ML20154K939

Responds to NRC Re Violations Noted in Insp Rept 50-395/88-13.Corrective Actions:Format for Displaying Data for post-trip Review Upgraded & Chemical Treament/Flushing of Reactor Bldg Cooling Unit Performed

ML20154K939
Person / Time
Site:	Summer
Issue date:	09/16/1988
From:	Bradham O SOUTH CAROLINA ELECTRIC & GAS CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
NUDOCS 8809260052
Download: ML20154K939 (24)

Text

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10CFR2.201

, th Caro Ina Ele:tric & Gas Company olne . adham JenOnsMe SC 29065 Nuclear OperaSc ns WO3) 345-4040

__ September 16, 1988 Document Control Oesk U. S. Nuclear Regulatory Commission Washington, OC 20555

Subject:

Virgil C. Summer Nuclear Station Docket No. 50/3?"

Operating License No. NPF-12 R Response to Notice of Violation NRC Inspection Report 88-13 Gentlemen:

Enclosed is the South Carolina Electric & Gas Company (SCE&G) response to the Notice Of Violation (NOV) dated August 17, 1988 (EA-88-151). Attachments 1 '

and 2 to this letter are the SCE&G "Response to the Notice of Violation" (see 10 CFR 2.201) and "Answer to the Notice of Violation" (see 10 CFR { 2.205),

respectively.

As indicated during the Enforcen,ent Conference, SCE&G denies t1at a violation of plant technical specifications occurred. In sum, SCE&G bel: eves that (1) post-trip recovery reviews were adequate, (2) degraded Service Water System flow was discovered by the Licensee within a reasonable time period, (3)

SCE&G was conservative in declaring both Reactor Building Cooling Units inoperable after degraded flow conditions were discovered (4) Reactor Building Cooling Units were at all times capable of performing their intended safety function, and (5) the post trip review program was effective in that it was via this process that the Licensee identified the degraded flow condition. If the Staff ultimately determines that a violation did occur, i the SCE&G position is th'.t the event that gave rise to the NOV has been '

incorrectly categorized by the NFC as a "cause for significant concern" and I should not be categorized greater than a Severity Level IV enforc eent action.

SCE&G would like to emphasize that, even though it believes nc Tk:hnical Specification Limiting Conditions for Operations were violated, and there was no significant impact on the ability to protect the health and sefety of the 1 public as a result of this incidant, it recognizes and appreciates tha

• potential seriousness of incidents of this nature. SCE&G believes that its record in identifying and correcting potential problem areas in its operations is an excellent one, and will make every effort to assure that it continues in that manner. SCE&G believes that the corrective actions taken in regard to this incident demonstrate the strength of its concern with safe operation cf the Virgil C. Summer Nuclear Station.

B'309260052 800916 PD3 0 Al>0CK 05000395 N PNV p j

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Cocument Control Desk 1 Page 2 September 16, 1988 r

I If you should have any questions, please advise. l Very truly yours, l 1

hlho

0. S. Bradham Y

~; HID/OSB: led  ;

Attachments c: D. A. Nauman/J. G. Connelly, Jr./0. W. Dixon, Jr./T. C. Nichols, Jr. l E. C. Roberts  :

, W. A. Williams, Jr.  !

) General Managers

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i J. Nelson Grace C. A. Price /R. M. Campbell, Jr. ,

j R. B. Clary ,

J. R. Proper ,

K. E. Nodland  !

J. C. Snelson  !

, G. O. Percival f R. L. Prevatte t

. J. B. Knotts, Jr.  !

NSRC l l

i RTS (IE 880013) t NPCF i File (815.01)  !

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• Attachment 1 to Docuacnt Control Desk Letter Septembe.r 16, 1988 Page 1 of 3 ATTACHMENT 1 RESPONSE TO NOTICE OF VIOLATIOM VIOLATION NUMBER 50-395/89-13-01

1. INTRODUCTION As discussed further below in the 10 CFR {f 2.201 and 2.205 i responses, SCE&G believes that it should not have necessarily known  !

and was not required by procedure to verify that flow to at least one train of the Reactor Building Cooling Units (RBCU's) had degraded to a point that Action Statement 3.6.2.3 was invoked until the completion of its post-trip recovery analysis. In addition, post- ,

trip review procedures allow plant startup pending an in-depth review l of system and component trends following the trip.

