the required 1/8" w.c. This evaluation confirms that use of a 600 cfm value for minimum air bottle flow rate will maintain the assumption used in the BVPS control room dose calculations, which support the BVPS Unit 2 IWSAR dose values.

Recognizing the potential that the dose for the BVPS Unit 2 Locked Rotor Accident analysis could also show an increased control room dose as a result of using a lower CREBAPS bottle flow rate, a higherpressurization fan flow rate,and a_ reducedpurge fan flow rate, the BVPS Unit 2 Locked Rotor Accident analysis was re-enalyzed. The re-analysis used zero (0) cfm for the bottle flow rate (to bound allpossible conditions),

1030 cfm for the pressurization fan flow rate (as provided by EM 115766), and 16,900 cfm for the purge flow rate (as provided in EM 115712!.

An additional analysis input revision involved the value used for reactor cs ra'It system (RCS) mass. Lower RCS mass is conservative by resulting in higher calcu. Med doses. A ,

lower RCS mass will result in a higher RCS concentration for a fixed activity source input. Thus the postulated primary to secondary leakage would occur at a higher concentration, yielding a higher calculated dose. The previous vclue of 4.2E+ 5 lbm (from the UFSAR) was revised to 3.51E+ 5 lbm (NDINEM:1144). The revised numberis  ;

in the range for RCS mass originally provided to SWEC in 1987 for use in BVPS dose \

calculations. The 1987 letterindicated to use 3.90E+ 5 lbm +/ 10%, which l represented 100% power best estimate values, for use in defining the relative masses of '

steam and/or water in the steam gsnerators and RCS. The letter also stated that uncertainties (+/- 10%) should be supplied to the initialmass in the appropriate direction by the dose analyst. RCS mass values usedin severalpast RadiologicalEngineering and SWEC calculation packages has varied from 3.9E+ 5 to 4.2E+ 5 lbm. During the re-analysis of several DBAs performed as part of this extent-of condition review, the RCS mass parameter was noted as having influence on radioactivity released from fuel and transferred through the various compartments of a dose analysis model. Bscause different values were documented, determin tion of a bounding value was deemed to be

,1ecessary.

A 1998 re-assessment of RCS volumes (which accounts for both no steam generator tubes plugged a'1d up to 30% steam generator tubes plugged, the currant limit) ~

supported a best estimate minimum value of 3.51E+ 5 lbm, originating from NDINEM:1144, since it represents a bounding conservative value.

l NPDAP 118 Rev 4. Form Rev : Oh2I/97 Page8 MNS

t 10 CFR 50.59 EVALUATION WORKSHEET (continued)

It was also noted that decreases in the steam generator liquid and steam masses would result in a higher, or more conservative, dose result. The values for steam generator liquid and steam mass were then reassessed to determine if they should be revised to more conservative values. A data set ofliquid and steam masses for various power levels and level of tube plugging was provided by Westinghouse and accepted for use as noted in EM 115783. Bounding values were selected from this data for ucs in dose calulations, it was also decided to use the occupancy factors as shown below to more correctly model the situation.

Thus, the Unit 2 Locked Rotor Accident (LRA) was re-analyzed for the controlroom using the following parameter changes:

CREBAPS flow rate Ocfm Pressurization fan flow rate 1030 cfm Purge fan flow rate 16900 cfm RCS mass 3.51E+ 5 lbm Steem Genera:orliquid mass 92000 lbm Steam Generator steam mass 7190 lbm Occupancy factors ,

0-1 day: 1.0 1 - 4 days: 0.6 4 - 30 days: 0.4 A further change was made in the re-analysis in that the ingrowth of radionuclide progeny (daughterproducts) was included in the final totals. This was done to provide 1 the most conservative result, given the input parameters.

A change was also made in the re-analysis in that the previous calculation included a j contribution from a concurrent iodine spike. The SRP does not address the iodine spike in the section on the locked rotor accident. In additon, there is an assumption of 18%

failed fuelincluded in the calculation.

l 1

l NPDAP 818 Rev 4. Form Rev : 07/21/97 Page9 sos 9wa

10 CFR 50.59 EVALUATION WORKSHEET (continued Since there were parameter changes in the calculation which would affect the dose to the pt ulic offsite, that analysis was also re-performed. In addition to the changes identified above, BVPS has recently developed new X/O values for the EAB and LPZ.

These were developedin 1996 as described in calculation ERS-SFL 96 021 These X/O values are based on BVPS site meteorology data collected from 1/1/86 to 12/31/95.

The values were calculatedpursuant to RG 1.145 and NUREG/CR-2858. These values are:

EAB, 0-2 hours 1.25E-3 sec/ cubic meter LPZ, 0-8 hours 6.04E 5

• LPZ, 8-24 hours 4.33E-5

• LPZ,1-4 days 2.10E 5

• LPZ, 4-30 days 7.44E-6

• The methodology to develop the abnve values was previously describedin a Unit 2 proposed Technical Specification change. This request, number 2A-115, was submitted to the NRC on October 22,1997. This methodology was also submitted to and approved by the NRC for Unit 1 in Technical Specification Amendment No. 205. That analysis, ERS-SFL-96-021, titled "RG 1.145 Short-Term Accident X/Q Values for EAB &

LPZ, Unit 1 and Unit 2, based on 1986-1995 Observations", documents updated Unit 1 and Unit 2 exclusicn area boundary (EnB) and low population zone (LPZ) values. Thus, the X/Q values usedin the re-analysis have been updated based on the 1986 to 1995 .)

data.

With the changes described almve, the calculated doses to controlroom (CR) personnel are:

NEW CR: Thyroid: 1.7 Re:n Gamma: 1.6E-2 Rem Beta: 2.3 E-1 Rem l UFSAR: Thyroid: 1.1 Rem Gamma: 1.1E-2 Rem Beta: 1.5E 1 Rem .

Similarly, the doses to the public o'ffsite are calculated to be:

NEW EAB: Thyroid: 3. 7E + 1 Rem Gamma: 3.6EO Rem Beta: 2.2EO Rem '

UFSAR EAB: Thyroid: 3.25E+ 1 Rem Gamma: 3.41E0 Rem Beta: 2.09EO Rem and NEW LPZ: Thyroid: 1.6E+ 1 Rem Gamma: 3.6E-1 Rem Beta: 2.3E-1 Rem UFSAR LPZ: Thyroid: 1.44E+ 1 Rem Gamma: 3.48E-1 Rem Beta: 2.17E-1 Rem NPDAP 8.18 Rev 4. Form Rev.: 07/21/97 Page10 fo!4wKs

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10 CFR 50.59 EVAL.UATION WORKSHEET (continued)

The above discussion addresses the following factors: (1) an estimated increase in calculated control room dose due to a new lower air bottle discharge rate. (2) an increase in the calculated control room dose due to the increased pressurization f&,1 flow rate, (3) an increase in the controlroom dose due to use of the reduced purge fan flow rate, (4) an increase in the controlroom dose and dose to the public at the EA9 and LPZ due to the use of the more conservative values for other parameters. These factors have resultedin a need to revise the calculated doses to the public at the EAB and LPZIisted '

in UFSAR Table 15.0-12 and the calculated controlroom doses listedin UFSAR Table 15.0-13 to new higher dose values. NRC Guidance published on 1/9/98 states that any issue which would cause the UFSAR listed dose value to increase willbe determined to be an Unreviewed Safety Question. Thus this 10CFR50.59 evaluation causes the determination that an Unreviewed Safety Question exists due to an increase in consequences of an accidentpreviously evaluatedin the UFSAR.

1.A.6 Identify ne design basis accidents,if any, for which failure modes associated with the change can be an initiating event.

BVPS.2 UFSAR Table 15.0-13 lists the applicable design basis accidents for which controlroom doses are calculated.

1.A.7 Discuss the effect of the change on the probability of occurrence of the design basis accidents identified in Part 1, A.6.

1 The design bases accidents descdbedin 1.A.6 are unaffected by the proposed UFSAR table changes. This 10CFF 50.59 evaluation only addresses a change in the methodology assun~ptions needed as a result of an identified previous non-conservatism.

Although the methodology provides a basis for determining the overalleffectiveness of the control room emergency pressurization and filtration systems in meeting GOC 19, the proposed UFSAR changes do not directly affect the probability of an accident since they are an administrative concept and do not have a physical constraint or impact on plant systems.

Thus the proposed UFSAR changes do not increase the probability of occurrence of an accident as previously described in the current BVPS Unit 2 UFSAR.

