ML19347D904
| ML19347D904 | |
| Person / Time | |
|---|---|
| Site: | Zimmer |
| Issue date: | 04/10/1981 |
| From: | CINCINNATI GAS & ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19347D902 | List: |
| References | |
| NUDOCS 8104140431 | |
| Download: ML19347D904 (100) | |
Text
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U ZPS-1-MARK II DAR AMENDMENT 14
< APRIL 1981 WM. H. ZIMMER POWER STATION
(
INSTRUCTIONS FOR UPDATING YOUR DJSIGN ASSESSMENT REPORT Changes to the MARK 11 DAR are identified by a vertical line in the right margin of the page. To update your copy of the ZPS-1 DAR, remove and destroy the following pages and figures and insert pages and figures as indicated.
REMOVE INSERT TABLE OF CONTENTS Pages vii and viii Pages vii and viii Page ix Page ix Page xii Page xii Page xvi Pages xvi and xvii CHAPTER 2.0 Page 2.1-2 Page 2.1-2 CHAPTER 5.0
(. >
Page 5.2-1 Page 5.2-1 Pages 5.4-2 through Pages 5.4-2 through 5.4-9 5.4-12 CHAPTER 6.0 Page 6.1-4 Page 6.1-4 Page 6.4-3 Page 6.4-3 CHAPTER 7.0 Page 7.1-4 Page 7.1-4 Page 7.1-7 Page 7.1-7 Page 7.1-12 Page 7.1-12 Page 7.3-1 Pages 7.3-1 and 7.3-la Page 7.3-8 Page 7.3-8 After Figure 7.3-3 Figure 7.3-4, Sheets 1 and 2 of 2 CHAPTER 8.0 Pages 8.1-1 through Pages 8.1-1 through 8.1-5
, 8.1-5 l
Pages 8.2-1 through Pages 8.2-1 through 8.2-20 8.2-5 Pigures 8.2-1 and 8.2-2 Figures 8.2-1 through 8.2-13 1
! m,y,% 1 P00R ORIGINAL
ZPS-1-MARK II DAR
- JfENDMENT 14
.LPRIL 1981
, ( REMOVE INSERT CHAPTER 8.0 (Cont'd)
After Figure 8.2-13 Pages: Title Page for Attachment 8A, pages
! 8A-1 through 8A-5, and Figures 8A-1 through 8A-8; Title Page for Attachment 8B and pages 8B-1 through 8B-3; Title Page for Attachment 8C and pages 8C-1 through 8C-3; 2 Title Pages for Attachment 8D and pages 8D-1 through 8D-14; Title Page for Attachment 8E and pages 8E-1 and 8E-2; and Title Page for Attachment 8F and Figure 8F-1.
APPENDIX B Page B-iii Page B-iii After page B-38 (Figure Page B-38a QO20.62-1)
APPENDIX G Pages G.5-1 and G.5-2 Pages G.5-1 and G.5-2
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P00R ORIGINAL 2
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 1
TABLE OF CONTENTS (Cont'd)
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v PAGE 7.6.2 Reevalation for the Reactor Pressure Vessel Supports and Internal Components 7.6-4 8.0 SUPPPESSION PCOL WATER TEMPERATURE MONITORING SYSTEM 8.1-1 8.1 SYSTEM DESIGN 8.1-1 8.1.1 Safety Design Basis 8.1-1 8.1.2 General System Description 8.1-1 8.1.3 Normal Plant Operation 8.1-2 8.1.4 Abnormal Plant Operation 8.1-3 l14 8.1.4.1 Plant Transients 8.1-3 8.1.4.2 Abnormal Events 8.1-3 8.1.4.3 Primary System Isolation 8.1-3 8.1.4.4 Stuck-Open Relief Valve 8.1-4 8.1.4.5 Automatic Depressurization System (ADS) 8.1-5 14 8.1.5 Transients of concern 8.1-5
- 8. 2 SUPPR ESSION POOL TEMPERATURE REPONSE 8.2-1
() 8.2.1 Introduction 8.2.1.1 System Description 8.2-1 8.2-1 8.2.1.2 Background,-Response to NRC Request 8.2-2 8.2.1.3 Conservatisms 8.2-3 8.2.2 Temperature Response Analysis 8.2-5 8.2.2.1 Model Description 8.2-5 8.2.2.2 General Assumptions and Initial Conditions 8.2-6 8.2.2.3 Description of Non-LOCA Events 8.2-7 8.2.2.3.1 SORV at Power 8.2-7 8.2.2.3.2 Isolation / Scram 8.2-8 8.2.2.4 Small Break Accident 8.2-8 8.2.3 Results/ Conclusions 8.2-8 8.2.4 Summary 8.2-9 L4 ATTACHMENT 8A 8A-1 ATTACHMENT 8B MAIN COFDENSEP 8B-1 ATTACHMENT 8C MANUAL SCRAM 8C-1 ATTACHMENT 8D EMEPGENCY OPERATING PROCEDURE 8D-1 ATTACHMEMT 8E FSAR CHAPTER 15.0 EVENTS VERSUS LONG TEFM POOL TEMPERATURE EVENTS 8E-1 4
C)
(
ATTACHMFNT 8F SUPPRESSION POOL TEMPERATURE vs.
MASS FLUX 8F-1 vii
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981
/^) TABLE OF CONTENTS (Cont'd)
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PAGE 9.0 PLANT MODIFICATIONS AND RESULTANT IMPROVEMENTS 9.1-1 9.1 STRUCTURAL MODIFICATIONS 9.1-1 9.2 BAL ANCE-OF- PLANT (FO P) PIPING AND EQUIPMENT 9.2-1 9.2.1 BOP Piping 9.2-1 9.2.1.1 Drywell Piping 9.2-1 9.2.1.2 Wetwell Piping 9.2-1 9.2.1.3 BOP Piping 9.2-2 9.2.2 Equipment 9.2-2 9.3 NSSS PIPING AND EQUIPMENT 9.3-1 9.4 SRV DISCHARGE QUENCHER 9.4-1 10.0 PLANT SAFETY MARGINS 10.1-1 10.1 CONSERVATISMS IN PLANT DESIGN 10.1-1 fg 10.1.1 Conservatisms in Pool Dynamic Loads 10.1-1 sj 10.1.2 Structural Conservatisms 10.1-2 10.1.3 Mechanical Conservatisms 10.1-2 10.1.3.1 Conservatisms in BOP Piping Analysis 10.1-2 10.1.3.2 Conservatisms in BOP Equipment 10.1-4 10.1.4 Conservatism in NSSS Design 10.1-4
11.0 CONCLUSION
S 11.0-1 APPENDIX A COMPUTER PROGRAMS A.1-1 APPENDIX B MRC QUESTIONS WITH RESPONSES B.1-1 APPENDIX C SOIL-STRUCTURE INTERACTION MODEL C.0-1 APPENDIX D MASS ENERGY RELEASE METHODOLOGY D.1-1 APPENDIX E MASS OPERABILITY ACCEPTANCE E.1- 1 APPENDIX F LINE INVENTCRY MODEL - MAIN STEAM-LINE PUPTURES F.1-1 APPENDIX G SUBMEPGED STRUCTUPE METHODOLOGY G.1-1 APPENDIX H T-QUENCHER EEEVALUATION FOR PIPING SYSTEMS H.1-1 ;
(*h L.)
viii 14
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981
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(
) THE WM. H. ZIMMER NUCLEAR POWER STATION - UNIT 1 MARK II DESIGN ASSESSMENT REPORT LIST OF TABLES NUMBER- TITLE PAGE 2.1-1 Plant Modifications 2.1-7 114 2.2-1 Piping Acceptance Criteria 2.2-7 2.2-2 Load Combinations and Acceptance Criteria 2.2-8 2.2-3 Drywell Piping Assessment: Comparison of Piping Support Load Magnitudes (UPS ET-B) 2.2-10 2.2-4 Drywell Piping Assessment: Comparison of Piping Support load Magnitudes (EMERGENCY-C) 2.2-11 2.2-5 Drywell Piping Stress Assessment 2.2-12 2.2-6 Piping Overstress 2.2-13 2.2-7 Piping Stress Summary 2. 2- 14 2.2-8 Load Combinations Evaluated for the Wetwell Piping 2.2-15 2.5-1 Summary of Load Cases for Equipment for Study Purposes 2.5-3 2.5-2 Load case Definitions 2.5-4
(~s 2.5-3 Feview of Previous Pesults 2. 5- 5
(-) 2.5-4 NSSS Safety-Related Components Assessed 2.5-6
- 2. 5- 5 NSES Safety-Related components Assessed 2.5-7 2.5-6 Zimmer Main Steam System Calculated Snubber Loads 2.5-8 2.5-7 Zimmer Fecirculation System Calculated Snubber Loads 2.5-9 3.2-1 List of Equipment Peing Monitored During In Situ SEV Test 3.2-2 3.3-1 Test 'datrix 3.3-3 3.3-2 Test Matrix - Definition of Abbreviations and Footnotes 3.3-5 4.0-1 Prinary Containment Principal Design Parameters and Characteristics 4.0-2 5.2-1 SRV Discharge Line Clearing Transient Parameterization 5.2-15
- 5.2-2 SRV Eubble Dynamics Parameterization 5. 2- 16
! 5.2-3 Transient Analysis Assumptions 5. 2- 17 5.2-4 Relief Valve Inputs - Zimmer Analysis 5. 2- 18 5.2-5 Zimmer Transients Pesults 5. 2- 19 5.3-1 Acoustic Loading on Peactor Pressure vessel Shroud 5. 3- 24 5.4-1 Zimmer Position on NFC Lead Plant Acceptance Criteria (NUREG-0487 14 l
and NUPEG-0487, Supplement No. 1) 5.4-2
( [) 6.1-1 Design Load Combinations 6.1-4 1-P00R ORIGINAL
ZPS-1-MARK II DAR ~ AMENDMENT 14 APRIL 1981 TABLE OF CONTENTS (Cont'd)
)
PAGE 7.6-3 Class 1E Control Par.els and Local Panels and Packs Seismic Qualification Test Summary 7.6-8 8.2-1 Pool Temperature Analysis Results 8.2-12 8.2-2 Important System Characteristics 8. 2-13 8.2-3 Pool Temperature Conditions - Case la 8.2-15 8.2-4 Pool Temperature Conditions - Case 1b 8. 2- 16 14 8.2-5 Pool Temperature Conditions - Case 2a 8. 2- 17 8.2-6 Pool Temperature Conditions - Case 2b 8. 2- 18 8.2-7 Pool Temperature Conditions - Case 3a 8. 2- 19 8.2-8 Pool Temperature Conditions - Case 3b 8.2-20 i
O 4
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xii
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981
(~} TABLE OF CONTENTS (Cont'd)
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7.1-26 Containment Wall Post-Tensioning Layout 7.1-27 Containment Wall Feinforcing Layout 7.1-28 Reactor Support Concrete Plug 7.1-29 Reactor Support - Feinforcing Layout Before Modification 7.1-30 Drywell Floor Feinforcing Layout 7.1-31 Drywell Floor Column Reinforcing Layout 7.1-32 Design Sections - Primary Containment, Reactor Containment, Feactor Support, and Basemat 7.1-33 Design Sections - Crywell Floor 7.1-34 Design Sections Drywell Floor Column 7.1-35 Typical Interaction Diagram for Basemat 7.1-36 Typical Interaction Diagram for Containment
- 7. 2- 1 Basemat Liner Detail 7.2-2 containment Liner Cetail 7.3-1 Downcomer in the Suppression Pool 7.3-2 Downcomer Bracing Layout 7.3-3 Connection of Bracing to Downcomer 7.3-4 Embedment Plate 7.5-1 Embedment Load Change Inside Containment for N + CO f~s (DFFF) + SSE
(_) 7. 5- 2 Embedment Load Change Inside Containment for N + CHUG
+ SRV SSE TQ 7.5-3 Embedment Load Change Inside Containment For N + CO (Empirical) + SRVTQ + SSE 7.5-4 Embedment Load Change Cutside Containment for N + CO (DFFF) + SSE 7.5-5 Embedment Load Change Outside Containment For N + CC (Empirical) + SRVTO + SSE 7.5-6 Embedment Load Change Outside Containment for N + CO (Empirical) + SRVTQ + SSE 7.6-1 Design and Evaluation Flow 8.2-1 Wm. H. Zimmer Nuclear Power Station 8.2-2 Residual Heat Removal System 8.2-3 Residual Heat Pemoval System Containment Cooling Mode 8.2-4 Residual Heat Femoval System Shutdown Cooling Mode 8.2-5 Residual Heat Removal System Hot Standby Mode 8.2-6 Residual Heat Removal System Low Pressure Coolant Injection Mode 14 8.2-7 Feedwater System (FW) 8.2-8 Pool Temperature Response - Case la SORV at Full Power, 1 PHF Available
- 8. 2-9 Pool Temperature Response - Case 1b SORV at Full Power, 2 RHR's Available 1 8.2-10 Pool Temperature Response - Case 2a Isolation / Scram, i 1RHR Availatle )
/~ 8.2-11 Pool Tenperature Response - Case 2b Isolation / Scram, I
\}
~/ 2 FHB's Available )
xvi l
l
ZPS-1-MARK II DAR AMENDMENT 14 !