In the alternative. SCE&G maintains that Service Water Booster Pumps (SWBP's) degraded flow to the RBCU's was not sufficiently safety significant to warrant escalated enforcement action.

l1. DISCUSSION The Notice of Violation states:

l "During the NRC inspection conducted on May 1 - 31, 1988, a violation of HRC requirements was identified. In accordance with the ' General Statement of Policy and Procedure for NRC Enforcement Actions,' 10 l CFR Part 2, Appendix C (1988), the violetion is listed below:

"Technical Specification (TS) 3.6.2.3 requires two ici:,,endent j groups of RBCU's be operable in Modes 1, 2, 3, ar.d 4. ACTION Statement 'b' of TS 3.6.2.3 requires that with both trains of RBCU's inoperable and both trains of reactor building spray system operable, to restore at least one train within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least hot standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

"TS 3.0.4 specifies entry into an operational mode shall not be made unless the conditions of the Limiting Condition for Operation are met without reliance on provisions contained in the ACTION requirements.

'TS 6.8.1 requires procedures to be established and implemented covering the activities referenced in Appendix 'A' of Regulatory Guide 1.33, Revision 2. February 1978. Apoendix 'A' of Regulatory 1.33. Revision 2, specifies that administrative l

t procedures be established and implemented. Station Administrative Procedure 132 requh'es that the shift engineer i review tne computer post trip review printout (that includes I service water flow) prior to plant restart. It specifically states to ascertain the cause of each alarm and determine that any recuired automatic action functioned properly.

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' Attachment 1 to Document Control Dask Letter Septemb:r 16, 1988 Page 2 of 3 "Contrary to the above, entry into Mode 2 was made at 8:24 p.m.

on May 12, 1988, and subsequently into Mode 1 at 12:21 a.m. on May 13, 1988, with both trains of RBCU's inoperable. The post trip review failed to detect that RBCU's were inoperable due to low service water flow, and therefore, the plant was in TS 3.6.2.3 Action Statement 'b.'

aThis is a Severity Level III violation (Supplement 1)."

ADMISSION OR DENIAL OF THE ALLEGE 0 VIOLATION SCELG does not agree that it should have necessarily known prior to entering Mode 2 on May 12, 1988, and subsequently into Mode 1 on May 13, 1988, that the flow to the RBCU's had been degraded to a level i that required the implementation of Action Statement 3.6.2.3. In l addition, after determining that a degraded flow condition existed. l SCE&G conservatively declared both trains inoperable when, in fact,  !

it was likely that only one of two trains actually had degraded flow below technical specification requirements. ,

i SCE&G also believes that post-trip reviews were adequate and were performed according to procedure. In fact, post-trip reviews resulted in the discovery of the service water low-flow condition.

REASON FOR THE VIOLATION 1

Not Applicable l

l CORRECTIVE STEPS TAKEN AND RESULTS ACHIEVED Notwithstanding the SCE&G denial of the above violations, certain actions have been taken to improve the ability of plant personrel to quickly perform comprehensive post-trip analyses:

l The format for displaying data for the post-trip review has been upgraded, making potential system problems more readily apparent (graphical comparison of specific plant parameters plotted against their expected or alarm values).

Chemical treatment / flushing of the RBCU's, and intpecting/ cleaning of clams out of the intake structure has resulted in restoration of flow (greater than the minimally required flow rate) from the SWBP's to the RBCU's.

Modifications to the RBCU's to allow more frequent on-line flushing /backflushing capabilities are currently being considered.

' Attidhment 1 to Document Control Desk Letter September 16. 1988 ,

Page 3 of 3 A modification that reduces minimum flow requirements by providing flow isolation of two of the four RBCU's is currently in progress. Only one RBCU is required to meet design basis cooling for the system.

Additional emphasis will be placed on operator response to annunciators and the importance of following all applicabic procedures.

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. Attachment 2 to 01cument Control Desk Letter l

September 16, 1930 l Page 1 of 3 ATTACHMENT 2 ANSWER TO NOTICE OF VIOLATION  !

ENFORCEMENT ACTION 88-161 l A.

SUMMARY

OF POSITION l

l As noted in Attachment 1 above, SCE&G denies the subject violation. I In the alternative, should the Staff maintain its position regarding the occurrence of a violation, as-found conditions were not sufficiently safety significant to warrant escalated enforcement action.