1.3. Evaluction Questions 1.8.1 Based on Part 1, A.2, MAY the proposed change increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the UFSAR?

O YES S NO 1.8.2 Based on Part 1, A.3, MAY the proposed change increase the consequences of a malfunction of equipment important to safety evaluated previously in the UFSAR?

O YES 9 NO SPDAP 318 Rev 4, Form Reva 07f.'1/97 Page11 Wwu

10 CFR 50.59 EVALUATION WORKSHEET (ccntinued) 1.B.3 Based on Part 1, A.S. MAY the proposed change increase the corwequences of an accident evaluated previously'in the UFSAR?

8 YES O NO 1.B.4 Based on Part 1, A.6 and A 7, MAY the proposed change increase the probability of occurrence of an accident evaluated previously in the UFSAR?

O YES S NO IF THE ANSWER TO ANY OF THE ABOVE QUESTIONS IS YES, THE CHANGE REPRESENTS AN UNREVIEWED SAFETY QUESTION PART 2: POTENTIAL FOR CREATION OF NEW TYPE OF UNANALYZED EVENT 2.A Supporting Information 2.A.1 Based or) Part 1, assess the impact of the change and/or failure modes associated with the change, to determine if the impact has modified the plant response to the point, where it can be considered a new type of accident. Discuss the basis for this determination.

The proposed UFSAR table changes do not impact the plant response since they are based on a methodology assumption concem and do not have a physical constraint or impact on plant systems. This 10CFR50.59 evaluation only adoresses a change in the methodology assumptions needed as a result of an identified previous analyses input .

non-conservatism.

Thus the proposed UFSAR changes do not cause an impact to the plant response to the point where it can be considered a new type of accident different than previously ciescribed in the current BVPS Unit 2 UFSAR. '

2.A.2 Determine if the failure modes of equipment important to safety associated with the change represent a new unanalyzed type of malfunction. Discuss the besis for this determination.

The proposed UFSAR table changes do not impact the plant response since it is based on a methodology assumption concem and does not have a physical constraint orimpact on plant systems. This 10CFR50.59 evaluation only addresses a change in the methodology assumptions 1:seded as a result of an identifiedprevious analyses input non conservatism.

7nus the proposed UFSAR changes do not cause an impact to the plant response to the point where it can create the possibility of a malfunction of equipment important to safety of a different type than previously describedin the current BVPS Unit 2 UFSAR.

2.B Evaluation Questions 2.8.1 Based on Part 2, A.1, MAY the proposed activity create the possibility of an accident of a different type than any evaluated previously in the UFSAR?

O YES 9 NO NPDAP 818 Rev 4, Form Rev 07ntM7 Page12 New

l 10 CFR 50.59 F. VALUATION WORKSHEET (continusd) 2.B.2 Based on Part 2, A.2, MAY the proposed activity create the possibility of a malfunction of I

equipment important to safety of a different type than any evaluated previously in the UFSAR?

O YES S NO IF THE ANSWER TO ANY OF THE ABOVE QUESTIONS IS YES, THE CHANGE REPRESENTS AN UNREVIEWED SAFETY QUESTION PART 3: IMPACT ON THE MARGIN OF SAFETY 3.A Supporting information 3.A.1 ldentify the acceptance limits which form tha licensing basis for the TECHNICAL SPECIFiCATIGNS (i.e., the accident analysis and other design basis) that .:ould be affected by the +.enge.

Unit 2 Technical Specification 3/4.7.7, "ControlRoom Emergency Air Cleanup and Pressurization System."

3.A.2 Discuss the impact of the change on the acceptance limits which form the basis for the TECHNICAL SPECIFICATIONS.

The cal.:ulated control room dose and the dose to the public at the EAB end LPZ for the Locked Rotor Accident has been determined to be increased above the value currently listedin the UFSAR. This was determined to be an Unreviewed Safety Question in Section I.A.5 of this evaluation. There is no chance in plant systems, only methodology assumption changes, resulting in this determination. There is no change in a safety limit or setting proposed by this evaluation. This change is therefore not a change in the acceptance limits or a decrease in the margin of safety.

3.8. Evaluation Question 3.B.1 Based on Part 3, A.1 and A.2, does the proposed activity reduce the margin of safety as defined in the basis for any NICAL SPECIFICATION?

O YES O NO IF THE ANSWER TO ANY OF THE ABOVE QUESY'ONS IS YES THE CHANGE. REPRESENTS AN UNREVIEWED SAFE"Y QUESTION NPDAP 818 Rev 4, Form Reva 07/21,97 Page13 !05ans

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10 CFR 50.59 EVALUATION WORKSHEET (continued)

PART 4: 10 CFR 50.59 EVALUATION CONCLUSION Based on the evaluation in Parts 1,2 and 3 the change:

O Does not involve an UNREVIEWED SAFETY QUESTION.

El involves an UNREVIEWED SAFEETY QUESTION. Contact Safety and Licensing Department before presenting to the Onsite Safety Committee.

PART 5: ENVIRONMENTAL EVALUATION ( Applicable to 8VPS Unit 2) 5.A Supporting Information '

5.A.1 identify any significant increase in any adverse environmental impact previously evaluated in the Final Environmental Statement - Operating License Stage, environmental impact appraisals, or in any decisions of Atomic Safety and Licensing Boaro.

No adverse environmental (non-radiological) impact has been identified. The Unit 2 Environmental Protection Plan addresses the protection of non-radiological environmental values. ~ No aquatic / water quality, terrestrial, or noise issues are affected by this change in radiologicaldose calculations.

5 A.2 Identify any significant change in effluents or power level.

The Locked Rotor Accident was not analyzed in the Environmental Report. The change in radiological dose calculations does not affect effluents or power level.

5.A.3 Identify any matters not previously reviewed and evaluated in the Environmental Protection Plan, Final Environmental Statement - Operating License Stage, or NPDES permit, which MAY hwe a significant adverse environmental impact.

There are no environmentalprotection issues that would involve an unreviewed environmentalquntion or that involve a change to the Environmental Protection Plan.

5.d. Evaluation Question Based upon Part 5, A.1, A.2 and A.3, the change:

Ei5 Does not involve an UNREVIEWED ENVIRONMENTAL QUESTION.

O involves an UNREVIEWED ENVIRONMENTAL QUESTION. Contact Safety and Licensing -

Department before presenting to the Onsite Safety Committee.

NPDAP 818 F.ev 4. Form Rev.: 07/21/97 Page14 5059WKS

o 10 CFR 50.59 EVALUATION WORKSHEET (continu:d)

PART 1: EFFECT ON DESIGN BASIS (UFSAR) ACCIDENTS 1.A' Supporting Intermation 1.A.1 Identify the safety SYSTEMS and/or SYSTEMS importent to safety affected by the change.

The proposed changes to the BVPS Unit 2 UFSAR sections and tables either address or

'list the controlram doses calculated from the Locked Rotor Accident. The prcposed UFSAR text and table revisions are changes resulting from revisions to the dose calculation which involve the controlroom pressurization and filtration systems for Unit 1

' and Unit 2 as describedin the UFSAR and within technica! specification 3/4. 7. 7 and bases. The proposed UFSAR changes originated from identifying a non conservative assumption usedin the controlroom dose calculations forlicensing both units regarding the possible controlroom pressurization fan flow rate - Condition Report 972390. An extent of condition re review ofinput assumptions fcr ControlRoom doses then identified several other conservative changes which were also factoredinto the BVPS Unit 2 dose calculations. This 10CFR50.59 evaluation only addresses the re-evaluation conducted for the Locked Rotor DBA. The re evaluation for the SmallBreak LOCA and DBA LOCA are eddressedin a separate 10CFR50.59 and this-10CFR50.59 draws no conclusions regarding the two above listed DBAs.

4 The safety systems are the BVPS Unit 1 and Unit 2 controlroom emergency pressurization and filtration systems as described in No. 3 in the Introduction Section.

1.A.2 Discuss the effects of the change and/or f ailure modes associated with the change on the probability of failure of the systems identified.