-APRIL 1981 i i
! TABLE OF CONTENTS (Cont'd) ,
O '
4 8.2-12 Pool Temperature Response - Case 3a SBA, 1 PHR
, Available 14 8.2-13 Pool Temperature Response - Case 3b SBA, 2 RHR's Available ,
j 9.4-1 .T-quencher Discharge Device :
i 9.4-2 Plan Location of T-quenchers i
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Xvii
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 conservatively _ based on existing steam condensation data.
(~,3j Because condensation oscillation occurs over a wide range of blowdown conditions, two CO loads were defined. The first is a high mass flux CO load which would correspond to the early portion of a large break LOCA. The main components of this 14 load are defined as:
- a. Sinusoidal Pressure Fluctuations 4.5 psi @ 2-7 Hz 2.2 psi @ 11-13 Hz
- b. Random Pressure Fluctuations 4
Steam Bubble Collapse: 15-50 Hz The 2 to 7 hertz component specified represents an increase of about 20% over the DFFR/NUREG-0487 load. The 11 to 13 hertz component is an additional load to account for any vent acoustic effects. The higher frequency portion of the load is added to bound random high frequencies which may appear in test data.
At lower mass fluxes there may be a possibility of a higher contribution from the vent acoustic effect with a corresponding decrease in the low frequency component. The main components 14 of this load are defined as:
(s_'T
/
- a. Sinusoidal Pressure Fluctuations i 2.2 psi 2-7 Hz i 3.8 psi 11-13 Hz
- b. Random Pressure Fluctuations Steam Bubble Collapse: 15-50 Hz The 2 to 7 hertz component here is 50% of the low frequency component used in the high mass flux load while the vent acoustic amplitude has been conservatively assumed to be even higher than the amplitude specified in the lower 2 to 7 hertz range in the DFFR. The higher frequency load is defined as described above.
The Zimmer Empirical Condensation Oscillation Load bounds the requirements of the NRC Lead Plant Acceptance Criteria (NUREG-0487), as demonstrated by Figure 2.1-1.
2.1.4 Chuacino Chugging loads are divided into two areas. The chugging lateral load is the self loading of the downcomer vent during chugging and affects the design of the downcomers, bracing, and drywell
() floor. The chugging event also generates a hydrodynamic load which loads the submerged boundaries of the suppression pool.
2.1-2
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 g
()_
5.2 SAFETY / RELIEF VALVE (SRV) LOADS - PRESENT DESIGN LOADS (T-0UENCHERS)
Actuation of safety / relief valves (SRV) produces direct transient loads on components and structures in the suppression chamber region and the associated structural response produces transient loadings on piping systems and equipment in the ccatainment region and reactor building. These transient SRV loadings are discussed in the following subsections.
Prior to actuation, the discharge piping of an SRV line contains atmospheric air and a column of water corresponding to the line submergence. Following SRV actuation, pressure builds up inside the piping as steam compresses the air in the line. The resulting high-pressure air bubble that enters the pool oscillates in the pool as it goes through cycles of overexpansion and recompression. The bubble oscillations resulting from SRV actuation and discharge cause oscillating. pressures throughout the 14 pool, resulting in dynamic loads on pool boundaries and submerged structures. These dynamic loads cause a dynamic structural response sufficient to affect piping systems and equipment in the containment and reactor buildings. The assessment of the affected systems for these responses is discussed in Chapter 7.0.
Steam condensation vibration phenomena can occur if high-
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'l pressure, high-temperature steam is continuously discharged at high-mass velocity from rams head devices into the pool, when the pool is at elevated temperatures. This phenomena is 14 mitigated by maintaining a low pool temperature as discussed in Chapter 8.0 and by installing quencher discharge devices.
The characteristics of the SRV actuation load vary depending on the piping configuration and the discharge device (rams head or quencher) located at the exit of the SRV line. Typically, the quencher device produces lower dynamic loads. Zimmer rower Station used a bounding load calculated for a rams head device as an original design basis for structures, equipment, and piping systems. A bounding quencher load is now used. To provide increased plant safety margins for containment SRV loads and to increase the threshold temperature limit for steam condensation vibration, SRV quencher devices are installed in the plant.
Pool temperature transients for several postulated cases involving a stuck-open safety / relief valve are presented in Section 8.2. The calculated maximum pool temperature was calculated to be a few degrees below the threshold temperature limit for steam condensation instability for a rams head discharge device.
In order to increase the margin between the calculated maximum temperature and this threshold temperature limit, it was decided t to install a quencher device having a higher suppression pool 5.2-1
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MARK II OWNERS GROUP LOAD OR PHENOMENON LOAD SPECIFICATION NRC REVIEW STATL'S ZIMMER POSITION ON ACCEPTANCE CRITERIA d) Pool Swell Acceleration predicted by the PSAM. Acceptable (1) Acceptable Acceleration Pool acceleration is utilfred in the calculation of acceleration drag during pool swell.
e) Wetwell Air Wetwell air compression is cal- Acceptable (1) Acceptable Compression culated by the PSAM. Defines the pressure loading on the wetwell boundary above the pool surface during pool swell, f) Drywell Pressure Plant unique. Utilized by PSAM Acceptable if based Acceptable History to calculate pool swell loads. on NEDM-10320. other-wise plant unique 14 reviews required. (1)
- 2. Loads on Sugmerged Maximum bubble pressure predicted Acceptable (1) Acceptable Boundaries by the PSAM added uniformly to local hydrostatic below vant exit (wells and basemat) linear attenua- g tion to pool surface. Applied to @
walls up to maximum pool swell J.
elevation.
- v. h' M
- ,. 3. Impact Loads
- d. ::
a) Small Structures 1.5 x Pressure-Velocity correla- NRC criteria I.A.6 (1) Acceptable ey tion for pipes and I beams. g Constant duration pulse.
b) Large Structures None - Plant unique load where Plant unique review Acceptable. Zimmer has no large structures in the pool applicable, where applicable, swell zone.
c) Grating No lepact load specified. Pdrag NRC Criteria I.A.3 (1) Acceptable. Zimmer has no grating in pool swell area.
vs. open area correlation and velocity vs. elevation history from the PSAM.
- 4. Wetwell Air Compression a) Wall Loads Direct application of the PSAM Acceptable (1) Acceptable calculated pressure due to wet-well compression.
b) Diaphragm Upward 2.5 psid NRC Criteria I.A.4 (1) Acceptable gg
,m Loads pg-3 I
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/
! 8
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TABLE 5.4-1 ' Cont'd)
MARK II OWNERS CROUP LOAD SPECIFICATION NRC REVIEW STATUS ZIMMER POSITION ON ACCEPTANCE CRITERIA
@ _OR PHENOMENON Use 10% of the maximum bubble NRC Criteria in Acceptable
- 5. Asymmetric Load pressure statically applied to Section II.A.3 of of the submerged boundary Supplement I to (GE letter MFN-076-79, March NUREG-0487 (2) 16, 1979).
C. Steam Condensation and Chugging Loads 14
- 1. Downcomer Lateral Loads 8.8 kip static NRC Criteria I.B.1 (1) Acceptable a) Single Vent Loads Prescribes variation of load NRC Criteria I.B.2 (1) Acceptable b) Multiple Vent Loads per downcomer vs. number of downcomers.
- 2. Submerged Boundary Loads Sinuaoidal presaure fluctuation Acceptable (1) Acceptable. As described in the discussion of the Zimmer h a) High Steam Flux Empirical Loads approach in Chapter 2.0, more conservative s Loads added to local hydrostatic. 5' loads have been used for assessment and redesign.
w' a-Amplitude uniform below vent exit-linear attentuation to pool (
M e surface. 4.4 psi peak-to-peak g amplitude. 2 to 7 Hz frequencies.
b) Medium Steam Flux Sinusoidal pressure fluctuation Acceptable (1) Acceptable f Loads added to local hydrostatic. Am-plitude uniform below vent exit-linear attenuation to pool surface.
7.5 psi peak-to-peak amplitude.
2 to 7 Hz frequencies.
Representative pressure fluc- Acceptable pending Acceptable. Conservatism inherent in the use of 4T data c) Chugging Loads is discussed in Section 3.3.1.1.6.
tuation taken from 4T test resolution of FSI added to local hydrostatic. concerns.
- uniform loading Maximum amplitude uniform below condition vent exit-linear attenuation to pool surface, +4.8 psi maximum overpressure, -4.0 psi maximum underpressure, 20 to 30 Hz frequency.
mm Zimmer specific (Lead Plant) Eg chugging load criteria (based u. m on 4T test results) submitted $%
July 1980. Revision 1 to the **
r.
Lead Plant chugging report was submitted reptember 1980.
/-.
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TABLE 5.4-1 (Cont'd)
MARK 11 OWNERS CROUP 14AD OR PHENOMENON LOAD SPECIFICATION NRC REVIEW STATUS ZIMMER POSITION ON ACCEPTANCE CRITERIA
- asymmetric loading Maximum amplitude uniform below condition vent exit-linear attenuation to pool surface. 20 psi maximum overpressure. -14 pai maximum underpressure, 20-30 Hz fre-quency, peripheral variation of amplitude follows observed statistical distribution with I' maximum and minimum dia-metrica11y opposed.
- 11. SRV-Related Hydrodynamic Loads A. Pool Temperature Limits for None specified NRC Criteria 11.1 Acceptable KWU and GE four-ara quencher and 11.3 (1)
Quencher Air Cleaning Loads The methodology documented in NRC Criteria in The Zimmer station is being assessed for the T-quencher the Susauchanna DAR is used Section 11.B.5 of loads. These loads are considered to be conservative for the T-quencher load Supplement 1 to and demonstrate the adequacy of the Zimmer design. A definition. NUREG-0487 (2) presentation on the impact of modifications to the SRV y frequency range was given in the February 13, 1979 g meeting. Results of an assessment of the SRV T-quencher j, frequency range was presented at the July 26, 1979
- v.
meeting. The conservatism of the T-quencher load in both h b amplitude and frequency is described in Subsection 5.2.2.1 M O' %
A further demonstration of the conservatism of the lead a plant approach has been documented by Long Island g; Lighting Co. (SNRC-374 March 30, 1979, Mr. Novarro
[LILC0]toMr.S.A.Varga[JRC]transmittingareport entitled " Justification of Mark 11 Lead Plant SRV Load Definition.")
In-plant tests will be run to demonstrate the adequacy and conservatism of the design loads.
B. Quencher Tie-Down Loads
- 1. Quencher Arm Loads Vertical and lateral arm loads Acceptable Acceptable a) Four-Arm Quencher developed on the basis of bound-ing assumptions for air / water discharge from the quencher and conservative combinations of maximum / minimum bubble pressure gg acting on the quencher. gm ra h
b) KWU T-Quencher T-qeencher arm loads, as specified Acceptable Acceptable. These loads have been calculated using the gp for SSES. (3) sv (hodology and assumptions aescribed in DFFR for four-arm $ 4 ochers, as recommended in the Acceptance Criteria. KWU g s uencher methods were used to verify conservatism.
_ -_ _ _ . ~_._ -.
%- , ,1 TABLE 5.4-1 (Cont'd) .
MARK II OWNERS GROUP LOAD SPECIFICATION NRC REVIEW STATUS ZIMMER POSITION ON ACCEPTANCE CRITERIA LOAD OR PHENOMENON
- 2. Quencher Tie-Down Loads Includes vertical and lateral Acceptable Acceptable a) Four-Arm Quencher arm load transmitted to the base-mat via the tie-downs. See 11.C.1.a above plus vertical transient wave and thrust loads.
Thrust load calculated using a standard momentum balance. Ver-tical and lateral moments for air or water clearing are cal-culated based on conservative clearing assumptions.