As discussed further below, the degraded flow conditions did not

! result in the RBCU's being unable to perform their safety function.

B. DISCUSSION l

1. Post-Accident Recovery Process As previously discussed at the June 24, 1988 enforcement l conference, the post-trip recovery process is a two-tier system.

l The first tier involves analysis of the following issues: (1) l

' the cause of the trip; (2) whether the cause of the trip still exists; (3) whether the event requires the implementation nf the Emergency Plan; (4) whether any Limiting Safety System Setting  :

has been exceeded; and (5) whether any Safety Limit has been l l exceeded.

J As correctly referenced in the N0'!, the Shift Engineer shall i also review the Plant Process Computer Sequence of Events i Printout ;o (1) ascertain the cause of each item on the l printout. (2) verify that Reactor Trip Breakers opened as  ;

required,(3)verifythataManualReactorTripwasinitiated, I (4) verify a Turbine Trip as required, (5) verify Main Steam  !

l Line Isolation as required, (7) verify other safety equipment '

l start as required, and (8) verify Emergency Feedwater start as required. Additional actions by the Shift Supervisor, Control Room Supervisor and Reactor Operator are required prior to returning the plant to power. (See Station Administrative l Procedure 132 5 6.4.3.) Startup of the Service Water Booster  :

Pumps (SWBP) and verification of initial flow through the RBCU's  !

(4000 gpm) was verified by the Shift Engineer. The Shift 1

Engineer should not have necessarily known and was not required I by procedure to verify whether the RBCU flow had decreased  !

subsequent to initial SWBP's startup and establishment of l

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minimum flow. In addition, the Shift Engineer could have l accepted the downward flow trend as a result of Operator action  !

being taken to secure the SWBP's. Flow rates to the RBCU's could remain at =2000 gpm due to the operating Service Water Pumps.

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Attachment 2 ts Document Control Desk Letter 4

September 16, 1988 Page 2 of 3 Subsequent to a restart determination, the Reactor Trip Package (RTP) was forwarded to the Independent Safety Engineering Group (ISEG) for the second tier review /analy;is. It was this review that discovered the degraded flow condition.

SCE&G personnel satisfactorily completed all steps of the procedure and properly authorized the restart of the unit. As discussed during the Enforcement Conference, the reason the initial post-trip review did not discover that a SwBP low-flow alarm had actuated was due to the fact that limited computer output capacity preclud(J its inclusion in the RTP.

While SCE&G acknowledges a deficiency, the circumstances did not result in a violation of existing procedures or Technical Specifications.

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2. Discovery of the SWBP Low Flow Condition 10 CFR Part 2, Appendix C, Section V.A states that, "Licensees are not ordinarily cited for violations resulting from matters not within their control, such as equipment failures that were not avoidable by l reasonable licensee quality assurance measures or aanagement l controls." As previously stated, SCE&G believes that the cited '

degraded flow condition was not necessarily apparent and was not required by procedure to be verified during the initial post-trip review. The initial reviewer reviewed the plant process alarm printout which did not indicate a low flow alarm. Subseauent component / system performance was appropriately analyzed during the i "TP review by ISEG after plant startup had been authorized.

l 3. Safety Significance _of As-found Conditions Analysis by SCE&G concluded that flow in RBCU Train A had reduced to approximately 3900 qpm and Train B had reduced to approximately 2000 gpm some time after the SWBP's had initially started after the trip.

Conservatively SCE&G declared both trains inoperatale and entved the appropriate Technical Specification Action Statement 3.6.2.3. The Train A indicated flow was likely at or above the required flow when considering instrument error (1 400 gpm).

Even if the lowest flow rate is assumed, analysis concludes that flow was sufficient to meet all design basis conditions. Therefore, there is no safety significance regarding the degraded flow condition (analysis previously provided; copy attached for reference).

s Attachment 2 to Document Ctntrol Desk Letter September 16, 1988 Page 3 of 3

4. Severity level of the Violation Should the Staff ultimately determine that a violation occurred as stated, it should be categorized at no greater than a Severity Level IV enforcement action. 10 CFR Part 2, Appendix C, Supplement 1, %

0.1 states that a Severity level IV violation involves "A less significant violation of a Technical Specification Limiting Condition for Operation where the appropriate Action Statement was not satisfied within the time allotted by the Action Statement."