The proposed UFSAR changes originated from identifying a non-conservative assumption used in the control room dose calculation for licensing both units regarding the possible control room pressurization fan flow rate. Previously all fan flow rates used in the calculations were assumed to be 690 cfm even though plant technicalspecifications require actual tested flow rates between 800 and 1000 cim. This flow rate was used since minimizing the fan flow was assumed to maximize the dose to personnelin the controlroom. This was describedin section 6.4.4.1 of the U2 UFSAR. Allcalculations which modeled the control room emergency pressurization fans were re-evaluated using a maximized flow rate (rather than the currently used minimized flow rate) to determine if a more limiting (i.e., higher) control room dose would be calculated. Re-running control room dose calculations with a maximized control room emergency pressurization flow rats (i.e.,1030 cfm) rather than the minimized flow rate (i.e., 690 cfm) resulted in one BVPS Unit 2 dose result where the UFSAR listed values could have needed to be increased, to represent the most limiting bounding controlroom dose case for that DBA.

NPDAP 8 48 Rev 4. Form Rev.: 07/21/97 Page 4 $05GKS

10 CFR 50.59 EVALUATl2N WORKSHEET (centinued)

(Note that the controlroom habitability system design assumed is the design described :

in the BVPS Unit 2 UFSAR for the Unit 2 ControlRoom Habitability Systems.' BVPS Unit 2 recently shutdown due to the inability to completely meet this design. This 50,59 evaluation assumes that the original design as described in the BVPS Unit 1 Technical Specification 3/4.7.7.1 bases can be retumed for the BVPS Unit 2 controlroom emergency pressurization system, via a separate 50.59 evaluation (or a different design and 50.59 is provided recognizing the results of this 50.59). The issues involving single failure and the timing / tolerances of controlroom bottle depressurization and control room emergency pressurization fan initiation were resolved via DCP 2306 and OCP 2311. This evaluation makes no new conclusions regarding the effectiveness of the current BVPS .

Unit 1 or Unit 2 control room habitability systems to conduct their design functions, only that the design can adequately function as currently describedin the respective UFSAR, such that the generic methodology issue can be addressed on its own merits. No changes are assumed for the response of the controlroom habitability systems nor any ,

changes to current operator responses as described or intended by the current UFSAR descriptions. This 50.59 evaluation is addressing the extent of condition for the non-conservatism of methodology assumptions only.}

The failure mode listed in No. 4 of the Introduction Section addresses the ability of a

methodology to ' effective. No failure modes are listed which involve the operation or existence of a system, structure or component..

The proposed UFSAR page and table changes support the January 1998 dose calculation revision.~ These changes ensure that the methodology and assumptions remain valid and the UFSAR description involving comrol room dose calculation, methodology and assumptions remain correct.

- The proposed UFSAR table changes (coming as a result of the previous methodology assumption non-conservetism) do not increase the probability of occurrence of a malfunction of equipment important to safety because the changes do not involve

- equipment. . The dose revisions more correctly show that the design for the current Unit 1 & 2 control room pressurization and filtration system as described in their current UFSAR continue to adequately operate to meet their design basis that the controlroom is provided adequate radiation protection.

NPDAP 415 Rev 4. Form Rev; 07/21/97 Page5 505 %

10 CFR 50.59 EVALUATION WORKSHEET (continu:d) i 1.A.3 Discuss the effect of the change on the performance of the safety SYSTEN. ..

The safety systems (described in No. -3 in the Introduction Section) involved with the control room dose methodology assumptions are unaffected by the proposed UFSAR table changes. This 10CFR50.59 evaluation only addresses a change in the methodology assumptions needed as a result of an identified previous non-conservatism.

Although the methodology provides a basis for determining the overall effectiveness of the control room emergency pressurization and filtration systems, the proposed UFSAR .

changes do not directly affect the operation of these systems since they do not have a physical constraint or impact on these systerns.

-The proposed UFSAR changes do not change the performance of the control room emergency pressurization and filtration systems as previously described in the current BVPS Unit 2 UFSAR since they do not involve equipment modifications.

I I

1.A.4 Identify the design basis accidents in the UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) to be reviewed for potential impact by the change.

BVPS 2 UFSAR Table 15.0-13 lists the applicable design basis accidents for which controlroom doses are calculated.

Since BVPS Unit 1 and Unit 2 share a combined controlmom, a design basis accident on either Unit will affect the common control room, and hence either Units' controlroom

. personnel. However, Unit 1 design basis accidents are not analyzedin this 10CFR50.59 evaluation because 1) this evaluation only addresses the Unit 2 license and UFSAR, and

2) these events have been analyzed under the Unit 1 license in the Unit 1 UFSAR Unit 1 design basis accidents are not addressedin the Unit 2 UFSAR.

NPDAP 818 Rev 4. Form Rev.' S7/21/97 Page6 MN t .

10 CFR 50.59 EVALUATION WORKSHEET (continued) 1.A.5 Discuss 5)w the parameters and SYSTEMS, affected by the change, affect the assumptions and radiological consequences of the accident (s) identified in Part 1, A.4.

As described in 1.A.2, a minimum controlroom emergency pressurization system fan flow rate of 690 cfm was used in all BVPS calculations used to calculate control room doses for all DBAs. Recently it was identified that maximizing the actualpressurization system fan flow rate may increase the calculated controlroom dose As a result, all Unit 2 events which credit isolation of the controlroom in their calculation of controlroom dose were re evaluated to determine whether maximizing or minimizing the actual expected controlroom emergency pressurization system fan flow rate after a DBA would increase the calculated control room doses.

NED recently issued EM 115792 which provided new flow rates for the purge fan flow rates. These new values were developed to determine the minimum expected air flow rate for radiological dose calculations. These flow values are used in the purge of the controlroom which occurs between eight and sight and one half hours after isolation of l

the control room. The flow rate for Unit 2 is now 16,900 cim, rather than the i9,800 cfm used previously.

A test was conducted involving the complete discharge of the BVPS CREBAPS bottled

' air. The test results suggested that the value used in the BVPS Unit 2 dose calculations for minimum possible CREBAPS air bottle discharge flow rate, which was previously thought to have been chosen as a conservatively low value, may not have been chosen to be low enough to bound allpossible conditions.

(The primary intent of the CREBAPS air bottles is to maintain an air flow discharge into the control room in order to maintain the controlroom at a poritive pressure relative to the outside environment (defined to be 1/8" w.c.) such that intrusion of outside environment (assumed to be contaminated by the postulated DBA) is minimized. The Standard Review Plan requires that dose calculations assumes a constant 10 cfm unfiltered inleakage even when the control room is maintained at its design positive pressure (to model some intrusion due to control room door openings). Thus the calculation always models some levelof radioactivity within the controlroom envelope.

An unintended but beneficial effect of CREBAPS air bottle discharge into the control room is the continuing dilution effect of the air mass within the controlroom. With the introduction of radioscrivity free air into the controlroom, the air exiting the control room (through the small existing leak paths in the controlroom, e.g. under the bottom of doors) will carry out some (small) amount of radioactivity that resides within the control room at that time. If actual CREBAPS bottle discharge flow rate is less than the minimum discharge flow rate modeled into the dose calculation, the calculated dose rate willnot be maximized. Thus, the modeled minimum CREBAPS bottled air discharge flow should bound all actual credible minimum flow rates possible. (Note: Assuming a perfectly tight control room boundary envelope without any leak paths, the control room could be maintained at the designed positive pressure of 1/8" w.c. with an extremely low bottle discharge rate. The value assumed must make some assumption regarding the credible, reasonable tightness achievable for the controlroom envelope.}

NPDAP $ 18 Res 4. Form Reva 07/21/97 Page7 Wb

4 i 10 CFR 50.59 EVALUATION WORKSHEET (ctntinuM NED evaluated the test data, in addition _ to the current tightness of the controlroom envelope. A new value of 600 cfm was provided by NED (EM 115766) as the minimum credible average value for CREBAPS air bottle discharge flow rate into the control room, considering the test data obtained on January 13,1998. (Note: the dose calculation assum9s a constant (i.e:, average) value throughout the 60 minutes that the bottled air is modeled to be discharging.] The 60') cfm value represents a value which conservatively bounds a dose analysis assumption for addressing dilution effect and also a value which willprovide the required 1/8" w.c. This evaluation confirms that use of a 600 cfm value

'for minimum air bottle flow rate willmaintain the assumption used in the BVPS control  ;

room dose ca!culations, which support the BVPS Unit 2 UFSAR dose values.-

Recognizir'g the potential that the dose for the BVPS Unit 2 Locked Rotor Accident anclysis could also show an increased control room dose as a result of using a lower -

CREBAPS bottle flow rate, a higherpressurization fan flow rate,and a reduced purge fan l_ flow rate, the BVPS Unit 2 Lockr2 Rotor Accident analysis was re analyzed. The re-(

analysis used zero (0) cfm for ths cottle flow rate (to bound allpossible conditions),

1030 cfm for the pressurization fan flow rate las provided by EM 115766), and -16,900 cfm for the purge flow rate (as provided in EM 115712).