T-quencher tie-down loads, as Acceptable Acceptable. These loads have been calculated using the g4 b) KWU "T"-Quencher methodology and assumptions described in DFTR for four-specified for SSES. (3) are quenchers, as recommended in the Acceptance Criteria.
KWU T-quencher methods were used to verify conservatism.
III. LOCA/SRV Submerged Structure loads y
A. LOCA/SRV Jet Loads vs Methodology based on a quast-one- NRC Criteria in Acceptable. Rams head - N/A. For LOCA jet see f
- 1. LOCA/ Rams Head SRV Subsection 5.3.2.1. g y* Jet Loads dimensional model. Section II.C.1 of >
Supplement 1 to
- e
$ NUREG-0487. (2) C No loads specified for lead plants. NRC Criteria III.A.2 (1) Acceptable. The spherical zone of influence defined in ts
- 2. SRV-Quencher Jet the Acceptance Criteria is not appropriate for the two- N Loads Model under development in long-term program.
arm quencher. A zone of influence for each arm will be defined as a cylinder with an axis coincidental with the quencher arm. The length of the cylinder will be equal to the length of the quencher arm plus 10 end cap hole diameters. The radius'of the cylinder is expected to be cuite small. However, because no structures are within 5 feet of the quencher arm. 5 feet will be assumed.
B. LOCA/SRV Air Bubble Drag Loads
- 1. LOCA Air Bubble Details of methodology are included NRC Criteria 2 Acceptable Loads in Appendix C. Section II.C.2 of Supplement I to NUREG-0487. (2)
- 2. SRV-Rams Head Air The methodology is based on an NRC Criteria 111.8.2 (1) Acceptable. Used in Item III.B.3. gg wm Bubble Loads analytical model of the bubble Standard drag is calculated and included for all sub- E Ei charging process including bubble
- rise and oscillation. Accelera- merged structure load calculations. Other submerged tion drag alone is considered. structure concerns in NUREG-0487 are addressed in $
>. h Appendix C. "" g.
e
. = . . . - - . _ . -
(1 s-
. (q; v
TABLE 5.4-1 (Cont'd)
MARK 11 OWNERS CROUP IAAD OR PHENOMENON LOAD SPECIFICATION NRC REVIEW STATI'S ZI M ER POSITION ON ACCEPTANCE CRITERIA
- 3. SRV-Quencher Air No quencher drag model provided NRC Criteria III.B.3 (1) The bubble location and radius will be defined Bubble 1.oads for lead plants. Lead plants appropriately for T-quenchers. Bubbles are located near the arms. The bubble size is predicted from the propose interim use of rams head model (See 111.B.2 above). line air volume.
Model will be developed in long-term program.
I C. Steam Condensation Drag No generic load methodology load plant load spe- Described in Subsections 5.3.1.3.5 and 5.3.1.3.6.
toads provided. Generic model cification and NRC under development in long- review.
term program.
IV. Secondary loads A. Sonic Wave Load Negligible load - none specified Acceptable B. Compressive Wave load Negligible load - none specified Acceptable No generic load provided Plant unique load Described in Zimmer Closure Re' port C, Post Swell Wave Load specification and NRC review.
D. Seismic Slosh Load No generic load provided Plant unique load Described in Zimmer Closure Report h specification and a we NRC review.
M E. Fallback Load on Negligible load - none specified Acceptable Submerged Boondary g
F. Thrust Loads Momentum balance Acceptable l G. Friction Drag Standard friction drag calculations Acceptable Loads on Vents H. Vent Clearing loads Negligible load - none specified Acceptable wm NOTES: (1) Per NUREC-0487 (NRC Mark 11 Lead Plant Acceptance Criteria)
(2) Per NUREG-0487. Supplement 1 El gm a (3) Per Fusquehanna Steam Electric Station (SSES) Design Assessment Report (DAR). cm 4
(j [Q, N (,/
TABLE 5.k-1 (Cont'd)
MARK II OWNERS CROUP IDAD SPECIFICATION NRC REVIEW STATUS ZIMMER POSITION ON ACCEPTANCE CRITERIA LOAD OR PHENOMENON Interim technical Acceptable, Rodabaugh criteria may be used in some cases if FUNCTIONAL position (7/19/78) NRC finds acceptable.
CAPABILITY 4
Verjfy using RELAP / Acceptable MASS-ENERGY RELEASE FOR !OD ANNULUS PRESS.
15% peak broadening Acceptable QUESTIONS MEB-2, MEB-5 to be used.
Closely spaced modes Acceptable. NSSS scope uses modified sususation MEB-3, MEB-5 combined Per 1.92 per approved CESSAR.
Dynamic analysis Acceptable ME8-1 g methods acceptable 3'
OBE Damping - Level ,
w MEB-2 ACCe ptable SSE Damping - Level C or D E
Seismic slosh plant Acceptable a MEB-6 $
unique review load Combinations: Acce pt able. See load combination table for Case #2 and #7.
MEB-7a and b AP+SSE OBE+SRV Functional capability See load combination table.
ME8-8 and piping acceptance criteria 4$
b
!E
TABLE 5.b-1 (Cant'd)
MARK 11 OWNERS CROUP LOAD PHENOMENON IDAD SPECIFICAT10ft NRC REVIEW STATUS ZDtMER POSITION ON ACCEPTANCE CRITERI A
' l. N+SRV To B Acceptable
- 3. N+SRV,}g+SSE to C Acceptable
- 4. N+SRV +0BE+15A to C Acceptable 9 ads
< 5. N+SRV +0BE+1BA to C Acceptable
- 6. Acceptable N+SRVads+SSE+1BA to C
- 7. N+SSE+DBA to C Acceptable
- 8. N to A Acceptable N
- 9. N+0BE to B Acce ptable 3 A
.* Applied to containment structure only (See M 02G.22 and !
N+SRV +SSE+DBA to C
, 7
- 10.
- DFFR 5.2.4.) h 1
n N
I b i
1 I
l
="s g4 i ;
\w) \v.- ) (w._
TABLE 6.1-1 DESIGN LOAD COMBINATIONS
- ASY'yET-LOAD D L FO Pg R E Egg Pg Pg Tg Rg Rg SRV ADS ALL RICAL SINGLE EQN COND Tc Normal 1.5 0 X X I w/o Temp 1.4 1.7 1.0 1.0 - - - - - - - - -
2 Normal X w/ Temp 1.0 1.3 1.0 1.0 1.0 1.0 - - - - - - -
1.3 0 X 3 Normal Sev. Env. 1.0 1.0 1.0 1.0 1.0 1.0 1.25 - - - - - - 1.25 0 X X 1.0 1.25 - 1.0 1.0 - 1.25 X 0 1 4 Abnormal 1.0 1.0 - - - - -
1.25 1.0 1.0 - 1.0 0 0 0 X 4a 1.0 1.0 1.0 - - - - - -
5 Abnormal Sev. Env. 1.0 1.0 1.0 - - -
1.1 -
1.1 - 1.0 1.0 -
1.1 X 0 X 1.0 1.0 1.0 - - -
1.1 - -
1.1 1.0 1.0 -
1.0 0 0 0 X u Sa 7.
, 6 Normal j.
1.0 1.0 1.0 1.0 1.0 1.0 -
1.0 - - - - -
1.0 0 X X 16
- i. Ext. Env. @'
a.
7 Abnormal 1.0 1.0 1.0 - - - - 1.0 1.0 - 1.0 1.0 1.0 1.0 X 0 X ;*
Ext. Env.
7a 1.0 1.0 1.0 - - - -
1.0 -
1.0 1.0 1.0 1.0 1.0 0 0 0 I c, D
s LOAD DESCRIPTION D = Dead Loads Egg = Safe Shutdown Earthquake Live Loads = SBA and IBA Pressure Load D,Dm L =
PB F = Prestressing Loads TA = Pipe Break Temperature Load Eh
>. m TO = Operating Temperature Loads RA = Pipe Break Temperature Reactions $$
Load RO = Operat ing Pipe Reactiot.s "" 7 Pg = DEA Pressure Loads (including all Pg = Operating Pressure Loads p of hydrodynamic loadings)
SRV = Saf ety/ Relief Valve Loads p = Reactions and Jet Forces Due to R
EO = operating Basis Earthquake Pipe Break SBA = Small Break Accident IBA = Intermediate Break Accident
- In any load combinations, if the ef fect of any load other than D reduces iie design forces, it will be deleted from the combination.
2PS-1-MARK II DAR AMENDMENT 14 APRIL 1981 CHUG Chugging load defined between 20 to 30
()
e-hertz; and Condensation oscillation load CO(EMPIRICAL) including higher frequency contributions (up to 50 hertz).
Includes the envelope of two independent CO empirical loads (see Chapter 2.0). 14 6.4.2.1.2 Pipina Systems Outside the Reactor Buildina All essential and safety-related piping systems outside the reactor building considered the effects of seismic loads. The load combinations considered included the following:
LOAD COMBINATION ACCEPTANCE CRITERIA N + OBE Service Level B N + DBE Service Level C where the loads are as defined in Subsection 6.4.2.1.1.
6.4.3 Balance-of-Plant Equipment 6.4.3.1 Loadina Combinations The table below defines the combinations of the normal, seismic, and pool dynamic loads considered in the equipment qualification:
ALLTO and SRVASY TQ)
- b. N + Envelope (OBE and SSE) + Envelope (SRVADS TQ and SRV TO) + CO (Zimmer empirical medium mass flux)ASY
- d. N + Envelope (OBE and SSE) + Envelope (0.6 SRVADS-TQ and SRVASY-TO) + CO (Zimmer empirical high mass flux)* l l
- e. N + /(AP)z + (ssE)z
- See Chapter 2.0 for explanation of condensation oscillation.
6.4.3.2 Acceptance Criteria
- a. Allowable Stress Limits
() 1. ASME Equipment 6.4-3
l 2PS-1-MARK II DAR AMENDMENT 14
> APRIL 1981
- 3. The transfer functions of the response were obtained by the
() computer program FAST. From Equation (1) Tk (u) = k (u) F (u) in which Rk (u) was the Fourier transform of the responses saved in step (2) and F (u) is the Fourier transform of the white noise load used in step (1) of the above.
- 4. For steady-state solution of the harmonic load, by definition from Equation (1), the transfer function itself was the response.