SCELG believes that the facts discussed above clectly indicate that t the degraded flow condition did not result in a safety significant issue in that all affected systems could have performed their design basis functions. Therefore this should not be categorized as higher than a Severity Level IV enforcement action.

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Attachment 3 to Docume: Control Letter September 16, 1988 SOUTH CAROLINA ELECTRIC & gas COMPANY w nea.wmme EMGINEERING SERVICE { CGSS: 21641 (Office) File: 4.534 14.3700 SUBJECT V. C. Summer Nuclear Station Date July 19, 1988 Reactor Building Cooling Units Consequence Analysis Reference CGGS-37485 dated July 19, 1988 To A. R. Koon Attention of A "Consequence Analysis

• to evaluate the impact on plant performance of reduced R8CU heat removal capability due to the combination of reduced RSCU fan flow of 54,200 ACFM and reduced service water fic w of 2,200 gpm for the 'B train' RBCU's has been performed.

The "Consequence Analysis" evaluations performed have d7monstrated that the degraded RBCU performance combined with postulated accidents does not result in exceeding any regulatory guidelines or loss of any equipment required for safe shutdown.

If additional discussion is necessary, please contact me at extension 4703.

/s.T. Estes, Jr Senior Mechanic Engineer Design assis Engineering C0 R. B. Clary, Manager Design Engineering cca 8. T. Estes G. V. Meyer WPCF Filt/R. B. Clary

[ Gilbert / Commonwealth engineers andconsultants G<LBERT{0VVONWf ALTH, AC .

• O Ser 1498, Reaci ng PA 19603 / Te; 115 7754600 s Cable Gdaso< t re en 836-411 July 19,1938

.\tr. ft. II. Clary, .\1anager CGGS-37485 Design Engineering South Carolina Electric & Gas Company P.O. Ilos 83 Re: V. C. Summer Nuclear Station Jenkinsville,SC 29065 G/C W O. 04 5650 500 iteact, iluilding Cooling Units.

Attention: .\f r. B. T. Estes Consequence Analysis file Code: 1.1.6 500/4.14 Response Code NRR

Dear .\f r. Cla ry:

Per your request, a "Consequence Anal 3 sis" to evaluate the impact on plant performance of reduced RBCU heat removal capability due to the combination of reduced RBCU fan now of 54,200 ACF.\1 and reduced service water Cow of 2,200 gpm for the'B train' RBCC's has been performed. The taskr identified for evaluation and their Gnal status is as follows:

TASK FIN Al. STATUS

1. Establish degraded itBC C performance New values provided by American Air Filter

2. Evaluate SW system pressure and the VeriDed calculation yielding minimum SW impact on RBCU performance pressure at IlflCU's

3. .\tSt.B Pressurefremperature analysis Verified calculation yielding pressure / temperature inside Reactor !!uilding above licensing values but within regulatory guidelines

4. l.OC A Pressure, Temperature analysis Verifled calculation yielding pressure temperature above licensing values but within regulatory guidelines 5.

Equipment QualiGeation iEQ) Evaluation Verined calculation, all equipment qualiGed for Tasks 3 and 4 pressure temperature conditions 6 Offsite, Control Room Doses No change from previously evaluated do. es

7. Instrument loop Accuracies Verined calculation, all Ril I E instrument loop accuracies analysed for pressure. tem perat ure conditions equal or greater than tha>e evaluated in tasks 3 and 4 NP la i 44 4 **. % ll

' (Mlbert/ Commonwealth eaanneers anda W *;

G4 SERT 40MMONWEALTH tNC.

Mr. R. B. Clary, Manager July 19,1948 CGG M 7488 Page 2 1

Each of these tasks is addressed in more detail within Attachment 1 of this letter.

Very truly yours, ,

R. E. Anderson '

Applied Engineering Analy is Task Engineer

&K.E.N x and ['

Engineering Project Manager l

l REA/ KEN:tln Attachment cc: NPCF w/att B. T. Estes w/att l

D. A. lavigne w/att s K. E. Nodland (2) w/2 att S. R. Hunt w/att G. Meyer L W. Kunkel w/o att w/att P.L. Bunker w/att J. L Skolds w/stt D. H. Stevens A. R. Koon, Jr.

w/att w/att E.J. Anselmi w/att D.J.Lengel w/att C.M.Hees w/att R. E. Anderson w/att S. M. Cisek w/att

CGGS-37485 Attachment 1 Page 1 of 8 V.C. Summer Nuclear Station RDCU Consequence Analysis

COGS .17485 -

Attachment 1 PagG 8 of 8 i TABLE OF CONTENTS Page

. COVER SHEET 1  !