An additional analysis input revision involved the value used for reactor coolant system .

(RCS) mass. Lower RCS mass is conservative by resulting in higher calculated doses. A lower RCS mass will result in a higher RCS concentration for a fixed activity source input. Thus the postulatedprimary to secondary leakage would occur at s higher concentration, yielding a higher calculated dose. The previous value of 4.2E+ 5 lbm (from the UFSAR) was revised to 3.51E+ 5 lbm (NDINEM:1144). The revised numberis in the range for RCS mass originally provided to SWEC in 1987 for use in BVPS dose calculations.- The 1987 letterindicated to use 3.90E+5 lbm +/- 10%, which' represented 100% power best estimate values- for use in defining the relative masses of steam and/or water in the steem generators and RCS. The letter also :tated that

. uncertainties (+/-' 10%) should be supplied to the initial msss in the appropriate direction by the dose analyst. RCS mais values usedin severalpast RadiologicalEngineering and SWEC calculation packages has varied from 3.9E+5 to 4.2E+5 lbm. During the re-analysis of several DBAs performed as part of this extent-of condition review, the RCS mass parameter was noted as having influence on radioactivity relcased from fuel and transferred through the various compartments of a dose analysis model. Because different valuss were documented, determination of a bounding value_ was deemed to_be necessary.

A 1998 re-assessment of RCS volurnes (which accounts for both no steam generator

. tubes plugged and up to 30% steam generator tubes plugged, the current limit) suoported a best estimate minimum value of 3.51E+ 5 lbm, originating from NDINEM:1144, since it represents a bounding conservative value.

- SPDAP S 18 Rev 4. Form Reva 07/21/97 Page8 !0NW

10 CFR 50.59 EVALUATION WORKSHEET (continued)

It was also noted that decreases in the steam oenerator liquid and steam masses would result in a higher, or more conservative, dose result. The. values for steam generator liquid and steam mass were then reassessed to determine if they should be revised to more conservative values. A data set of liquid and steam masses for various power levels and level of tube plugging was provided by Westinghouse and accepted for use .ss i noted in EM 115783. Bounding values were selected from this data for use in dose calulations.

It was also decided to use the occupancy factors as shown below to more correctly model the situation.

Thus, the Unit 2 Locked Rotor Accident (LRA) was re-analyzed for the control room using the following parameter changes:

CREBAPS flow rate Ocfm Pressurization fan flow rate 1030 cfm Purge fan flow rate 16900 cfm RCS mass 3.51E+ 5 lbm Steam Generator liquid mass 92000 lbm Steam Generator steam mass 7190 lbm Occupancy factors ,

O - 1 day: 1.0 1 - 4 days: 0.6 4 - 30 days: 0. 4 A further change was made in the re analysis in that the ingrowth of radionuclide progeny (daughter products) was included in the final totals. This was done to provide the most conservative result, given the input parameters.

A change was also made in the re-analysis in that the previous calculation included a r contribution from a concurrent iodine spike. The SRP does not address the iodine spike in the section on the locked rotor accident. In additon, there is an assumption of 18%

failed fuelincludedin the calculation.

NPDAP 818 Rev 4. Form Reva 07/2t/97 Page9 505m Ks

10 CFR 50.59 EVALUATION WORKSHEET (continued) 3 Since there were parameter changes in the calculation which would affect the dose to the public offsite, that analysis was also re-perfctmed.- In addition to the changes identified above, BVPS has recently deveicped new X/Q values for the EAB and LPZ.

These were developedin 1996 as describedin c?culation ERS SFL 96 021. These X/O values are based on BVPS site meteorology data c0!!ected from 1/1/86 to 12/31/95.

The values were calculated c trsuant to RG 1.145 and NUREG/CR 2858. These values are:

EAB, 0-2 hours 1.25E 3 sec/ cubic meter  !

LPZ, 0-8 hours 6.04E 5

• LPZ, 8 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 4.33E-5 " -

LPZ,14 days 2.10E 5 "

LPZ, 4 30 days 7.44E-6 *

. The methodology to develop the above values was previously describedIn a Unit 2 proposed Technical Specification change. This request, number 2A 115, was submitted to the NRC on October 22,1997. This methodology was also submitted to and approved by the NRC for Unit 1 in Technical Specification Amendment No. 205. That

)- analysis, ERS-SFL 93-021, titled "RG 1.145 Short-Term Accident X/Q Values for EAB &

LPZ, Unit I and Unit 2, based on 1986-1995 Observations" documents updated Unit ! ,

and Unit 2 sxclusion area boundary (EAB) and low population zone (LPZ) values. Thus, the X/Q values usedin the ra-analysis have been updated based on tim 1986 to 1995 data.

With the changss described above, the calculated doses to controlroom 101) corsonnel are:

NEW CR: Thyroid: 1.7 Rem Gamma: 1.6E 2 Rem Beta: 2.3 E-1 Rem UFSAR: Thyroid: 1.1 Rem Gamma: 1.1E-2 Rem Beta: 1.5E-1 Rem Similarly, the doses to the public o'ffsite ..re calculated to be:

NEW EAB: Thyroid: 3. 7E+ 1 Rem Gamma: 3.6E0 Rem Beta: 2.2E0 Rem UFSAR EAB: Thyroid: 3.25E+ 1 Rem Gamma: 3.41EO Rem Beta: 2.09E0 Rem and NEW LPZ: Thyroid: 1.6E+ 1 Rem Gamma: 3.6E-1 Rem Beta: 2.3E 1 Rem UFSAR LPZ: Thyroid: 1.44E+ 1 Rem Gamma: . 3.48E-1 Rem Beta: 2.17E 1 Rem NPDAP 8 '.8 Rev 4. Form Rev.: 07/21/97 Page 10 soso n s

9. i 10 CFR 50.59 EVALUATION WORKSHEET (c:ntinued)

The above discussion addresses the following factors: (1) an estimated increase in calculated control room dose,due to a new lower air bottle discharge rate. (2) an increase in th a calculated control room dose due to the increasedpressurization fan flow rate, (3) an increase in the controlroom dose due to use of the reducedpurge fan flow rate, (4) an increase in the control room dose and dose to the public at the EAB and LPZ due to the use of the more conservative values for otherparameters. These factors have resulted in a need to revise the calculated doses tc the public at the EAB and LPZ listed in UFSAR Table 15.0-12 and the calculated contro! room doses listedin UFSAR Table 15.0-13 to new higher dose values. NRC Guidance published on 1/9/98 states that any issue which would cause the UFSAR listed Jose value to increase willbe determined to be an Unreviewed Safety Question. Thus this 10CFR50.59 evaluation causes the determination ti:nt an Unreviewed Safety Question exists due to an increase in consequences of an accident previously evaluatedin the UFSAR.

1.A.6 Identify the design basis accidents, if any, for which failure modes associated with the change can be an initiating event.

BVPS 2 UFSAR Table 15.0-13 lists the applicable design basic accidents for which .

control room doses are calculated.

! 1.A.7 i ' cuss the effect of the change on the probability of occurrence of the design basis

6. -dents identified in Part 1, A.6.

The design bases accidents describedin 1.A.6 are unaffected by the proposed UFSAR table changes. This 10CFR50.59 evaluation only addresses a change in the mi

• odology assumptions needed as a result of an identifiedprevious non-conservatism.

\

Although the methodology provides a basis for determining the overall effectiveness of the control room emergencypressurization and filtration systems in meeting GDC 19, the proposed UFSAR *hanges do not directly affect the probability of an accident since they are an administra ' ve concept and do not have a physical constraint or impact on plant systems.

Thus the proposed UFSAR changes do not increese the probability of occurrence of an accident as previously described in the current BVPS Unit 2 UFSAR.

1.B. Evaluation Questions 1.B.1 Based on Part 1, A.2, MAY the proposed change increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the UFSAR?

O YES S NO 1.B.2 Based on Part 1, A.3, MAY the proposed change increase the consequences of a malfunction of equipment important to safety evaluated previously in the UFSAR?