For SRV loads with variable frequency, the transfer functions ! were scanned in the frequency range of the loading. The maximum response could be obtained as the product of the transfer functions and the Fourier transforms of the load, using the FAST program. Response acceleration time histories were also generated to obtain response spectra using the RSG program. T In order to consider a conservative frequency content, (s'J three KWU time history traces reported in the SSES DAR 14 were expanded into longer and shorter time history dura-tions by multiplying the time scales by a factor of 2.0 and 0.9, respectively. In addition, the pressure scale were multiplied by a factor of 1.5 for each of the three traces. l The resulting structural responses to the various SRV T-quencher loads were combined with the other appropriate loads as per the load combinations shown in Table 6.1-1. The margin factors from these load combinations are presented in Tables 7.1-17 through 7.1-24. 7.1.2 Structural Analysis of LOCA Loads The analysis of the structure for the LOCA loads was performed as a set of analyses covering each LOCA related phenomenon l 7.1-4 l l
2PS-1-MARK II DAR AMENDMENT 14 APRIL 1981 (]) Magnitude: t 3.75 psi Frequency: 2 to 7 hertz l The spatial distributions of the condensation oscillation loads are shown in Figure 7.1-13 for rams head design basis and in Figure 7.1-14 for T-quencher design basis, as 2 3.75 psig acting at a frequency 2 to 7 hertz on the basemat, containment, and reactor pedestal. The structural model describea in Subsection 7.1.2.1 was used for the rams head design basis, and the one described in Subsection 7.1.1 was used for the T-quencher design basis. The load was assumed to be harmonic in time, and only the steady-state response was considered as being of interest. For this purpose, frequency response variations were determined for all response components of interest using the computer program FAST, Appendix A, which obtained the complex frequency response by calculation of the discrete Fourier transform of both load and response. The frequency range of 2 to 7 hertz on the frequency response was considered relevant in evaluating the structural response. The resulting structural responses to the condensation g oscillation loads were combined with the other appropriate loads (s~s/ as per the -load combinations shown in Table 6.1-1. The margin factors from these combinations are presented in Tables 7.1-2 through 7.1-24. In, addition to the above CO load (2-7 hertz), an empirical limiting CO load was considered in combination with the T-quencher design basis of the 2PS-1 containment. This load waa intended to be a best estimate of the conservative load specification which resulted from the full-scale condensation oscillation test to be conducted in the 4T facility. All the details for this load are described in Chapter 2.0. 14 This ZPS-1 empirical CO load was incorporated for the T-quencher design basis. The spatlal distributions of this load are shown in Figure 7.1-14. The resulting structural responses to this empirical CO load were combined with the other appropriate loads as per the load combination shown in Table 6.1-1. The margin factors from the load combinations are presented in Tables 7.1-25 through 7.1-28. 7.1.2.4 Chuacing Analysis The chugging loads used in the analysis are described in Section 5.3 and presented in Figure 7.1-15. The finite-element model used in the analysis is described in Subsection 7.1.2. (]) 7.1-7
2PS-1-MARK II DAR AMENDMENT 14 APRIL 1981 () The forces of reactor support margin factor were obtained by analysis using the model described in Subsection 7.1.2.1. Margins shown in Table 7.1-14 for loading conditions 4a, Sa, and 7a on the'drywell floor are for the LOCA effects, including the lateral loads on the downcomers. As.per DFFR Subsection 4.4.6.6, a net upward load of 9 psid acting on the drywell floor has been considered. Margins shown in Table 7.1-14 for loading conditions 1, 2, 3, and 6 on the drywell floor are for all the valves discharge loading which clearly governs the design of the drywell floor rather than j t.ne asymmetric two valve discharge loading. Loading conditions 4, 5, and 7 in Table 7.1-14 include all loads resulting from a small pipe break combined with the loads due to the discharge of all 13 SRV's. This was done for reasons of analytical expediency, since the discharge of all 13 SRV's transmits significantly more energy to the drywell floor than the i 6 valve ADS discharge. Since 2PS-1 can take this higher loading case, the actual ~1oading from the ADS valves was not considered. i For the drag loads on the downcomer, the maximum load described in Section 5.2 was used for all loading combinations which l 14 include SRV. loads irrespective of the discharge mode (ALL, ASYMMETRIC, or ADS). .; ) T-QUENCHER DESIGN BASIS LOAD COMBINATION WITH NRC CO LOAD (DFFR)
- a. Basemat Tables 7.1-17 through 7.1-20
- b. Containment wall Tables 7.1-21 through 7.1-24 LOAD COMBINATION WITH EMPIRICAL LIMITING CO-LOAD
- a. Basemat Tables 7.1-25 and 7.1-26 i
- b. Containment wall Tables 7.1-27 and 7.1-28 i
The marain f ctors were calculated as results of the assessment : based on the NRC acceptance criteria (modified for the T-quencher). All the margin factors were greater than 1.0 except the following cases: (:1 J l 7.1-12 i n- - , ,- ~ ~ - - ,, , , .- - -,- , , , - - -
2PS-1-MARK II DAR ATTACHMENT 14 APRIL 1981 () 7.3 OTHER STRUCTURAL COMPONENTS A reassessment for the additional effects of pool dynamic loads was made for steel and concrete structures in the reactor building including steel framing and galleries, cable pan hangers, conduit hangers, HVAC duct hangers, and concrete slabs, beams and shear walls. The design and analysis procedure for the reassessment was the same as described in Subsections 3.8.3 and 3.8.4 of the ZPS-1 FSAR for the original design of the plant. Load combinations described in Section 6.3 of this report were used for the assessment. 7.3.1 Downcomers and Downcomer Bracing 7.3.1.1 General Description There are 88 downcomers anchored in the drywell floor. The downcomers are also connected to the containment structure by horizontal bracing at elevation 496 feet. Figure 7.3-1 shows the downcomers in the suppression pool, and Figure 7.3-2 shows the layout of the horizontal bracing. The bracing members are attached to embedment plates which are anchored to the containment wall by a system of bolts and backup cap devices. A typical detail of this attachment is shown in Figure 7.3-4. (m) As shown in the Figure 7.3-4, the length of the bolt is such that it does not interfere with the vertical or hoop tendons in place. Leak tightness is restored by providing stainless steel pipe caps over the bolts and stainless steel plate inserts around the embedment plate. A typical detail of connection of bracing to downcomer is shown in Figure 7.3-3. The downcomers and downcomer bracing are subjected to static and dynamic loads due to normal, upset, emergency, and faulted plant operating conditions. The loadings cases were obtained from the DFFR and are identified in detail in Subsection 7.3.1.2. The loading combinations are explained in Subsection 7.3.1.3. The design limits are identified in Subsection 7.3.1.4, and the analytical methods are presented in Subsection 7.3.1.5. ' 7.3.1.1.1 Downcomer Properties The following are the properties of the downcomers:
- a. outside diameter - 25.00 inches;
- b. wall thickness - 0.500 inch; f- c. weight per unit length - 131 lb/ft;
(_) 7.3-1
- .- . .-_ . . _ - _ ~ - . . . . - _ -
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 .
- d. material - SA-516, Grade 60; and Q
i e. damping coefficient - 3%. 7.3.1.1.2 Bracina Properties j The following are the properties of the bracing:
- a. outside diameter - 8.625 inches;
- b. wall thickness - 0.875 inch; i
'l i O a t O 14 7.3-la
O O O TABLE 7.3-1 LOAD COMBINATIONS AND ACCEPTANCE CRITERIA FOR DOWNCOMER AND DOWNCOMER BRACING LOAD NRC LOAD COMBINATION T-QUENCHER CASE ASME STRESS (NUREG-0487) DESIGN-BASIS CRITERIA 1 N+SRV N+SRV X B (UPSET) 2 N+SRVX+0BE MSRV +OM B (UPSM g
?
3 N+SRV + } X Y 14
- 4 N+SRV + CHUG N+SRVADS+IBA(SBA) C (EMERGENCY)
H 5 M RV +0E+ CHUG C (MERGMCY) N+SRVADS+OBE+IBA(SBA) 6 N+SRVADS+
+ AIb A L
7 N+SSE+DBA N+SSE+CO C (EMERGENCY) >> 8 N N A (NORMAL) P5 g
$5 9 N+0BE N+OBE B (UPSET) I
[ 10 N+SRV +SSE+DBA - X CONTAINMENT STRUCTURE ONLY JUSTIFICATION PROVIDED BY GE. e i.
AMENDMENT 14 APRIL 1981 O , l l I l l i 5'- 0" 9" J '-2" 7 " 7" l'-2" 9" _ DRILL ED LEVELING SCREW SEE DET. F (S-46 8) % [I l HOLE (TYR ) A x / , l
" 3 / ~
=
m i A A QP- E6f 49 =i _l r A I o ,
' ' 4 i ,
-l *
/ =
g _t CLEARANCE LINES SEE l - , to NOT E 20 FOR "AS-BUILT " D " O BOLT SPACING SEE 8 b-db--db-db -_ H 1 - j SCHEDULE (S-470) f 6"_ _ __6 " i d. 2I4 'x 42'x 5'-O" (TYP) (TYP) (A-588 GR50) FOR ANCHOR B OLT S SEE DETAIL C P00R ORIGINAL l WM. H. ZIMMER NUCLEAR POWER STATION. UNIT 1 l MARK II DESIGN ASSESSMENT REPORT FIGURE 7.3-4 EMBEDMENT PLATE (SHEET 1 of 2)
AMEN 0 MENT 14 APRIL 1981
; i
~.
N T. HORIZ. TENDONS
.-PRIMARY CONTAINMENT WALL p
- 9. VERT. TENDONS M O EMBECO 636 GROUT AS MANUFACTURED W _
W9 4/I 2" $ DRILLED HOLE ,16" MAXIMUM BY THE MASTER BUILDER CO. MIXING & DEPTH FROM INSIDE FACE OF
.- PRIMARY CONTAINMENT WALL PLACING OF COMPOUNDS SHALL BE ACCORDING TO MANUFACTURER'S ~
RECOMMENDATIONS SEE NOTE 22 '
/ 1 4 2' @ x3 2" i WASHER IMPERMEABLE MATERIAL " .I ( A 588 GR 50 )
WRAPPED 8 TAPED TO BOLT- -G _ j EMBECO 636 GROUT '
"g .
! c-~ ROUGHEN SURFACE
' ' r-L3x2 (3" MAX. UNDER BASE i.)s ,
SEE DET. B 1 (S-468) X = [L . Ib6" HOLE _lN :
# \ p ;__. 3 1, BASE It.
b, l/4" S.S (* ) ^m , :.l LINER llL(f.R )M .' 6, - W- t h U[ :
)M \Y
[ ' _ LEAK TEST - it.2/"d 34 INSIDE FACE OF PRIMARY CHANNEL (l.P.) ! -
; 3/ V 16 f CONTAINMENT WALL l l 4'/ 2" PROJECTED FlELD CUT . tim ',Mr ~ ,Y N ~
AFTER lt. IS INSTALLEDd,- kl N r* Rf/ /// N = ll/16 16 7 l/8 / FILLET WELD MA x:- -: % ' NINCREASED UP TO 1/4 4"$ PIPE (A 312 TYPE 304 ke;j/ '
/
/
OR 316 )(S.S. )(SCH . 40 )- --
/ \
BACK-UP RING (S.S) TYP.)
,', SEE DET. J (S-468 i N gy 6" & PIPE ( A 312 TYPE 304 OR 316 ) / // 'A (S.S)(SCH. 40)(OFFSET AS REQ'D. /p/
TO MAKE WELD }-
/ / '\ TEST BOSS DRILL 8 TAP 7 j p/ FOR I/4 $ PIPE I
j-
-7 Q. I 2/ "$ x 2'- 2" LG. (SA.193 GR. B7) BOLT
/
II. I/"x4/2 2 I
@ ( A 240 TYPE / ' W/ HEAVY HEX. NUTS (SA-194 GR. B7 ) i JAM NUTS 8 HARDENED CIRCULAR 304 OR 316 ) ( S . S . ) a WASHER l>_ /I 2' x 7 '/4 $ ( A 240 TYPE i
304 OR 316 ) (S . S ) -- " SECTION A-A wu. s.zimmen sucte n eowen sT* Tion. unit i MARK 11 DESIGN ASSESSMENT REPORT Q FIGURE 7.3-4 EMBEDMENT PLATE (SHEET 2 of 2)
ZPS-1-MARK II DAR AMENDMENT 14 l APRIL 1981 i CHAPTER 8.0 - SUPPP ESSION POOL HATEP TEMPER ATUPE MONITORING SYSTEM 8.1 SYSTEM DESIGN 8.1.1 Safety Design Basis The safety design basis f or setting the temperature limits for the suppression pool temperature monitoring system are based on providing thc operator with adequate time to take the necessary action required to ensure that the suppression pool temperature will always remain below the pool temperature limit established by the MFC. in analysis of suppression pool temperature transients can be found in Section 8.2. The system design also provides the operator with necessary information regarding localized heatup of the pool water while the reactor vessel is heing depressurized. If relief valves are selected f or actuation, they may be chosen to ensure mixing and uniformity of heat energy injection to the pool. 8.1.2 General System Dcscriction The suppression pool temperature monitoring system monitors the pool water temperature in order to prevent the local pool water 74 temperature from exceeding the pool temperature limit during SRV /7,l discharge and provides the operator with the information
~'
necessary to prevent excessive pool temperatures during a transient or accident. Temperatures in the pool are recorded and alarmed in the main control room. The instrumentation arrangement in the suppression pool consists of 18 local temperature sensors ir individual guide tubes mounted of f the pool walls. The local temperature sensors consist of 18 dual-element, ccpper constantan thermocouples located 1 foot below the low water level. Twelve of the sensors are located off the outer suppression pool wall at azimuths 280, 450, 860, 1170, 1470, 1830, 2170, 2400 2630, 2770, 3250, and 3440 mhe other six are located off the pedestal at azimuths 550, 1420, 2020, 2460, 2980, and 3440 The sensors and readou* devices are assigned to ESS-1 and ESS-2 divisions and local discharge areas are monitored by two sensors, one from each division. This represents a conservative neasurement of local pool water heatup. All instrumentation will be qualified 9eismic Category I. "he time constant of the thermocouple installation will he .o greater than 15 seconds. The difference between reasurement reading and actual temperature will be within 20 E. s \i The display techniaues for monitoring the pool temperature are: e ,-1 P00R ORlGlNAl
i Z PS- 1-f1 AP K II DAR AMENDMENT 14 APRIL 1981
- a. to continuously input to the computer system the measurement made by Element 1 of each of the nine thermocouples in ESS-1 which can be displayed individually or averaged by the computer to display the bulk temperature;
- b. to sequentially record on a multipoint recorder the measurement made by Element 1 of each of the nine thermocouples in FSS-2 at a rate of 5 sec/ point when all nine are below the alarm level, ard at a rate of 1 sec/ point when any of the nine are above the alarm level;
- c. to continously record on a strip-chart recorder the bulk temperature obtained by electrically averaging the nine Element 2 thermocouples of ESS-1; and
- d. to continuously input to the computer system and display on a hardwired indicator the bulk temperature as obtained by electrically averaging the nine Element 2 thermocouples of ESS-2.