TABLE OF CONTENTS 2 t

INTRODUCTION 3 E I

1. Degraded RBCU Capability 3 I

2. SW System Pressure / impact on  ! '

RBCU Capability i i

3. MSLB Pressure / Temperature inside 4 '

Reactor Building f

! 4. LOCA Pressure / Temperature 5  :

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i 5. Equipment Qualification Evaluation 6

6. Offsite/ Control Room Doses 7  !

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7. Instrument Loop Accuracles 7 i l L s '

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SUMMARY

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CGGS-37485 Attachment 1 Page 3 of 8 INTRODUCTION A "Consequence Analysis evaluating the impact of degraded RBCU performance on the post accident response of the V.C. Summer Nuclear Power Plant has been completed.

This evaluation identifies and quantifies the post accident impact on safety concerns due to degraded RBCU performance resulting from the comb!nt'lon of reduced RBCU fan flow (fan capacity at lower Tech. Spec. limit) and reduced r9rvice water flow to the

'B' train RBCU's of 2,200 gpm, which corresponds to the minimam documented RBCU flow rate which would have been present during plant restart.

The major assumptions used to perforni this evaluation are as follown

1. Loss of Offsite Power

2. 'A' Train Diesel Generator fa!!ure to start I

3. One 'B' Train RBCU fan is in service

5. 'B in Ser ice ater fl w of 2 2 0 m,1,100 gpm to each RBCU

6. Service Water temperature of 66.70F (maximum SW temperature during degraded flow conditions)

7. No Reactor Building Cooling Unit identified leakage.

Section 1: Establish Degraded RBCU Performance American Air Filter (AAF), the RBCU vendor, was asked te evaluate the heat removal capabihtles of the V.C. Summer RBCU's for the conditions given in Table 1. Table 2 provides the calculated heat removal rates developed by AAF using the methodology described in their approved Topical Report No. TR7101 A.

Section 2: Service Water System Pressure Analysis The reduced SW flow through the Reactor Building Cooling Units will n'fect the pressure in the SW System. A new RBCU outlet SW pressure was calculated using the reduced 2,200 gpm flow (1,100 gpm/RBCU) and a normallow pond levei of 420.5 ft. The pressure at the RBCU was calculated working back from the discharge and assuming that the increased system pressure drop contributing to the flow reduction is entirely

CGGS-37485

' ' Atttchment 1 Page 4 of 8 upstream of the RBCU outlet. This assumption will result in a bounding worst case minimum SW pressure value being calculated at the RBCU outlet. The calculated pressure is 5.95 psia at the RBCU coil outlet.

The 5.95 psia SW pressure at the RBCU outlet results in a saturation temperature of 169oF. When the temperature of the SW in the RBCU coil reaches 1690F, some steam formation will occur due to heat transfer and will result in decreasing the heat transfer coefficient for the coil downstream of the point where the saturation temperature is recched. Performance dsta for the RBCU coils at 1,100 gpm indicates that a 200oF Reactor BL.tdtng post ate! dent temperature will result in a SW outlet temperature from the RBCU of 1630F. Since this is less than the 169oF saturation temperature, no steam formation in the RBCU's will occur with Reactor Building temperatures of 2000F or less.

I Section 3: MSLB Pressure / Temperature Analysis inside Reactor Building The Licensing Basis Matri Steam Line Breaks (MSLB) were reanalyzed using the RBCU j i

heat removal capacity identifled in Table 3. Table 3 provides a comparison of the degraded RBCU performance versus the RBCU performance used for the Licensing Basis Accidents. The degraded RBCU performance is based on the data provided by AAF for conditions 3 and 4 (see Tables 1 and 2). Energy removal by the RBCU was cotservatively set to zero when the Reactor Dullding (RB) temperature exceeds 2000F.

This approach very conservatively bounds RBCU performance when Service Water flashes within the RBCU due low Service Water pressure as discussed in Section 2.