O YES S NO )

NPDAP 8 i& Rev 4. Form Rev/. 07/397 Page11 505Ms

-_ __a

.- - - - - -. ,. . ..- _ _ . . - - - . - - ~ ~ . - . -

10 CFR 50.59 EVALUATION WORKSHEET (continued)

(

1.B.3 Based on Part 1, A 5, MAY the proposed change increase the consequences of an accident evaluated previously'in the UFSAR? l i

0 YES O NO i i 1.B.4 - Based on Part'1, A.6 and A.7, MAY the proposed change increase the probability of occurrence of an accident evaluated previously in the UF3AR?

O YES O NO 4

IF THE ANSWER TO ANY OF THE ABOVE QUESTIONS IS YES, THE CHANGE REPRESENTS AN UNREVIEWED SAFETY QUESTION

' PART 2: POTENTIAL FOR CREATION OF NEW TYPE OF UNANALY2ED EVENT 2.A Supporting Information

~ 2.A.1 ; Based on Part 1, assess the impact of the change and/or failure modes associated with the change, to determine if the impact has modified the plant response to the point where it can be considered a new type of accident. Discuss the basis for this determination.

The proposed UFSAR table changes do not impact the plant response since they are based ca a methodology assumption concem and do not have a physical constraint or impact on plant systems. This 10CFR50.59 evaluation only addresses a change in the methodology assumptions needed as a result of an identifiedprevious analyses input non conservatism.

Thus the proposed UFSAR changes do not cause an impact to the plant response to the point where it can be considered a new type of accident different than previously described in the current BVPS Unit 2 UFSAR.

2.A.2 Determine if the failure modes of equipment important to safety associated with the change represent a new unanalyzed type of malfunction. Discuss the basis for this determination.

The proposed UFSAR table changes do not impact the plant response since it is based on a methodology assumption concern and does not have a physical constraint or impact on plant systems. This 10CFR50.59 evaluation only addresses a change in the methodology assumptions needed as a result of an identified previous analyses input non-conservatism.

Thus the proposea UFSAR changes do not cause an impact to the plant response to the point where it can create the possibility of a malfunction of equipment important to safety of a different type than previously describedin the current BVPS Unit 2 UFSAR.

2.B L Evaluation Questions 2.8.1 . Based on Part 2, A.1, MAY the proposed activity create the possibility of an accident of a different type than any evaluated previously in the UFSAR?

O YES S NO NPDAP R 18 RevA Form Rev.: 07n t/97 Page 12 so.w u

1:

10 CFR 50.59 EVALUATl2N WORKSHEET (continued) 2.8.2 Based on Part 2, A.2, MAY the proposed activity create the possibility of a malfunction of equipment important to safety ( ' a different type than any evaluated previously in the UFSARt O YES E NO IF THE ANSWER TO ANY OF THE ABOVE QUESTIONS IS YES, THE CHANGE REPRESENTS AN UNREVIEWED SAFETY QUESTION PART 3: IMPACT ON THE MARGIN OF SAFETY 3.A Support ,g Information 3.A.1 Identify the acceptance limits which form th'd licensing basis for the TECHNICAL SPECIFICATIONS (i.e., the accident analysis and other design basis) that cuuld be affected by the change.

Unit 2 Technical Specification 3/4. 7. 7, " Control Room Emergency Air Cleanup and Pressuriza* ion System."

3.A.2 Discuss the impact of the change on the acceptance limits which form the basis for the TECHNICAL SPECIFICATIONS.

The calculated control room dose and the dose to the public at the EAB and LPZ for the Locked Rotor Accident has been determined to be increased above the value currently listed in the UFSAR. This was determined to be an Unreviewed Safety Question in Section 1.A.5 of this evaluation. There is no change in plant systems, only methodology assumption changes, resulting in this determination. There is no change in a safety limit or setting proposed by this evaluation. This change is therefore not a change in the acceptance limits or a decrease in the margin of safety.

3.8. Evaluation Question ,

3.B.1 Based on Part 3, A.1 and A.2, does the proposed activity reduce the margin of safety as defined in the bevis for any TECHNICAL SPECIFICATION?

)

O YES 8 NO IF THE ANSWER TO ANY OF THE ABOVE QUESTIONS IS YES THE CHANGE REPRESENTS AN UNREVIEWED SAFETY QUESTION NPDA( 818 Rev 4, Form Reva 07/21/97 Page 13 foso n s

f 10 CFR 50.59 EVALUATION WORKSHEET (continued)

PART 4: 10 CFR 50.59 EVALUATION CONCLUSION Based on the evaluation in Parts 1,2 and 3 the change:

O Does not involve an UNREVIEWED SAFEYY QUESTION.

O involves an UNREVIEWED SAFETY QUESTION. Contact Safety and Licensing Department before presenting to the Onsite Safety Committee.

PART5: ENVIRONMENTAL EVALUATION (Applicable to BVPS Unit 2) 5.A Supporting Information '

5.A.1 Identify any significant increase in any adverse environmentalimpact previously evaluated l

in the Final Environmental Statement Operating License Stage, environmental impact appraisals, or in any decisions of Atomic Safety and Licensing Board.

No adverse environmental (non-radiological) impect has been identified. The Unit 2 Environmental Protection Plan addresses the prctection of non-radiological environmental values. No aquatic / water quality, terrestrial, or noise issues are affected by this change in radiological dose calculations.

5.A.2 Identify any significant change in effluents or power level.

The Locked Rotor Accident was not analyzed in the Environmental Report. The change in radiological dose calculations does not affect effluents or power level.

5.A.3 Identify any matters not previously reviewed and evaluated in the Environmental Protection Plan, Final Environmental Statement - Operating License Stage, or NPDES permit, which MAY have a significant adverse environmentalimpact.

There are no environmenalprotection issues that wouldinvolve an unreviewed .

onvironmentalquestion or that invelve a change to the EnvironmentalProtection Plan. \

l 5.B. Evaluation Question Based upon Part 5, A.1, A.2 and A.3, the change:

O Does not involve an UNREVIEWED ENVIRONMENTAL QUESTION.

O Involves an UNREVIEWED ENVIRONMENTAL QUESTION Contact Safety and Licensing Department before presenting to the Onsite Safety Committee.

NPDAP 818 Rev 4, Form Rev.: 07/21/97 Page14 $059WKS

ATTACHMENT C-2 Beaver Valley Power Station, Unit No. 2 License Amendment Request No. 120 REVISED SMALL BREAK LOCA' CALCULATION PARAMETERS Attached is tne 10 CFR 50.59 evaluation that documents the unreviewed safety question involving the calculation methodology for the Small Break LOCA.

I 10 CFR 50.59 EVALUATION WORKSHEET Plant Change or Procedure No.: UFSAR Changes Revision No.: -

Plant Cm. .ge or Procedure

Title:

Review of Dose Calculation Results for the Small Break LOCA Against UFSAR Listed Values Unit Number: 2 Note: Personnel who initiate changen to the f acility or procedures as described in the UFSAR SHALL also initiate a UFSAR uange per NPDAP 7.3.

Steve Nass Preparer (Print) /r re Si ture

([73[f8 Date Mark Duranko /) /

Independent Rev. (Print) Independent viewer's Signature Date Jesus Arias ( }nIa4 \. "2. b9 %

Department Approval (Print) Depa ment Approval Signature Date OSC Concurrence: Meeting Number BV-OSC- -

96 Date / ~M S '

b :_ 2f ,

R.L,l = L Q.aea k~>

l W\ 'y Q INTRODUCTION

1. Describe the plant or procedure change (test or experiment) being evaluated and its expected effects below.

2. Also describe below why the plant or procedure is being changed. -

CHANGE DESCRIPTION:

1 As a result of issues in Condition Reports 972304 and 972390 involving control room habitability, BVPS control room dose calculations were re-evaluated. The re-evaluation initially involved an identified non-conservative fan flow rate assumption used in control room radiation dose calculations (i.e., maximum versus minimum control room emergency pressurization fan flow rate).

An extent of condition re-review of input assumptions for control room doses then identified several other conservative changes which were also factored into the BVPS Unit 2 control room dose calculations.

The control room re-evaluation only addressed any DBA and its associated control room dose which credited isolation of the control room during the DBA since the issue is limited to control room emergancy p.assurization and filtration system fan flow rate. There are three DBAs in the Unit 2 UFSAR which were recently re-evaluated, since they credit control room isolation. This 10CFR50.59 evaluation only addresses the re-evaluation conducted for the Small Break LOCA.

The re-evaluation for the Large Break LOCA DBA and Locked Rotor DBA are addressed in separate 10CFR50.59's and this 10CFR50.59 draws no conclusions regarding the two above listed DBAs.