Each instrumentation division has the capability of alarming both 14 local and bulk high temperature. The computer system provides temperature readout via CRT/ data logger on demand. The above ('" configuration provides the maximum flexibility for providing redundant pool temperature information to the operator. The quenching of the stram at the quencher discharge forms jet s that heat the water and generate convection currents in the suppression pool. These currents eventually rise and displace cooler water near the pool surface. During ar extended blowdown, a large temperature gradient is ex-pected initially . ear the quencter. After a short time the pool qradients will stabili7( wi th a tulk to local temperature difference of about 100 F The adequacy of the temperature monitoring systcm will l e confirmed Ly the in -plar t SPV testing. 8.1.3 *ormal Plant Oceration The temperature monitoring system is utilized during normal plant operation to ensure that the pool temperature will remain low enough to condense all quantities of steam that may Le released in any anticipated transient or postulated accident. When rams head devices were specified for design, there was an NRC concern that high pool temperature might result in high pool dynamic loads during SEV discharge because of unstable steam condensation. Installat ion of T quenchers has eliminatad this concern. During normal plant operation, the system is ir /? continuous operation recording the suppression pool water Ki temperature with a realout in the main control room. If the pool temperature rises atove normal operating temperatures, an alarm P00R ORIGINAL
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 is actuated in the control room allowing the operat or to take actions as required to maintain pool temperature below the pool temperature limit as described in Section 8.2. 8.1.4 Abnormal Plant Oreration BUR plants take advantage of the large thermal capacity of the suppression pool during plant transients which require relief valve actuation. The discharge of each relief valve is piped to the suppression pool, where the steam is condensed. This results in a pool water temperature increase but a negligible increase in containnent pressure. However, certain events have the potential for subetantial energy addition to the suppression pool and could result in a high local pool temperature if timely corrective action is not taken. When rams head discharge devices are used, test results and operating experience indicate that high magnitude oscillatory loads may occur wrer a high steam mass flux is injected into a pool with local temperature above 1700 F. Although analysis demonstrates that the pool temperature will remair below 1500 F, when the stean mass flux is high enough to cause these loads, T-quenchers have been installed instead of the rams heads to provide additional margin to the pool temperature limits. 14 n (,) 9.1.4.1 Plant "rans ients "his subsection discusses various plant transients which result in SRV disctarges to the euppression pool and which cou.1d possibly lead to high pool temperature. 8.1.4.2 Abnormal Events various events that result in encrgy being discharged to the supp re sa io.- pool via the sa fety/ relief valves are discussed in the folloeing naragrapnr. '40 s t of these transients are of short duration and nave little effect or the suppression pool temperature. 1owever, three events have the potential for substantially high enercy release to the pool that could result in uniesirably high pool temperatures if timely corrective action is not taken. These events are: (1) events that result in the isolation of the plant frcm the main condenser, (2) stuck-oren relief valve, and (3) automatic depressurizat ion sys-em ( ADS) operation. A brief description of each of tLtse events is civon ir the following sursections. 8.1.4.3 Primary System Isclaticn Chen the primary system is isolated from the main condenser, the reactor is scrammed automatically and the stored energy in the vessel internals, fuel relaxation erergy, and decay heat is (~J rejected to the suppression pool. The amount of heat rejected to the pool depends on reactor size, power level, anc primary system e '-' P00R ORIGINAL
ZPS-1 MAPK II CAR A:4ENDMENT 14 APRIL 1981 heat removal capability. This includes condensing type heat exchangers which remove steam directly from the RPV. 8.1.4.4 Stuck-Open Relief Valve The steam flow rate through a safety-relief valve (SRV) is proportional to reactor pressure. One method to terminate erergy input to the pool is to scram the reactor and depressurize the FPV in the event the relief valve cannot be closed. During the energy dump, the pool temperature will increase at a rate determined hy the RPV pressure, flow capacity of the SRV, primary system heat removal system capahility, and suppression pool water beat removal capability. 8.1.4.5 Automatic Depreseurization System (ADS) Activation of ADS result s in rapid depressurization of the PPV hy the opening of a designated numher of saf ety-relief valves. During this transient, the hulk suppression pool temperature ri se s. In a typical care, the FPV is depressurized below 150 psia in about 10 minutes. 14 8.1.5 Transients of corcern r~ There are seven plant depressurization transients that were ( '3) considered as limiting events (with rams heads) for energy released to the suppression pool. These events are numbered 1 through 7 for ease of reference and are described in the following paragraphs: Event 1 is a stuck-open relief valve with the reactor at full powe r. The plant is scrammed and decressurization hegun via the stuck-open valve. The initial pool tenperature is the maximum pool temperature allowed for continuous operation. This is the only event t hat is not truly limiting, since all FMP heat exchanger equipment is considered operational. Even t 2 is identical to Event 1, except that one FHF heat ex-changer is considered uravailahle. The remaining heat exchanger equipment is placed in the suppression pool cooling mode. Event 3 is a stuck-open relicf valve with the reactor isolated f rom the main conder ser and maintained at operating pressure and tempe rat ure . All FF.9 heat exchanger equipment is considered operational in this case, because if one heat exchanger is not available, the reactor may not remain isolated and pressurizcd. The bulk pool temperature assumed in this event is the highest allowed while the reactor is maintaining pressure. Event 4 is a controlled depressurization f rom the plant l condition descrihed in Ivent 3. This event demonst rates the ' (', ability of the plant to safely depressurize through a controlled transient from the limiting conditions. e.,., P00R ORIGINAL l i
Z PS-1-MAPK II DAR AMENDMENT 14 APRIL 1981 / : KJ Event 5 is a controlled depressurization several minutes af ter reactor isolation described in Event 3. One heat exchanger is considered unavailable tecause power operation in that condition is not prohibited. This event demonstates the ability to safely depressurize the plant through a controlled transient with only partial RHR heat exchanger equipment availability. Event 6 is a rapid depressurization of the reactor resulting from actuation of the automatic depressurization system (ADS). ADS 14 actuation takes place at the time the plant is scrammed. Because this event takes place quickly, no heat exchangers are brouaht into operation. This event represents the most rapid energy release to the pool and demonstrates that the plant may be safely depressurized in this manner without use of heat exchangers. Event 7 is also a rapid depressurization via ADS, but with the reactor isolated and pressurized and with a limiting pool temperature condition. All heat exchanger equipmc:nt is placed in the pool cooling made 0.5 hour after scram but is considered unavailable when depressurization begins. ,r )
%,y 0 P00R ORIGINAL 8.1-5
ZPS-1-MARK II DAR AMENDMENT 14 APPIL 1981
- 8. 2 SUPPRESSION PCOL TEMPERATUPE RESPONSE 8.2.1 Introduct ion This submittal is in respcnse to an URC request for suppression pool temperature calculations for specified transients. CGEE has worked with the Mark II Owners Group and the Mass / Energy Subcommittee ta develop a generic approach to this response.
This analysis was performed in accordance with the Mass / Energy
" White Paper" (revision 1) and incorporates a degree of conservrtism that causes the 2immer pool temperature calculations to slightly ( xceed the iulk pool temperature goal of 1900 F in two cases. C 3SE has identified conservatisms and performed a 2immer plant 2nique analysis and sensitivity study. This analysis and study indicates that 2PS-1 will not exceed the desired goal if 1900 F for the postulated events.
The Wm. H. Zimner Station Unit 1 (Z PS-1) takes advantage of the large thermal capacitance of the suppression pool during plart transients requiring safety / relief valve (SPV) actuation. The discharged steam is piped from the reactor pressure vessel (FPV) to the suppressio: pool where it condenses, res ult ing in a temperature increate of the pool water, but negligible increase 14 in the containment tressure. Most transients that result in
- relief valve actuations are of very short duration and have a i
s ~j small effect on the s1ppression pool temperature. However, certain postulated ev(nts with conservative assumptions present the potential for subs antial energy additions to the suppression pool that could result in high pool temperature. See Figure 8.2-1 for a graphical rouresentation. 8.2.1.1 System Cescriotien ZPF-1 uses the residual heat removal (FHR) system to remove heat from the suppression pool. The FHR system consists of three independent loops. Two of these loops are equipped with 100% capacity heat exchangers that are available in the pool cooling, shutdown cooling, containment spray, and steam condensing modes of operation. The PHP heat exchangers are U-tuhe, vertical head-down units which use Ohio River water for tube side coolant. All three PHR loope are available for the low pressure coolant injection (LPCI) mode of operation. See Figurcs P.2-2 through 8.2-6 for a graphical representation of the PHP system. In the pool cooling mode of operation, PHo pump suction is taken from the suppression pool, cycled through the heat exchanaer, and returned to the pool. In shutdown cooling, pump suction is taken f rom the reactor recirculation (FR) piping, cycled through the heat exchanger, and returned to the reactor pressure vessel (rPV) . The steam condensing mode removes steam from the RPV, (~ ; condenses it ir the heat exchanger, and returns it to the PPV via
> the reactor core isolation cooling (RCIC) system. The steam condensing mode is not used in this analysis. The LPCI mode of l e-'
P00R ORIGINAL
Z PS-1-M AFK II DAR AMENDMENT 14 APPIL 1981 ( _j operation is automatically initiated on high drywell pressure or on low FPV water level (- 14 6 inches) . In the LPCI mode, pumo suction is taken from the suppression pool and cycled back to the suppression pool through the minimum flow bypass line until the PPV reaches pressure permissive for injection into the RPV. See Subsection 5. 5.7 of the Zimmer FSAR for the RHP system description. ZPS-1 is equipped with twc steam-driven feedwater cump turbines (Figure 8. 2-7) which receive their steam supply from the main steamlines downstream from the mair steam isolation valves (MSIV's) . When MSIV closure occurs, the steam driving force to the feedwater turbines is eliminated, causing a rapid decrease in the amount of feedwater entering the RPV. This treatment of the feedwater system is not used in this analysis. ZPS-1 feedwater system is described in Subsection 10.4.7 of the ZPS-1 FSA7 8.2.1.2 Background, Fesocnse to NRC Fequest Cincinnati Gas S Electric Company was asked to deFonstrate, for several postulated transients at Zimmer plant, that the MFC pool temperature limit would not be exceeded. The NRC requested figures showing reactor pressure and suppression pool temperature versus tiFe for the f ollowing
'N events:
G
- a. stuck oper SRV during power operation assuming reactor scram at 10 minutes after pool temperature reached 1100 F and all RHR loops available;
- b. same as above except only one RHR trair available;
- c. stuck open SRV during hot standby condition assuming 1200 F pool temperature initially and only one PHP train available;
- d. automatic depressurization system (ADS) activated durino a small line break assuming an initial pool temperature of 1200 F and only one RHF train available; and
- e. the primary system is isolated and depressurized at a rate of 1000 F/ hour with an initial pool temperature of 1200 F ard only one PHR train available.
In addition, CGSE was askad to provide important plant parameters, such as, service water temperature, RHF heat exchanger capability, and initial pool mass for the analysis. gm This submittal of pool temoerature transients answers the NFC _, reauest. The Mark II owncrs Group "Nhite Paper" (Pevision 1) e -2 P00R ORIGINAL
Z PS-1-MAFK II DAR AMEMDMENT 14 APRIL 1981
/ identified the six cases to be analyzed by the Mark II plants.
These are:
- 1. Stuck-Cpen Felief Valve
- a. from power operation with loss of one PHP heat exchanger, and
- b. from power operation with spurious closure of MSIV's.
- 2. Isolation / Scram (SPV Cischarge)
- a. loss of one RHP heat exchanger, and
- h. stuck-oren relief valve.
- 3. Small treak accident
- a. loss of one RHP heat exchanger, and 14 L. loss of shutdown cooling.
Pesults of the analysis are listed in Table 8.2-1, and shown in
,_s Figures 8.2-8 through 8.2-13. A list of import ant plant (v ) parameters is in Table 8. 2-2.