A comparison of Licensing Basis and degraded RBCU cot 4 TEMPT LT-26 pressure / temperature results for the peak pressure MSLB (1.4 ft2 DER at 102% power) are provided in Table 4. The maximum calculated degraded RBCU analysis pressure, including an initial RB pressure of 1.5 psia to account for maximum allowable normal operation Technical Specification, is 51.23 psig. This peak calcalated pressure of 51.23 psig is well under the Reactor Building design pressure of 57.0 psig which Standard Review Plan (SRP) 6.2.1.1.A. Section ll.a speelfles for licensing of operating plants, and I i is also less than the peak calculated value of 51.8 psig which corresponds to the 10%

safety margin speelfled in SRP 6.2.1.1.A, Section ll.a for plants in the Construction Permit stage.

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CGGS-37485 l .

, Attachment 1 d

Page 5 of 8 A reanalysis of MSLB peak temperature (0.645 Split Rupture at 102% power) was not performed since this calculated temperature is due to superheating of the RB post

accident atmosphere which is quenched by RB sprays prior to RBCU Initiation.

, Conservative assumptions incorporated into this analysis includes i

! 1. Multip?e failures (Diesel Generntor Failure, Male Steam isolation Valve Fallure.

Emergency Feedwater Control Valve Failure) used for determining L ass / energy j release (Westinghouse BIT Removal Analysts Assumptions).

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2. Use of maximum Technical Specification allowable normal operation pressure of l 1.5 psig.

3. No heat removal by RBCU at RB temperatures above 20noF.

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l j Section 4: LOCA Pressure / Temperature Analysis j The degraded RBCU performance was also analyzed for the LOCA event (the long term 1 governing pressure / temperature event). The Double Ended Pump Suction LOCA

{ Contempt LT-22 model was updated for use of Contempt LT-26. Model changes

incorporated include

Licensing RBCU Consequence j *. malysis Analysis

1. Environmental conditions Temperature 900F 95oF conservatively changed tot llumidity 50% 70 %

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! 2. liest transfer coeffielents used with passive heat sink .nodels are the Tagamt heat s

transfer coefficient used through blowdown (approximately 17.2 seconds post j

accident) and the Uchida heat transfer coefficient thereafter.

J 3. Spray initiation and service water flow to RBCU timing is conservatively set to

! the same times as for the MSLB.

,I j 4. Convection and radiation heat transfer is allowed to the environment from the j outer face of the RB concrete shell and dome.

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CGGS-37485

. Attachment 1 Pago 6 of 8

5. The RBCU performance given in Table 3 was used.

The resultant RB pressure / temperature profile is tabulated in Table #5. These results show calculated peak pressure / temperatures slightly higher (60.0 psia vs. 59.36 psia, 267.90F vs. 268.70F) than tha Licensing Basis LOCA values. Calculated RB LOCA pressure remains more than 10% below design pressure. Calculated RB LOCA temperature remains below design of 233oF. l Conservative assumptions incorporated into the analysts include:

1. Use of maximum Technical Speelfication allowable normal operation RB pressure of 1.5 pstg.

2. RB sprays are assumed to automatically go into the recirculation mode for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

3. No heat removal by RBCU at RB temperatures above 2000F (the RBCU's begin to remove energy from the RB atmospt'ere at approximately 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> post accident).

Additionally, a LOCA evaluation assuming RB Spray operation for only 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per FSAR Section 6.2.2.2.1.2 was performed. Recults of this analysis are also tabulated in Table 5.

A comparison of these two cases shows that sprays running for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> yields the highest Reactor Building pressures and vapor temperatures; whereas, the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> spray case yle!ds the highest Reactor Building sump temperatures.

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Section 5

Equipment Qualification Evaluation An evaluation was performed to determine the effects of increased LOCA and MSLB temperatures and pressures resulting from RBCU reduced flow conditions on Class 1E equipmert which would have been exposed to the postulated accident t nvironment.

The evaluation compared the newly calculated LOCA and MSLB temperature and pressure versus time profiles with the LOCA test profiles to which the Class 1E equipment was subjected during environmental qualification testing. Acceptance

CGGS-37485 Attachment 1 Page 7 of 8 criteria was based on whether the tested profile enveloped the postulated accident profiles.

Any potential deviations or postulated temperature excursions which exceeded those of the test profile were documented and evaluated. It was determined that the affected Class IE equipment would have been quallfled if they had been exposed to the postulated LOCA/5tSLB environment occurring during RBCU low flow conditions.