. _ . _ . . . ~ . . _ - _ __.___..___--__.-._m__ _ . _ _

_e l l

U. 110 CFR 50.59 EVALUATION WORKSHEET (continu:dl  !

t CHANGE DESCRIPTIOM: (Continued)

Radiological consequences of a SmbH Break LOCA was analyzed in response to Westinghouse i

Nuclear Safety Advisory Letter (NSAL)93-016. NSAL 93-016 suggested;the possibility that the

( containment pressure during a SBLOCA may not reach the -setpoint required to actuate the containment sprays.. Containment- sprays are designed to - remove iodine ' released to the

~

Containment, and are credited in the radiological consequence analysis for the large Break LOCA

.DBA. If source terms equivalent to those assumed for a large Break LOCA DBA are used, without taking credit for iodine scrubbing of the containment sprays, offsite dose consequences might be.

greater for a SmaH Break LOCA than for the large Break LOCA DBA. SmaH Break LOCA is not a-limiting event (i.e. SnbaH Break LOCA has a much lower peak clad temperature (PCT) than large Break LOCA), therefore the ; methodology and assumptions used in the SmaH Break LOCA ,

radiological consequence analyses are not required to conform ' with the guidance given in

' Regulatory Guide 1.4. The methodology used for the SmaH Break LOCA radiological consequence analysis was included in revision 7 (April 1995) of the BVPS 2 UFSAR section;15.6.5.5. The methodology howeve.- had not previously been reviewed and approved by the NRC. ' -

UFSAR pages and tables which describe SmaH Break LOCA dose analysis assumptions wiH be revised or clarified where necessary. UFSAR Table 15.0-13 wiH be revised to show the results of the revised SmeH Break LOCA Dose Calculation. .

4

3. = identify the operating parameters, design parameters, and systems affected by the change.

The Control Room Emergency Air Pressurization and Filtration System as described within =

BVPS Unit 1 UFSAR section 9.13.4 and within BVPS -1 Technical Specification 3/4.7.7.1 l and bases.

The Control Room Emergency Air Pressurization and filtration System as described within BVPS Unit 2 UFSAR section 6.4 and within BVPS 2 Technicd Specification 3/4.7.7 and bases.

The design calculations used to support that the above systems fulfin the GeneralDesign Criteria (GDC 19) criteria to provide adequate radiation protection to control room personnel.

The following items in the SmaH Break LOCA Analysis calculation were affected: 1) a new lower minimum control room emergency pressurization fan flow rate, 2) a new lower minimum air bottle discharge rate, 3) use of a specific X/O value for the RWST, 4) a correction for the assumed continuous unfiltered control room inleakage, and 5) adding in a c smau controlroom gamma dose contribution fxm activity deposited on the controlroom

' emergency ventilation system filters.

4. Identify the credib'le f ailure modes associated with the change.

The credible failure mode consideredis the potential failure of a controlroom dose

calculation to adequately meet regulatory calculation criteria and meet GDC 19 limits and l 10CFR100 limits, using certain assumptions and approved methodology.

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10 CFR 50.59 EVALUATl:N WORKSHEET (continued)

5. Provide references to the location of information used for the safety evaluation.

Unit 2 UFSAR sections 5.4, 6.4,12.2.1.3, 16.0, 15.3 Unit 2 UFSAR Tables 15.0 12, 15.0 13,15.0-14, Unit 1 TechnicalSpecifications 3/4.7.7.1 Unit 2 TechnicalSpecifications 3/4.7.7 ERS SFL 92 020, ControlRoom X/O Values the Beaver Valley Power Station (NUS EPC 91 1025)

NUREG/CR 5055, dated 1988, " Atmospheric Diffusion for ControlRoom Habitability Assessments" NED Letter # ND1MNE:7954, dated 1/12/98, 'Inpu:s for CREBAPS (DCP 2306) RadCon Scoptag Analysis" ERS S&W 92-012 " Doses to BV1/BV2 ControlRoom, EAB & LPZ Due to Release from the U2 RWST via ECCS Leakage following a LOCA" ERS SFL 94-014 " Safety Analysis of the Radiologic Consequences of a Smal! Break LOCA at U2 with Delayed CNMT Spray Actuation and NUREG 1465 Core Damage Progression' BVPS Unit 2 Technical Specification Amsndment No. 46 and 57 NPD3SHP:1913, dated October 11,1993, titled "U2 Amendment 57" EM 115766, dated 1/15419, Qilution Flow For Health Phy:..,;s Control Room Dose Calcs. "

EM 115790, dated 1/21/98, "Possible Effects on Dilution Flow Assumptions for Health Physics ControlRoom Dose Calculations Resulting from DCP 2306".

Stone & Webster letter dated January 27,1998, "BVPS2 Control Room Habitability Assessment loss Of Coolant Accident" Stone & V'ebsterletter dated 12/30/97, 'BVPS 1 & 2 Col.trolRoom Habitability Assessment loss of Coolant Accident" The following references were reviewed, but determined to not be applicable or have a direct affect on the conclusion drawn within this 50.59 evaluation:

ERS SFL 93 004, Safety Analysis of Consequences of CR Damper Response Delay at Unit 2 Accidents.

Unit 1 UFSAR sections 9.13.4, 11.3.5

6. Identify specific UFSAR parts (i.e. r,eges, tables and/or figure numbers) modified or potentially modified by this facility change, procedure change, test or experiment.

Unit 2 UFSAR Section 15.6.5.5 Unit 2 UFSAR Tables 15.0 12, 15.0-13, 15.0-14, 15.6 15, i6.6 16, 15.6-17 NPDAP 818 R. .... Form Rev.: 07nl/97 Page3 5059us

10 CFR 50.59 CVALUATION WERKSHEET (continued)

PART 1: EFFECT ON DESIGN BASIS (UFSAR) ACCIDENTS

{

t 1.A Supporting Information 1

1.A.1 Identify the safety SYSTEMS and/or SYSTEMS important to safety affected by the change.

The proposed changes to the BVPS Unit 2 UFSAR sections and tables either address or list the LPZ, EAB or control room doses calculated from a Small Break LOCA. The  ;

proposed UFSAR text and table revisions are changes resulting from revisions and 1 clarifications to the dose calculation which involve the controlroom pressurization and filtration systems for Urit 1 and Unit 2 as describedin the UFSAR and within technical specification 3/4. 7. 7 and bases. The proposed UFSAR changes originated from identilying a non consewative assumption used in the co' trol room dose calculations for licensing both units regardine the possible controlroom pressurizat!on fan flow rate -

Condition Report 972390. An extent of condition re-review ofinput assumptions for controlroom doses then identified other conservative changes which were also factored into the BVPS Unit 2 controlroom dose calculation. This 10CFR50.59 evaluation only addresses the re evaluation condur'ed for a SmallBreak LOCA. The re-evaluation for the Large Break LOCA DBA and Locked Rotor DBA are addressedin separate 10CFR50.59's end this 10CFR50.59 draws no conclusioos regarding the two above listed DBAs.

The safety systems are the BVPS Unit 1 and Unit 2 controlroom emergency pressurization and filtration systems as describedin No. 3 in the Introduction Section.

1.A.2 Discuss the effects of the change and/or failure modes associated with the change on the probability of failure of the systems identified.

The proposed UFSAR changes originated from identifying a non conservative assumption used in the control room dose calculation for licensing both units regarding the possible controlroom pressurization fan flow rate. Previously all fan flow rater used in the calculati'ns were assumed to be 690 cfm even though plant technical specifications require as .' val tested flow rates between 800 and 1000 cim. This flow rate was used since mimmizing the fan flow was assurned to maximize the dose to personnelin the controlroom. This was describedin section 6.4.4.1 of the U2 UFSAR. Allcalculations which modeled the controlroom emergency pressurization fans were reevaluated using a maximized flow

• ate (rather than the currently used minimized flow rate) to determine if a more limiting (i.e., higher) control room dose would be calculated. Re-running control room dose calculations with a maximized controlroom emergency pressurization flow rate (i.e.,1030 cim) rather than the minimized flow rate (i.e., 690 cim) resulted in one BVPS Unit 2 dose result where the UFSAR listed values could have needed to be increased, to represent the most limiting bounding controlroom dose case for that DBA.

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10 CFR 50.59 EVALUATION WORKSHEET (:ontinued)

EM 115790 identified that pressurization tan flow rate may be lower than 690 cim, and provided a new icwer boundirig flow rate of 600 cfm for use in the radiological consequence analysis. Using this value resultedin BVPS Unit 2 SmallBreak LOCA voses presented in UFSAR Section 15.6.5.5 potentially changing.