8.2.1.3 Conservatisms Subsection 8. 2. 2. 2 lists all the general assumptions and initial conditions. Aany of the assumptions used in the analysis maximize heat addition to the suppression pool. For example, with the exception of case la, MSIV's are assumed to close 3.5 seconds after scram. This eliminates the condenser as a heat aink and maximizes heat addition to the suppression pool. Justification for maintaining condenser availability in case la can he found in Attachment 9B. This analysis further maximizes heat additior to the RPV by assuming hot feedwater flows into the EPV throughout the transients. In reality, however, MSIV closure eliminates steam supply to the turbine-driven f eedwater pumps which causes termination of feedwater flow into the RPV within seconds. The ECCS pumps would he the source for providing liquid inventory to maintain EPV level on isolation events. Since hot feedwater has a much higher enthalpy than ECcs suction from the condensate a storage tanks or the suppression pool, the obviods difference in i heat addition to the RPV is a conservatism which has a sionificant impact on pool temperature. A sens itivity study on feedwater flow is in Attachment 8A. (_) Using desigr fouling values for the PHP heat exchangers, coupled with the assumption that 5% of the tubes are plugged, plus use of P00R ORIGINAL
ZPS-1-MAFK II DAR AMENDMENT 14 APFIL 1981 design temperature for the service water system greatly reduce tre ef fectiveness of the PHR heat exchangers. A clean heat exchanger is almost twice as ef fective as one with design fouling. Design temperature of the service water system is 950 F; however, the average recorded temperature f or the month of August is 840 F. A record high river temperature of 880 F was recorded only twice; first occurring in August 1975, later in July 1977. C1SE has performed an in-depth study of RHP heat exchanger ef fectiveness, which can be found in Attachment 8 A. The technical specification values that maximize heat addition to the suppression pool are: *he events begin with the minimum suppression pool volume allowed by the technical specifications. Initial pool temperature is assumed to the maximum allowed prior to reaching alarm setpoints. The operation manually scrams the reactor at 1100 F pool temperature, which is the maximum allowed by the technical specifications. A maximum technical specification temperature of 1200 F was chosen for the initiation of manual depressurization of the FPV. Justificat ion for a manual scram of the reactor at 1100 F pool temperature is provided in Attachment 8C. No credit is taken for heat lost to the surroundings. The amount of energy required to brat up the sutmerged structures in the 14
- wetwell is neglected. In the small break case, the amount of I
l energy required to heat up the drywell is all assumed to be directed into the suppression pool. Also, the energv " held up" in the drywell is ionorEd. Furthermore, the inventory in the feedwater system is assumed to remain at a constant temperature throughout the transient; i.e., no credit is taken for heat lost through the system boundary. These cases hound all of the Chapter 15. 0 acc ident analyses in the FSAP, primarily because only one heat exchanger is used in cases la, 2a, and 3a. Also, stored energy dumped into the the condenser is minimized due to spurious MSIV closure. See Attachment 8E a comparison to Chapter 15.0 events. Using the conservative, nonmechanistic assumptions enumerated in this subsection, two of the Zimmer cases submitted slightly exceed the bulk pool temperature goal of 1900 F (see Table
- 8. 2-1) .
CGSE has evaluate] the options available which could be incorporated into the analysis in order to reduce the bulk pool temperatures. Options include: rapid depressurization of the FPV, mechanistic treatment of the feedwater system, or elimination of some of the extra conservatisms used in the analysis. In this ovaluation, CGSE has performed sensitivity studies on feedwater coastdown rates, initial suppression pool , ~3 temperature, service vater temperature, suppression pool volume, m) FHF heat exchanger tube foulina, and the number of heat exchangere in service. The results of this study are in P00R ORIGlEL
ZPS-1-MAFK II DAR AMENDMENT 14 APRIL 1981 ( - Attachment 8A. CGSE has chosen not to implement any unnecessary rapid depressurizations by opening additional SRV's, because it violates technical specification maximum cooldown rates, increascs challenges to the safety relief valves, induces a load transient on primary containment structures, and is not consistent with accepted operating procedures. It is not believed that rapid depressurization is a desirable method of reducing bulk pool temperature. The final values listed in Tabic 8.2-1 are in accordance with
" White Paper" (revision 1) developed under the direction of the Mass Energy Subcommittee, which was financed by the vark II Cwners Group. CGSE has attempted to work in accordance with the generic " White Paper" assumptions; however, in this analysis, feedwater provides makeup to the RPV, and the HPCS/RCIC is not utilized.
Pesults of the Zimmer specific cases which were performed by GCSE are also included in Attachment 8A for your information. Attachment 9A illustrates the degree of conservatism that was used in the Zimmer pool temperature calculations and further shows that the Zimmer results are within the limits specified by the NRC. 74 8.2.2 Temoerature nesponse Analysis (~ t !
This analysis was performed for the quencher SPV discharge device. Pool temperatures were calculated until a peak pool temperature was reached.
8.2.2.1 Model Descriction Mon-LOCA Events To solve the transient response et the reactor vessel and supression pool temperat ure due to the postulated events, a coupled reactor vessel and suppression pool thermodynamic model was used. The model is based on the principles of conservation of mass and energy and accounts for any possible flow to and from the reactor vessel and t he suppression pool. The model incorporates a control volume approach for the reactor pressure vessel and suppressior pool. It is capable of tracking a collapsed reactor vessel water level and having a rate of change of temperature or pressure imposed on it. The various modes of operation of the residual heat removal (EHR) system can be simulated, as well as the relief valves, HPCS, FCIC, and feedwater functions. Tre model also simulates system setpoints (automatic and manual) and operator actions and accepts as input the specific plant geometry and equiment capabilit y. (- 8.2-3 P00R ORIGINAL
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 (< Small Break Accident Model Ir the small break accident analysis, the mass and energy conservation laws are applied to a control volume which includes all of the reactor vessel contents and its walls. This control volume is subjected to the boundary conditions of decay heat input. The break and tre safety / relief valve flow rates and the associated fluid enthalples are derived f rom the state of fluid in the control volume urdergoing the transient and the specified flow areas and locations. The time-dependent break and safety / relief valve mass and energy flows are then input to another control volume containing the suppression pool. The pool temperature transient is obtained using the energy and mass balance equations on the suppression pool. A. 2. 2. 2 General Assumptions and Initial Conditions (" White Paper,"
~~
Pevision 1, Except As 1:cted By *) The following common assumotions were used throughotit the analysis of the 2PS- 1 suppressior pool temperature response:
- a. Decay heat per AMS 5-20/10.
14 l' L.); b. Design fouling of FEF heat exchangers.
- c. Wetwell air temperature equal to the suppression pool water temperature.
*d. Feedwater ir excess of instantaneous pool temperature is assumed to maintain level rather than condensate storage tank inventory via PCIC and HPCs. This assumptior raximizes heat addit ion to t he pool,
- e. Ir calculating the overall heat transfer coefficient of the vessel wall and internal structures, it is assured that the heat transfer is dominated by conduction. The heat transfer area of the reactor internals is obtained by assuming that the average retal thickness is 0.166 feet (2 inches) .
- f. The control volume of the reactor includes the reactor vescel, the recirculatior lines, the tee 1 water lines from the vessel to the nearest fredwater heaters, and the steamlines f rom the vessel to the inboard main isolation valves (MSIV) .
- g. The initial water level in the reactor vessel is calculated tased on the assumption that the voids in em the two-phase region collapse. "herefore, the ECCS
!j ON/OFF volumes are based or the total liquid volume e.2-6 P00R ORIGINM
ZPS-1-MAPK II DAR AMENDMENT 14 APRIL 1981
, of the reactor vessel, the feedwater lines, and the recirculation lines combined.
- h. The specific heat cf the reactor vessel and the internal is assumed to be 0.123 Btu /lbm/0F. The metal density is assumed to be 490 lbm/f t3
- i. A stuck-oper relief valve can be detected and the corresponding quencher within the suppression chamber identified.
- j. Additional safety / relief valves are manually opened as necessary to depressurize the reactor.
- k. Minimum tectnical specification suppression pool water level.
- 1. Maximum suppression pool initial temperature which was 950 F during power operation and 1200 F at hot standby.
- m. ASME safety / relief valve flow rate rated at 122.57.
- n. No credit is taken for heat lost to the surroundings; i.e. , all energy discharged f rom vessel is added to 14
(^3 the suppression pool. C/ 8.2.2.3 Descrintion of Mcn-LCCA Events This subsection describes the safety / relief valve discharJes for non-LOCA events (Subsection 8. 2. 2.4 describes the LOCA event) . A complete descrintion of the sequence of events for all of the cases, i.e., Events la, 1b, 2a, 2b, 3a, and 3b is given in Tables 8.2-3 through P.2-8. 8.2.2.3.1 SC9V at Power
- a. The SCEV is the initiating event and two single failures are considered separately:
- 1. loss of one PHP EX, and
- 2. MSIV isolation signal at t = 0 minutes.
- b. In accordance with the Technical Specifications, manual scram occurs at 1100 F. Manual scram is accomplished by arming and depressing the four manual scram buttons an.1 transferring the mode switch from "run" to "sl utiown", in accordance with CP.EOP. 01, Revisicn 5, in Attachment RC.
I ) c. Pool cooling initiated at t = 10 minutes. v e-P00R ORIGINAL
Z PS-1-MAPK II CAR AMENDMENT 14 APRIL 1981 k) d. For a.1. , main condenser remains available. Attachment BB. See
- e. For a.1., the operable RHR HX is placed in shutdown cooling mode. For a.2. , two RHR HX are available and no shutdawn cooling is used in the analysis.
8.2.2.3.2 Isolation / Scram i
- a. Isolation / scram is the initiating event and two single failures are considered separately:
- 1. loss of one RHP HX, and
- 2. spur ious failure of a saf ety/ relief valve in tne open position (SORV) .
- b. Pool cooling initiated at t= 10 minut es.
- c. A reactor depressurization is initiated at 1200 F.
- d. For a.2., the SORV is assumed to occur at t= 0 minutes.
,_s e. For a.1., tre operable RHR HX is placed in shutdown 14
() cooling mode. For a.2., two RHE HX are available and no shutdown cooling is used.
- 8. 2. 2. 4 Small Break Accident
- a. Two single failures considered separately:
- 1. loss of one PHP HX, and
- 2. loss of shutdown cooling mode.
- b. SCF AM on high drywell pressure and MSIV closure signal assured at t = 0 minutes.
- c. At t = 10 minutes, pool cooling is initiated.
- d. A reactor depressurization is initiated at 1200 F.
- e. For a.1., tre operable RHR HX is placed in shutdown cooling mode. For a.2., two RHR HX are available and no shutdown cooling is used.
8.2.3 Pesults/conculsions The results obtained from the ZPS-1 suppression pool temperature
,ey analyses are depicted in Figures 9.2-8 through 8.2-13. Summary
(,/ results are presented in Table 8.2-1. Conservative assumptions were used for transient events presented in this report. For P00R ORIGINAL
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 j example, maximum initial poci temperature, minimum initial pool mass, design fouling of RHR heat exchanger, continued addition of feedwater energy into the reactor vessel, and the initial reactor power corresponding up 102% of rated steam flow are conservative parameters that will affect the pool temperature. For the case of a stuck-open-relief valve f rom power (Figures 8.2-8 and 8.2-9), the peak bulk pool temperatures are 1790 F (loss of one RHP) and 1790 F (spurious isolation) . The cases of isolation / scram are given in Figures 8.2-10 and 8.2-11. The peak Fulk pool temperature for these cases are 1920 F (loss of one RHR) and 1860 F (SCRV) . To maximize pool temperature, SEA analyses were performed without actuation of ADS, as shown in Figures 8. 2-12 and 8. 2-13. The peak pool temperatures for these cases are 1960 F (loss of one RHF) and 1900 F (loss of shutdown cooling) . 8.2.4 Summary The initial Zimmer pool temperature calculations were performed in accordance with " Assumptions for Use in Analyzing Mark II PWF 14 Suppression Pool Temperature Response to Plant Transients Involving Safety /Felief Valve Discharge-Revision 1," transmitted [.1 from Mr. Bucholz (GE) , to Mr. Kniel (MPC) , on January 9, 1981. These assumptions are commonly referred to as the Mass / Energy Subcommittee " White Paper" (Revision 1) . The calculations indicate that the bulk poal temperature is acceptable for all events considered, even though some very conservative nonmechanistic assumptions were used. Item d. of Suhsection 8. 2. 2. 2 s tates t hat , "Feedwater in excess of instantaneous pool temperature is assumed to maintain level rather than condcnsate storage tank inventory via FCIC and HPCS. This assumpt ior maximizes heat addition to the pool." In this analysis, all the hot feedwater in the system, including feedwater heatere, in injected, as required, into the RPV. Furthermore, no credit is taken for heat escaping trom the feedwater through the pipe walls in these transients, which may require 5 to 6 hours to reach maximum pool temperature. The Timmer station bas two steam-driven turbine feedwater pumps. If treated mechanistically, these pumps would rapidly coast down to zero flow af ter MSIV closure. With feedwater no longer available for maintaining water level in the vessel, the ECCS pumps, taking suctior from the condensate storage tanks or the suppression pool, would automatically initiate on level 2 and inject water into tre LPV until vessel water level reaches level 8. This is considered mechanistic treatment of the 7y feedwater system. U 8.2-9
ZPS-1-MAFK II CAR AMENDMENT 14 APRIL 1981 ( ,
.he obvious dif ference in heat addition between feedwater makeup (nonmechanistic) and HPCS (mechanistic) is very significant in the cases analyzed. Eoth CGSF and GE have run Zimmer cases utilizing mechanistic feedwater treatment, and the resulting bulk pool temperatures are well below the goal of 1900 F in all cases analyzed.