Section 6: Offsite/ Control Room Doses The reduced air flow (54,200 ACF51) through the RBCU coils combined with reduced l

service water flow (1,100 gpm C 66.70F) to the RBCU coils does not Impact the offsite or control room doses (dose assessment based on 54,200 ACF51 RBCU fan flow previously reported in letters CGGS-37423 and CGGS-37450, dated June 24,1988 and June 30,1988 respectively). These calculated doses are conservatively based on design

containment pressure for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> post accident and 1/2 that value thereafter.

Thus, the relatively small changes in RB pressure / temperature response resulting from reduced RBCU performance will not result in calculated offsite and control room doses above the current Licensing Basis.

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l Section 7: Instrument Loop Accuracles F i

An analysis was made of the impact on IE instrument loop accuracles from increased StSLB and LOCA pressure / temperature resulting from the RBCU Consequence Analysis. I i

Calculations for these loops included insulation resistance (IR) degradation effects of  :

cabling from accident conditions as well as component errors.

1 There are no degraded protective function actuations as a result of the new RBCU i

temperature / pressure profile since all protective actuations occur within the first 5

) minutes of the initiating event. During this period, the accident profiles are essentially j identical to those previously evaluated.  !

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' ' CGGS-37485 Attachment 1 Pago 8 of 8 Display accuracles are generally unaffected by the new temperature / pressure profile since the existing calculations utilized higher bounding temperatures than results from the new RBCU analysis. Reactor Building Level and Steam Generator wide range level indications are the only Post Accident Display channels which were not completely bounded, however, no significant affects (less than 0.1%) were created and no margins were reduced beyond allownble values.

Summary:

l The "Consequence Analysis" evaluations performed have demonstrated that the degraded RDCU performance combined with postulated accidents does not result in exceeding any regulatory guidelines or loss of any equipment required for safe shutdown.

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TABLE 1 l RBCU Performance Capability - Degraded Conditions '

1 EA N R AL BUILDING TEMPERATURE l g,y) PRESSURE j CONDITION f (PSIA) l la  !

283 59.4

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lb 241 59.4 Ic 200 59.4 Id 160 59.4 2 241 44.

3 200 30.

4 160 22.

All evaluations based ont 54,200 ACFM, Fan Flow at inlet 1,100 gpm, Service Water flow to 'B' train in-service RBCU 66.7oF, Service water temperature 100%, Reactor Building Humidity 0.0005, Cooling Coll Fouling Facta NOTE: Reactor Building design pressure is 57.0 psig, 71.7 psia

TABLE 2 RBCU Performance Capability

1. ACFM at fan inlet 54203 54201 54200 54202 54200 54200 54200

3. lient Removal Capacity 97.79 71.75 48.29 28.40 74.21 51.69 30.40 in BTU /hr (x106)

3. Cooling Water GPM 1100 1100 *100 1100 1100 1100 1100 Entering Temp 'F 66.7 66.7 66.7 66.7 66.7 66.7 66.7 Leaving Temp 'F 254.6 201.8 156.2 118.9 206.7 162.8 122.6

4. Co!! Entering Air ACFM 67412 63455 60557 58370 66820 66456 64028 1

'F 283 241 200 160 241 200 160

5. Coll Leaving Air ACFM 54203 54201 54200 54202 54200 54200 54200

'F 280.6 228.8 180.1 137.3 231.5 182.5 133.9 l

l Density .1483 .2031 .2390 .2643 ,1405 .1136 .0059 {

6. Motsture Condensation 104,985 70,829 42,306 20,424 75,119 49,435 26,670 rate in Ib. water /hr I.

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TABLE 3 RBCU Performance Capability - Degraded Conditions Heat Rimoval Capability REACTOR RBCU BUILDING LICENSING BASIS LICENSING BASIS CONSEQUENCE -

TEMPERATURE MSLB (BTU /liR) LOCA (BTU /IIR) ANALYSIS (OF) (BTU /IIR) 283 125 x 106 100 x 106 0, 241 90 x 106 75.7 x 106 0.

200 57 x 106 51.8 x 106 51.7 x 106 160 29 x 106 28.5 x 106 30.4 x 106 1

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TABLE 4 Case Comparison - MSLB PRES 8URE' TEMPERATURE TIME (PSIA) (oF)

(SEC.)