(Note that the cont of room habitability system design assumedis the design described b the BVPS Unit 2 UFSAli for the Unit 2 ControlRoom Habitability Systems. BVPS Unit \

2 recently shutdown due to the inability to completely meet this design. This 50.59 evaluation assumes that the original design as described in the BVPS Unit 1 Technical Specification 3/4. 7. 7.1 l'ases can be retumed for the BVPS Unit 2 control room

\

emergency pressurizat!on system, via a separate 50.59 evaluation (or a different design and 50.S9 is provided recognizing the results of this 50.59). The issues involving single failure and the timing / tolerances of controlroom bottle dep.essurization and control room emergency pressurization fan initiation were resolved with DCP 2306 and DCP 2311.

This evaluation makes no new conclusions regarding the effectiveness of the current DVPS Unit 1 or Unit 2 controlroom habitability systems to conduct their design functions, only that the design can adequately function as currently described in the

\

respective UFSAR. such that the generic methodology issue can be addressed on its own merits. No changes are assumed for the response of the controlroom habitability .

systems nor any changes to current operator responses as described or intended by the current UFSAR descriptions. This 50.59 evaluation is addressing the extent of condition for the non conservatism of methodology assumptions only.)

The failure mode listed in No. 4 of the introduction Section addresses the ability of a methodology to be effective. No failure modes are llsted which involve the operation or existence of a system, structure or component.

The proposed UFSAR page snd table changes support the January 1998 dosu calculation revision.

The proposed UFSAR table changes do not increase the probab4y of occurrence of a malfunction of equipment important to safety because the changes do not involve equipment. The dose revisions more correctly show that the design for the current Unit 1 & 2 controlroom pressurization and filtration system as describedin their current UFSAR concinue to adequately operate to meet their design basis that the controlroom is provided adequate radiation protection.

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10 CFR 50.59 EVALUATION WORKSHEET (continued) 1.A.3 Discuss the effect of the change on the performance of the safety SYSTEMS.

The safety systems (described in No. 3 in the Introduction Section) involved with the control room dose methodology assumptions are unal.'ected by the proposed UFSAR table changes. This 10CFR50.59 evaluation addresses, a change in the assumptions y

needed as a result of an identifiedprevious non conservatism, and the methodology used for the analysis. l The proposed UFSAR changes do not directly affect the operation of these systems and do not have a physical constraint or impact on the' =ustems. Any modification done to enhance the current controlroom habitability syster , 'llbe addressed by a separate 10CFR50.59 evaluation.

The proposed UFSAR changes do not change the performance the of controlroom emergency pressurization and filtration systems as previously c'escribedin the current BVPS Unit 2 UFSAR siw e they do not involve equipment modificatinns.

l 1.A.4 Identify the design oasis accidents in the UPDATED FINAL SAFETY ANALYSIS REPORT I

(UFSAR) to be reviewed for potentialimpact by the change.

\

BVPS 2 'JFSAR Table 15.013 lists the applicable design basis accidents for which controlro0m doses are calculated.

Since BY'3 Unit I and Unit 2 share a combined controlroont, a design basis accident at

) either Unit will affect the common controlroom, and hence either Units' control room personnel. However, Unit I design basis accidents are not analyzedin this 10CFR50.59

(

evaluation becau:e 1) this evaluation only addresses the Unit 2 license and UFSAR, and

2) the controlroom radiological consequences of a Small Break LOCA were not analyzed under the Unit I license in the Unit 1 UFSAR. Unit I design basis accidents are not ac'dressedin the Unit 2 UFSAR.

1.A.5 Discuss how the parameters and SYSTEMS, affected by the change, affect the assumptions and radiological consequences of the accident (s) identified in Part 1, A.4.

As described in 1.A.2, a minimum controlroom emergency pressurization systern fan flow rate of 690 cfm was used in all BVPS calculations used to calculate control room doses for allDBAs. Recently it was identified that maximizing the actualpressurization system fan flow rate may increase the calculated controlroom dose for certain DBAs.

As a result, all Unit 2 events which credit isolation of the controlroom in their calculation of controlroom dose were re evaluated to determine whether maximizing or minimizing the actual expected control room emergency pressurization system fan flow rate after a DBA would increase the calculated controlroom doses.

A test was conductedinvolving the complete discharge of the BVPS CREBAPS bottled air. The test results suggested that the value used in the BVPS Unit 2 dose calculations for minimum possible CREBAPS air bottle discharge itcw rate, which was previously thought to have been chosen as a conservatively low value, may not have been chosen to be low enough to bound allpossible conditions.

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4 10 CFR 50.59 EVALUATION WORKSHEET (continued)

(The primary intent of the CRFBAPS air bottles is to mainta.'., an air flow discharge into the control room in order to m tintain the control room at a positive pressure relative to the outside environment (defined to be 1/8

• w.c.) such that intro : *>n of outside environment (assumed to be contaminated by the postulated DBA) is minimized. The Standard Review Plan requires that dose calculations assumer a constant 10 cfra l

unfiltered inleakage even when the control room is maintained at its design positive

\

pressure (to model some intrusion due to control room door openings). Thus the calculation always models some levelof radioactivity within the controlroom envelope.

An unintended but beneficial effect of CREBAPS air bottle discharge into the control room is the continuing dilution effect of the air mass within the controlroom. With the introduction of radioactive free airinto the controlroom, the air exiting the controlroom (through the small existing leak paths in the controlroom, e.g. under the botto n of doors) will carry out some (small) amount of radioactivity that resides within the control mom at that time. If actual CREBAPS bottle discharge flow rate is less than the minimum discharge flow rate modeledinto the dose calco'ation, the calculated dose rate willnot be maximized. Thus, the modeled minimum CREttAPS bottled air discharge flow should bound all actual credible minimum flow rates possible. (Note: Assuming a perfectly tight control room boundary envelope wehout any leak paths, the controlroom could be maintained at the designed positive pressure of 1/8" w.c. with an extremely.

Iow bottle discharge rate. The value assumed must make some assumption regarding the credible, reasonable tightness achievable for the controlroom envelope.}

NED evaluated the test data, in addition to the current tightness of the controlroom envelope. A new value of 600 cfm was provided by NED (EM 115766) as the minimum credible average value for CREBAPS air bottle discharge flow rate into the controlroom, considering the test data obtained on January 13,1998. (Note: the dose calculation l assumes a constant (i.e., average) value throughout the 60 minutes that the bottled airis modeled to be discharging.] The 600 cfm value represents a value which conservatively bounds a dose analysis assumption for addressing dilution effect and also a value which willalways provide the required 1/8" w.c. This evaluation confirms that use of a 600 cfm value for minimum air bottle flow rate willmaintain the assumption used in the BVPS controlroom dose calculations, which support the BVPS Unit 2 UFSAR dose values.

Recognizing the potential that the dose for the BVPS Unit 2 SmallBreak LOCA analysis could also show an increased control room don as a result of using a lower CRE8APS bottk flow rate or a higher fan flow rate, the BVPS Unit 2 SmallBreak LOCA analysis was re-analyzed. The re analysis used the new lower conservative value of 600 cfm for the minimum CREBAPS air bottle discharge flow rate las provided by EM 115766).

NED also reviewed the auto start and selection operation of the pressurization fan system. They concluded (EM 115790) that it was necessary, because of defined accuracy penalties in the pressure switches, to select a setting which correlates to 600 cim. This is done to ensure practicaland reliable system operation, and at the same time to ensure that the 1/8" w.c. pressure is not lost.

Thus, the Unit 2 Small Break LOCA was reanalyzed using a CREBAPS flow of 600 cfm and a single pressurization fan flow rate of 600 c!m.

NPDAP t 18 Rev 4 Form Rev; 07f2tM Page7 $U N

10 CFR 50.59 EVAL.UATION WORKSHEET (continued)

As part of the calculation for dose to the controlroom personnel, BVPS uses WO values for specific release points. The values of WO for the controlroom were duvelopedin )

1992. These WQ values are based on BVPS site meteorology data collected from 1986 thru 1990. The values were ct'culated using the methodology in NUREG 5055 which w;

• submitted to the NRC for review. The first TechnicalSpecification amendment was approved by the NRC ts BVPS Unit 2 Amendment No. 46. Following the submittal of additionalmeteorology data to the NRC, use of the Ramsdellnaethodology was approved for Unit 2 in Technical Specification Amendment No. 57. The contro,' room NO values developed for the RWST are:

0-8 hours 8.25E 5 sec/ cubic meter

~

8 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 4.07E 5 14 days 3.13E 5 "

4 30 days 1.89E-5 "

The newer values were usedin this calculation and willcontinue to be usedin future calculations of dose due to releases from the RWST.