In this analysis, no credit was taken for heat sinks in the drywell and the wetwell. In the SEA case, for example, all of the heat generated by tre inventory leaving the small break is assumed to be discharged, via the LOCA downcomers, into the suppression pool. It is believed that it can be demonstrated that significant heat " hold up" in the drywell would occur in this type of accident. Also, the submerged structures in the wetwell, i.a., the downcomer bracing, pool liner, etc., would absorb some of the heat energy entering the wetwell. These items could significantly reduce the bulk pool temperature. The Zimmer SFV dischargc devices are T quenchers. They are submerged 1A.5 feet at minimum pool volume (LWL), which increases the saturation temperature at the quencher device to approximately 2350 F. Test data indicates that the quencher demonstrates stability at mass fluxes corresponding to the Zimmer 14 peak bulk pool temperatures. See Attachment 8F for the ZPS-1 saturation temperature at the quencher, and the corresponding increased local pool terperature limits. (O < CGSE has performed extensive studies on the Zimmer RHP heat exchangers. It is our intentior to have regular maintenance performed on the FHP heat exchanger tute 1.D to remove corrosion causing agents and to ensure that fouling is kept to a minimum. Discussions with various tube cleaning companies are in the preliminary stages. Cur studies indicate that a relatively low fouling factor (10% of design) coupled with realistic but conservative service water temperature can reduce bulk pool temperature as much as 300 F. he have determined that regular cleaning of heat excnanger tubes will ensure a higher degree of heat exchanger efficiency. A clean heat exchanger is almost twice as effective in removing heat as one that has reached design fouling conditions. Plant modifications are being made on ZPS-1 to ensure that no single failure will result in the loss of shutdown cooling and an P"E heat exchanger loop in pool cooling. CGEE engineering and operating personnel are working tocether to ensure that emergency operatino procedures are cor.sistent with the assumptions used in this analysis. For your in formation, please find Feactor Scram Operating procedure OP. FOP.01 and Stuck Open Relief Valve of erating procedure CP.EOP. 30 in Attachment 8D. LJ 8.2-10 y
1 l ZPS-1-MAPK II DAR AMENDMENT 14 APRIL 1981 The Zimmer in-plant tests will ccnfirm the bulk to local ('/T s_ temperature difference. The results described herein demonstrate conformance with acceptable quercher pool temperature limits. 14 In summary, CGSE is taking steps to increase RHP system
" availability and to ensure that operator actions and procedures are consistent with the assumptions used in this ar.alysis. The ZPS-1 suppression pool temperature response for all cases analyzed is acceptable.
O O 8.2-11 i
O O O i TABLE 8.2-1 PCOL TEMPERATURE ANALYSIS RESULTS WHITE WHITE PAPER ATTACHMENT 8A NRC PAPER ZIMMER BULK ZIMMER CASE CASE POOL PEAK UNIQUE NUMBER NUMBER TEMPERATURE TEMPERATURES b la 179* F N/A 14
- 1. SORV at Power -
Loss of 1 RHR Hx N
- 2. SORV at Power - a lb 179* F N/A T
, Spurious Isolation Y
- Y Y 3. Isolation / Scram - e 2a 192* F 184* F* g
[ Loss of 1 RHR Hx x x 2b H~
- 4. Isolation / Scram - SORV e 186* F N/A
- 5. SBA-Loss of 1 RHR Hx d 3a 196* F 181* F* y, M
- 6. SDA-Shutdown d 3b 190* F N/A Cooling Not Available 4
Es t< ct mz co e H
- Assuming Feedwater coastdown to 0 flow occurs 30 seconds after SCRAM.
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 "N TABLE 8.2-2
. (J IMPORTANT SYSTEM CHARACTERISTICS 0
Initial Pool Mass 5.8 x 10 lbm Initial Pool Temperature 95* F ! Initial RPV Liquid Mass 457,300 lbm Initial RPV Steam Mass 17,800 lbm RPV and Internals Mass 2,479,000 lbm
- Initial Vessel Pressure 1,020 psia Initial Core Power (102% Rated) 2.35 x 10 6 Btu /sec Initial Steam Flow (102% Rated) 2,968 lbm/sec Initial CRD Flow 8.33 lbm/sec 14 CRD Flow After Scram (P RPV=0 psig) 23.6 lbm/sec
- CRD Enthalpy (From CSD) 68 Btu /lbm HPCS On Volume 8,566 ft 3 HPCS Off Volume 10,265 ft Vessel MAX P For Shutdown Cooling 150 psia ,
RHR K In Shutdown Cooling 190 Btu /sec*F RHR In Pool Cooling 190 Btu /sec*F ' RHR Flow Rate In Pool Cooling 702 lbm/sec RHR Flow Rate In Shutdown Cooling 702 lbm/sec S/RV Flow (122.5% ASME) P. psia FLOW lbm/sec 0 0 1500 368 i i Feedwater MASS (lbm) ENTHALPY ! (pipe and fluid) (Btu /lbm) 226,030 360 423,130 268
' 0s 110,340 221 1
8.2-13
~
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 () , TABLE 8.2-2 (Cont'd) 4-REACTOR POWER DECAY TIME (seconds) POWER DECAY (fraction of 1) 0 1.084 2 0.5026 6 0.6271 O.5249 10 a 20 0.2309 30 0.1372 31 0.1370 60 0.0492 .i . 100 0.0427-120 0.0400 14 121 0.0390
. 200 0.0358 600 0.0279 f 1000 0.0245 1001 0.0244 2000 0.0192 6000 0.0138 10,000 0.0120 ;
20,000 0.0101 4 6 x 10 .00739 5 1 x 10 .00624 5 2 x 10 .00512 1 x 10 .00298 6 4 x 10 .00180 l 4 1 1-l
- 1 l
8.2-14 l
ZPS-1-MARK II DAR AMENDMENT 14
' APRIL 1981
( )/ TABLE 8.2-3 POOL TEMPERATURE CONDITIONS - CASE la NRC Case b SORV at full power, 1 RHR available Manual Scram at Tpool = 110' F. Mechanistic closure of the turbine stop and bypass valves (Product Line Unique - BWR-4 or 5) . One RHR in pool cooling 10 minutes after high temperature 14 alarm. Main condenser reestablished through bypass system 20 minutes after scram using plant specified bypass capacity. Main condenser available using full bypass capacity until reactor vessel permissive for RHR shutdown cooling. RHR out-of-pool cooling when pressure permissive for RHR shutdown cooling is reached. Sixteen-minute delay for RHR transfer to shutdown cooling. (Additional SRV's opened, as C_s) required, during switchover to ensure no repressurization during switchover.) i O s_/ 8.2-15
ZPS-1-MARK II DAR AMENDMENT 14 l APRIL 1981 1 (') TABLE 8.2-4 POOL TEMPERATURE CONDITIONS - CASE lb NRC Case a SORV at full power, 2 RHR's available Manual Scram at Tpool = 110' F. 4 Nonmechanistic isolation at scram, with 3.5 seconds main isolation valve closure. Two RHR's in pool cooling 10 minutes af ter high pool temp-erature alarm. No manual depressurization is required. Depressurization rate is controlled by SORV only. RHR shutdown cooling not initiated. L O 8.2-16
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 () TABLE 8.2-5 POOL TEMPERATURE CONDITIONS - CASE 2a NRC Case e Isolation-Scram (nonmechanistic) , 1 RHR available Isolation Scram at t = o, nonmechanistic, with 3.5 seconds main isolation valve closure. One RHR in pool cooling 10 minutes after the event. 14 When Tpool = 120' F, begin manual depressurization by opening additional valves as needed. Depressurize at 100* F/hr. RHR out-of-pool cooling when pressure permissive for RHR shutdown cooling is reached. Sixteen-minute delay for RHR transfer to shutdown cooling. (Additional SRV's opened, as required, during switchover to ensure no repressurization during switchover.) i I t s l 1 l O l 8.2-17
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 TABLE 8.2-6 , O POOL TEMPERATURE CONDITIONS - CASE 2b l NRC Case e Isolation Scram- (nonmechanistic) , 2 RHR's available Isolation Scram at t = o nonmechanistic, with 3.5 seconds main isolation valve closure. 14 l SORV at t = o. Two RHR's in Pool Cooling at 10 minutes after the event. When Tpool = 120 F, begin manual depressurization by opening additional valves. Depressurize at 100' F. RHR shutdown cooling not initiated. O i i ) l O 8.2-18
l ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 ; l l TABLE 8.2-7 l l]'- POOL TEMPERATURE CONDITIONS - CASE 3a NRC Case d SBA Event Mode, 1 RHR Available Scram at t = o on high drywell pressure. Isolation at t = o (nonmechanistic) , with 3. 5 seconds main isolation valve closure. One RHR in pool cooling 10 minutes after high pool tempera-ture alarm. When Tpool = 120' F, begin manual depressurization by 14 opening additional SRV's as needed. Depressurize at 100' F/hr. RHR out-of-pool cooling when pressure permissive for RHR shutdown cooling is reached. Sixteen-minute delay for RHR transfer to shutdown cooling. (SRV's opened, as required, during switchover to ensure no repressurization during switchover.)
/~N
(_) Automatic RHR switchover to the LPCI mode occurs when the high drywell pressure is reached. Therefore, RHR pool cooling is assumed to be unavailable'for the first 10 minutes after scram. l O 8.2-19 l l
ZPS-1-MARK II DAR AMENDMENT 14 APRIL.1981 TABLE 8.2-8 {J] POOL TEMPERATURE CONDITIONS - CASE 3b NRC Case d SBA Event, 2 RHR's Available Scram at t = o on high drywell pressure. Isolation at t = o (nonmechanistic), with 3.5 seconds main
. isolation valve closure.
14 Two RHR's in pool cooling 10 minutes af ter high pool temperature alarm. When Tpool = 120' F, begin manual depressurization by opening SRV's as needed. Depressurize at 100 F/hr. RHR shutdown cooling not initiated. Automatic RHR switchover to the LPCI mode occurs when the high drywell pressure is reached. Therefore, RHR pool cooling is assumed to be unavailable for the first 10 minutes after scram. (~)'
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AMENDMENT 14 VESSEL PRESSURE P. S. I . A . APRIL 1981 i o o o O O O o O O O O O o o o o o o o o o o o o o l ro N - o m o N e o T M N o
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AMENDMENT 14 VESSEL PRE S SUR E - P.S.I. A. APRIL 1981 O O O O O O O O O O O O O O O O O O O O O O O O O
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ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 O ATTACHMENT 8A O I l O 1
' ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 ATTACHMENT 8A The purpose of Attachment 8A is to demonstrate the contribution made by various conservatisms used in the Zimmer pool temperature submittal, based on an independent study by CG&E Company. The
'two primary. items examined in this study are the feedwater system and RHR h' eat exchanger performance.