Licensing RBCU" Licensing RHCU" 1 18.84 18.84 186.0 186.0 2 22.13 22.13 222.8 222.8 5 28.00 28.00 231.3 231.3 10 35.31 35.31 224.1 224.1 15 40.65 40.65 236.0 I

236.9 20 41.87 41.87 239.7 239.7 40 44.91 44.91 246.4 246.4 60 46.55 46.55 249.5 249.5 100 49.30 49.45 254.5 254.8 140 $ 2.24 52.60 259.4 260.0 200 55.32 57.56 263.9 267.7 280 59.16 60.26 269.4 271.0 400 58.09 59.84 267.8 270.3 500 57.86 60,11 267.4 270.7 600 57.70 60.55 267.1 271.2 700 57.80 61.11 267.3 271.9 800 57.96 61.74 267.5 272.4 900 58.18 62.40 267.8 273.7 1

1000 58.43 63.10 268.2 274.6 l

1100 58.67 63.78 268.5 275.5 1200 58.87 64.43 268.8 276.4 1500 -

57.04 -

266.1 1800 -

50.85 -

256.4 1.5 psia should be added to this value to cover the max allowable Tech. Spec.

Normal Operation Pressure

" RBCU Consequence Analysis NOTE: Reactor Building design pressure is 57.0 psig, 71.7 psia

• \

'.* l TABLE 5

. Case Camparison - LOCA lient Transfer Coeff. Vapor / Sump Temperature (*F) to Passive Sinks Pressure (RH Sprays (RH Sprays

. Time (BTU /IIr - ft 2 *F) (PSIA) end at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) end at i hours)

(Sec.) Licensfrg RHCU* Licensing RHCU* Licensing RHCU* RBCO*

5 139.1 135.4 40.85 40.66 233.2/192 232.9/191 232.9/191 10 195.7 194.2 51.73 51.68 254.7/211 254.7/211 254.7/211 15 239.3 237.9 56.10 56.29 262.0/218 262.2/219 262.2/219 l 20 228.2 94.5 54.59 55.45 259.4/220 261.0/220 261.0/220 40 118.5 93.9 53.21 55.14 257.2/223 260.4/222 260.4/222 60 80.6 94.1 53.83 55.23 258.2/224 260.5/224 260.5/224 100 63.2 94.5 55.17 55.48 260.3/228 260.9/228 260.9/228 1

200 63.4 97.1 57.41 57.27 263.8/233 263.8/234 263.8/234 300 65.9 101.7 59.11 59.42 266.3/241 266.9/241 266.9/241

] 350 65.8 103.1 59.36 60.00 266.7/244 267.9/245 267.9/245 550 63.4 101.6 57.60 59.38 264.0/254 266.8/255 266.8/255 950 58.9 98.1 54.6 b 58.00 259.3/263 264.8/265 264.8/265 1880/2000 41.9 78.5 43.38 47.04 238.3/261 246.3/263 246.3/263 3880/3500 22.5 45.3 30.40 34.19 202.0/229 1

218.9/246 218.9/'446 1 5020/5000 21.4 52.3 29.72 36.66 199.4/236 224.4/244 224.4/244 10000 17.3 $ 1.5 26.76 36.28 187.0/223 221.9/237 213.1/242 20000 13.0 44.3 23.74 33.55 4

171.4/198 214.2/221 156.4/231 I

! 40000 10.5 36.8 21.89 30.23 159.4/183 203.1/209 122.3/217 s

54000 - -

21.51 29.39 156.6/178 200.0/204 118.5/211 1

60000 9.7 -

21.28 23.44

] 154.9/175 172.4/193 116.8/206 i

86000 - - 20.20 20.01 146.0/160 148.3/163 114.5/114 I 90000 6.4 -

18.85 18.33 133.8/164

! 132.9/166 108.8/177 l 1.1 + 5 - - 18.82 18.89 132.8/170 150.7/172 118.5/174 i

1.4+5 6.4 -

18.81 18.80 132.6/168

.; 147.8/169 118.5/169 l, i 1. 9 + 5 6.4 -

18.81 18.58 132.4/161 1

140.8/162 115.6/162 l 2.0 + 5 - . 18.81 18.54 3 132.4/160 139.5/161 115.1/160 f

I 5.0+5 -

18.60 129.9/145 -/- -/- 1 4

1.0 + 6 -

18.28 123.4/134 -/- -/- I

• - RBCU Consequence Analysis I

l NOTE: Reactor Building design pressure is 57.0 psig,71.7 psia r

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