Two errors were identifiedin the originalcalculation which were correctedin this analysis. An input value for assumed continuous unfiltered control room inleakage was found to be a factor of 10 higher than discussed in the cal;ulation, and a small gamma dose contribution fmm activity deposited on the filters was inadvertently neglected in the totalgamma dose.

With the flow changes for the CREBAPS air bottles andpressurization fans, the new WO values, and the calculation corrections, the doses to the controlroom personnelare l calculated to be:

1 NEW Calc.: Thyroid: 9.37 Rem Gamme: 3.2E-3 Rem Beta: 1.9E 2 Rem UFSAR: Thyroid: 11.0 Rem Gamma: 3.2E-3 Rem Beta: 3.2E-2 Rem The new dose results are less than or equal to those currently in the Unit 2 UFSAR.

1.A.6 Identify the der.ign basis accidents, if any, for which f ailure modes associated with the enange can be an initiating event.

BVPS 2 UFSAR Table 15.0-13 lists the applicable design basis accidents for which controlroom doses are calculated.

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10 CFR 50.59 EVALUATION WORKSHEET (continued) 1.A.7 Discuss the effect of the change on the probability of occurrence of the design basis accidents identified in Part 1, A.6.

The design bases accidents describedin 1.A.6 are unaffected by the proposed UFSAR text and table changes. This 10CFRSO.59 evaluation only addresses a change in the methodology assumptions needed as a result of an identified pic':fous non conservatism.

The proposed UFSAR changes do not directly affect the probability of an accident since they are an administrative coricopt and do not have a physical constraint or impact on plant systems. '

Thus the proposed UFSAR changes do not increase the probability of occurrence of an accident as previously described in the current BVPS Unit 2 UFSAR.

1.B. Evaluation Questions 1.B.1 Based on Part 1, A.2, MAY the proposed change increase the probability of occurrance of a malfunction of equipment important to safety evaluated previously in the UFSAR?

O YES E NO l

1.B.2 Based on Part 1, A.3, MAY the proposed change increase the consequences of a malfunction of equipment important to safety evaluated previously in the UFSAR?

O YES E NO 1.B.3 Based on Part 1, A.5, MAY the proposed change increase the consequences of an accident evaluated previously in the UFSAR?

O YES E NO 1.B.4 Based ori Part 1, A.6 and A.7, MAY the proposed change increase the probability of occurrence of an accident evaluated previously in the UFSAR?

O YES 0 NO IF THE ANSWER TO ANY OF THE ABOVE QUESTIONS IS YES, THE CHANGE REPRESENTS AN UNREVIEWED SAFETY QUESTION PART 2: POTENTIAL FOR CREATION OF NEW TYPE OF UNANALYZED EVENT 2.A Supporting Information 2.A.1 Based on Part 1, assess the impact of the change and/or f ailure modes associated with the change, to determine if the impact has modified the plant response to the point where it can be considered a new type of accident. Discuss the basis for this determination.

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, .' 10 CFR 50.59 EVALUATl3N WORKSHEET (continued)

The ptsposed UFSAR table changes alone do not impact the plant response since they are based on a methodology assumption concern and do not have a physical constraint or impact on plant systems.

• This 10CFR50.59 evaluation addresses a new methodology and changes in the assumptions needed as a result of an identifiedprevious analyses input non conservatism.

The methodology used to calculate radiological consequences from a non-limiting Small Break LOCA is not prescribed by Standard Review Plan or Reg. Guide 1.70. The analysis that was incorporatedinto the Unit 2 UFSAR revision 7 (April 1995) used methodology and assumptions that had not been previously reviewed and approved by the NRC and therefore may be considered a new methodology different than previously described in the BVPS Unit 2 UFSAR as defined by recent NRC guidance.

2.A.2 Determine if the failure modes of equipment important to safety associated with the change represent a new unanalyzed type of malfunction. Discuss the basis for this determination.

The proposed UFSAR table changes alone do not have a physical constraint or impact on plant systems. This 10CFR50.59 evaluation only addrasses a new methodology and changes in the assumptions needed as a result of an identifiedprevious analyses input non conservatism.

The proposed UFSAR changes do not cause an impact to the plant responsa to the point where it can create the possibility of a malfunction of equipment important to safety of a different type than previously descdbedin the BVPS Unit 2 UFSAR.

2.Il Eva!uatior: Questions 2 .11. 1 Based on Part 2, A.*,, MAY the proposed activity create the possibility of an accident of a different type than any evaluated previously in the UFSAR?

Although no new types of accidents are created, the analysis B YES O NO represents a new methodology different than any evaluated previously in the UFSAR.

2.B.2 Based on Part 2, A.2, MAY the proposed activity create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the UFSAR?

O YES E NO IF THE ANSWER TO ANY OF THE ABOVE QUESTIONS IS YES, THE CHANGE REPRESENTS AN UNREVIEWED SAFETY QUESTION NPDAP t.18 Tev 4. Form Rev.: 07/21/97 Page10 som Ks

10 CFR 50.59 EVALUATION WORKSHEET (continued)  :

i PART 3: IMPACT ON THE MARGIN OF SAFETY

> 3.A Supporting Information '

3.A.1' identify the acceptance limits which form the licensing basis for the TECHNICAL SPECIFICATIONS (i.e., the accident analysis and other design basis) that could be affected by the change.

• ' Unit 2 Ta:l'nical Specification 3/4.7.7, "ControlRoom Ememency Alt Cleanup and Pressurization System. "

3.A.2 Discuss the impact of the change on the acceptance limits which form the basis for the TECHNICAL SPECIFICATIONS.

The calculated control room doses for the Small Break LOCA have been determined to i be less than or squal to those currently listed In the UFSAR. This decrease in

( consequences Is not a decrease in the mamin of safety.

3.B. Evaluation Question 3.B.1- Based on Part 3, A.1 and A.2, does the proposed activity reduce the margin of safety as defined in the basis for any TECHNICAL SPECIFICATION?

O YES E NO e IF THE ANSWER TO ANY OF THE ABOVE QUESTIONS IS YES l

, THE CHANGE REPRESENTS AN UNREVIEWED SAFETY QUESTION PART 4: 10 CFR 50.59 EVALUATION CONCLUSION

. Bated on the evaluation in Parts 1,2 and 3 the change:

O Does not involveh UNREVIEWED SAFETY QUESTION. -

B . involves an UNREVIEWED SAFETY QUESTION. Contact Safety and Licensing Department j ' before presenting to the Onsite Safety Committee.

t..

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10 CFR 50.59 EVALUATION WORKSH3T (continued)

PART5: ENVIRONMENTAL EVALUATION (Appilcable to BVPS Unit 2) 5.A Supporting Information 5.A.1 Identify any significant increase in any adverse environmen'.al impact previously evaluated in the Final Environmental Statement Operating License Stege, environmentalimpact appraisals, or in any decisions of Atomic Safety and Licent.ing Board.

No adverse environmental (non radiological) impact has been identified. The Unit 2 Environmental Protection Plan addresses the protection of non radiological environmental values. No aquatic / water quality, terrestrial, or noise issues are affected by this change in radiologicaldose calculations. A Loss of Coolant Accident was analyzedin the EnvironmentalReport. Re-analysis of a SmallBreaic LOCA has not resultedin an increase in the release of radionuclides since the changes made related only to tho dosss to the i operators in the controlroom 5.A.2 Identify any significant change in effluents or power level.

The change in radiological dose calculations does not affect effluents or power level.

5.A.3 Identify any matters not previously reviewed and evaluated in the Environmental Protection l Plan, Final Environmental Statement Operating License Stage, or NPDES permit, which MAY have a significant adverse environmentL1 impact.

There are no environmentalprotection issues that wouldinvolve an unreviewed environmental question or that involve a change to the Environmental Protection Plan.

5.B. Evaluation Question Based upon Part 5, A.1, A.2 and A.3, the change:

E3 Does not involve an UNREVIEWED ENVIRONMENTAL QUESTION.

O Involves an UNREVIEWED ENVIRONMENTAL QUESTION. Contact Safety and Ucensing Department before presenting to the Onsite Safety Committee.

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