Feedwater Original cases 2a - Isolation Scram, and 3a - Small Break, were chosen for the feedwater sensitivity study because these i two cases slightly exceed the maximum desired bulk pool temp-erature'of 190* F. This sensitivity study concludes that if the feedwater system is treated mechanistically, bulk pool temperatures are well below the 190 F goal. See Table 8A-1 for results of this sensitivity study. The mechanistic 14 assumptions used for this study are consistent with Zimmer () plant operating procedures, system performance, and Mass Energy " White Paper", revision 1, transmitted fron. Mr. Buchholz (G.E. ) to Mr. Kniel (NRC) on January 9, 1981. Heat Exchanger Performance Original case 3b - Small Break, was chosen for this sensitivity study because two heat exchangers are in service in the pool
! cooling mode throughout the event. This study demonstrates that bulk pool temperature is very sensitive to the degree of tube fouling used, and the number of heat exchangers in service. It also demonstrates that bulk pool temperature for this event is well below the goal of 190 F using one heat exchanger only in the pool cooling mode, provided the tube fouling does not exceed 10% of design fouling, and the service water is reduced to the maximum recorded temperature.
- See Table 8A-2 for results of this sensitivity study.
i ( 8A-1
--e - -
e - -m -+
( TABLE 8A-1 FEEDWATER SENSITIVITY STUDY' CASES 2a and 3a FEEDWATER COASTDOWN BULK POOL COAL CASE DESCRIPTION RATE (Seconds) TEMPERATURE *F *F COMMENTS 2a Isolation Scram, 1 Hx, 100* F/Hr Cooldown *N/A (Nonmechanistic) 192 190 . White Paper Case 2a 2a Isolation Scram, 1 Hx, 100* F/Hr Cooldown 7 Mechanistic 183.25 190 See Figure 8A-1 2a Isolation Scram, 1 Ex, 100* F/Hr Cooldown 15 Mechanistic 183.5 190 'See Figure 8A-2 14 2a Isolation Scram, 1 Hx, 100* F/Hr Cooldown 30 Mechanistic 183.96 190 See Figure 8A-3 2a Isolation Scram, 1 Hx, 100" F/Hr Cooldown 6G Mechanistic 184.8 190 See Figure 8A-4 3a Small Break, 1 Hx, 100* F/Hr Cooldown *N/A (Nonmechanistic) 196 190 White Paper y Case 3a y 7 3a Small Break, 1 Hx, 100* F/Hr Cooldown 7 Mechanistic 181.7 190 See Figure 8A-5 g 3a Small Break, l'Hx, 100* F/Hr Cooldown 15 Mechanistic 182,94 190 See Figure 8A-6 ,[ l 3a Small Break, 1 Hx, 100* F/Hr Cooldown 30 Mechanistic 181.18 190 See Figure 8A-7 g
- =
3a Small Break, 1 9x, 100* F/Hr Cooldown 60 Mechanistic 178.9 190 See Figure 8A-8 5*
"s 5E oo .-3
- Feedwater is assumed to be available throughout the transient and maintains vessel water level rather "' g i than HPCS or RCIC.
O O TABLE 8A-2 O
-I
. HEAT EXCHANGER SENSITIVITY STUDY - CASE 3b (Pool Cooling Mode) .i-I Suppression Service.
!' Pool Mass ' Water Temp- Hx "K" Hx In' Bulk Pool' 'F Change From Case Ibm' erature 'F BTU /sec 'F Service Temperature *F* Base Case Variance 3b base 5.80E6 95 (design)- 190 2 190 White Paper Case 3b l
'i 207.8 l -1 5.80E6 95 (design) 198-(Design Fouling 1 Base case in sensi- ! w/5% Plugged)' tivity study 14~ 2 5.97E6 95 (design) 198 1 206 -1.8 Pool Level r 3 5.80E6 88 (hi record) 198 1 204.5 -3.3 S.W. ' Temperature 4 5.80E6 84 (avg. max. 198 1- 202.7 -5.1 S.W. h i Aug.) Temperature 4 o, e T 5 5.80E6 60 (avg. yearly) 198 1 192 -15.8 S.W.
'" Temperature X 4
U i 6 5.80E6 95 (design) 208 (Design Fouling) 1 205.7 -2.1 No Plugged - e Tubes - j 7- 5.80E6 95 (design) 248 (50% Design 1 198.2 -9.6 50% Design-Fouling) ' Fouling 8 5.80E6 95 (design) 295 (10% Design 1 190.9 -16.9 10% Design + Fouling) Fouling 4 5.80E6 95 (design) 341 (No Fouling) 1 184.6 -23.2 No %$ Fouling p@ i @. l' 10 5.80E6 95 (design) 323 (No Fouling, 1 187.0 -20.8 5% $ @ ., 5% Plugged) Plugged fH j
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I
O o TABLE 8A-2 (Cont'd) o' Suppression Service Hx "K" Hx In Bulk Pool *F Change From Pool. Mass Water Temp-Case Ibm erature *F BTU /sec *F Service Temperature *F* Base Case Variance 11 5.80E6 95 (design) 280 (10% Design 1 193.1 -14.7 10% Fouling, 5% Plugged) Fouling 12 5.80E6 95 (design) 236 (50% Design 1 200.3 -7.5 50% Fouling, 5% Plugged) Fouling 13 5.97E6 88 (hi record) 295 (10% Design 1 185.7 -22.1 Realistic Fouling) case 1** 14 5.97E6 84 (avg. max. 295 (10% Design 1 183.6 -24.2 Realistic Aug.) Fouling) case 2**- 15 5.97E6 60 (avg. yearly) 295 (10% Design 1 172.4 -35.2 Realistic !$ Fouling) case 3** y I 5.80E6 198 (Design Fouling 2 179.5 -28.3 Add 1 Hx g g 16 95 (design) w/5% Plugged) 3 4 17 5.80E6 95 (design) 208 (Design Fouling) 2 177.9 -29.9 No Plugged Tubes y
- =
18 5.80E6 95 (design) 248 (50% Design Fouling) 2 171.9 -35.9 50% Design Fouling 19 5.80E6 95 (design) 295 (10% Design 2 165.8 -42 10% Design Fouling) Fouling 20 5.80E6 95 (design) 341 (No Fouling) 2 160.5 -47.3 No Fouling
>g Ym 21 5.80E6 95 (design) 323 (No Fouling, 2 162.5 -45.3 No Fouling, p 5% Plugged) 5% Plugged ~m
~
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O ~ O TABLE 8A-2 (Cont'd) O Suppression Service Pool Mass Water Temp- Hx "K" ~Hx In Bulk Pool *F Change From Case Ibm erature 'F BTU /sec 'F Service Temperature *F* Base Case' Variance 22 5.80E6 95 (design) 280 (10% Fouling, 2 167.6 -40.2 10% ; 5% Plugged) Fouling, 5% Plugged j 23 5.80E6 95 (design) 236 (50% Fouling, 2 173.6 -34.2 50% I 5% Plugged) Fouling, 5% Plugged 14 'i j 'l
*' Desired maximum bulk poo temperature is 190* F. g a
** Cases 13, 14, and 15 use estimates of heat exchanger fouling after 1 year of intermittent operation, &
assuming units are flushed and laid up with demineralized water after each usage. 8
** 5
=
t vi X j M i CONCLUSIONS cs
- 1. Pool volume changes do not significantly change bulk pool temperatures. g
- 2. Changes in service water design temperature do not significantly affect bulk pool temperature.
- 3. Heat exchanger fouling is a very significant factor.
- 4. The number of heat exchangers available is very significant. ,
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l l ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 ATTACllMENT 8B O O
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 ( )' ATTACHMENT 8B MAIN CONDENSER The Mark II Owners Group (Reference 1) utilizes the main con-denser in the transient analysis of a stuck-open relief valve (SORV) at power with one residual heat removal system available. Both Mark I operating experience and Zimmer transient analyses have been evaulated and it was concluded that the main con-denser will be available in the unlikely event that a relief valve sticks open during reactor full-power operation. Therefore, CG&E Company concurs with the Mark II Owners Group position. The use of the main condenser as a heat sink requires that the bypass system be available, the circulating water system function, and that the main steam isolation valves (MSIV) remain open. These three requirements are addressed as follows: The only active components of the turbine bypass system are three hydraulically-operated control valves which are capable 14 of being opened by remote manual operation. [} Decay heat immediately following scram is less than 10% of the initial core power. Only one out of the three valves is required to meet the decay heat load (total bypass capability [three valves] is 25% of the reactor heated steam flow). The bypass system, as described in FSAR Subsection 10.4.4, is designed to control reactor pressure:
- a. during the reactor heatup to rated pressure, while the turbine generator is being brought up to speed and synchronized,
- b. during power operation when the reactor steam generation exceeds the transient turbine steam requirements, and
- c. during reactor cooldown.
Both a and b provide online operability checks to 2nsure that the bypass system is operable. Online ensurance of operability and redundancy of components
~s (one required of three) ensures the availability of the bypass ,
g)' system in the unlikely event that a relief valve sticks open i at reactor full-power operation. 8B-1
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 () The circulating water system function is to remove heat from the main condenser. It does so by taking water from the cooling tower, passing it through the main condenser, and returning it to the cooling tower. The only primary active components used to achieve the above function are the three 1/3-capacity circulating water pumps and two 100%-capacity cooling tower makeup pumps. All three circulating water pumps and one cooling tower makeup pump are in use when the reactor is at full power and, therefore, all pumps must be in operation when the SORV is postulated to occur. The failure of one circulating water pump or one cooling tower makup pump will not degrade the 4 circulating water system below what is needed to supply a sufficient amount of water for the removal of the decay heat that will be bypassed (through the bypass valves) to the main condenser. The ensured cperability of the pumps / system and the redundancy provided by the pumps ensures the availability of circulating water to the condenser in the unlikely event that a relief valve sticks open at reactor full-power operation. 14 In order to determine whether the main steam isolation valves remain open, the event sequence must be evaluated. The ' O- following event sequence for a SORV occurring at reactor full power has been chosen to maximize the severity of the transient. The initial conditions are the same as those utilized by the Mark II Owners Group (Reference 1). A safety / relief valve (SRV) spuriously opens and sticks open. As described in the event sequence for manually scramming the reactor with a SORV at reactor full power, (See Attachment 8D OP.EOP.01 and OP.EOP.30) the alarms sound and automatic con-trols adjust the generator load to the decrease in steam flow. Prior to the pool temperature reaching TS3 (110* F) the operator scrams the reactor by placing the reactor mode switch into " shutdown". This procedure maintains the MSIV's open when the reactor pressure falls below the low pressure MSIV closure setpoint. Following the scram, the feedwater system continues providing makeup to the RPV thereby maintaining reactor vessel water level, the turbine control valves (TCV) ~close as the RPV pressure drops, and the stuck open relief valve continues to depressurize the RPV. After scramming the reactor and stabilizing the water level, the operator places the RER system into suppression pool cooling. At this time, no MSIV isolation signals have been generated. O As the SORV continued to depressurize the vessel to the " low steamline pressure" main steam isolation signal, the MSIV's 8B-2
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 ' Q (j do not close. This isolation signal is bypassed when the I reactor modo switch is in shutdown. Twenty minutes after the initiation of the transient, the operator opens the bypass valves _thcreby utilizing the main condenser as a heat sink. The bypass valves do not open auto-matically due to the decreasing reactor pressure caused by the SORV. Condenser vacuum is maintained by the steam jet air ejectors (SJAE). The main condenser and SORV continue cooling and depressurizing the reactor. When the shutdown cooling pressure permissive is reached, the main condenser is no longer used as a heat sink and the RllR system is transferred to shutdown cooling. From the above event description one can see that no mechan-istic MSIV closure signals are generated and, therefore, the MSIV's remain open. The redundancy provided in the bypass and circulating water systems and the use of the systems during plant operations ensure their availability. The availability of the bypass and circulating water systems along with the main steamline isolation valves remaining open allow the operator to utilize the main condenser as a heat sink. Therefore, CG&E Company has justified that the main condenser is available for the (]) analysis of the postulated pool temperature transient in-volving a stuck open relief valve at reactor full-power operation coincident with the unavailability of one of the two redundant residual heat removal systems. REFERENCES
- 1. Letter plus enclosure dated January 9, 1981, from Mr.
Buchholz (GE) to Mr. Kniel (NRC) on Mark II Containment Program, " Assumptions for Use in Analyzing Mark II BWR Suppression Pool Temperature Response to Plant Transients Involving Safety / Relief Valve Discharge-revision 1." i i f V 1 1 l 8B-3 i
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 O l i I ATTACIIEMENT 8C O l l O 1 I l j
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 ATTACHMENT 8C ( MANUAL SCRAM As identified by the Mark II Owners Group (Reference 1), it is assumed that a manual scram of the reactor occurs when the suppression pool temperature reaches TS3 for the transient analysis of a stuck open relief valve at power. TS3 is de-fined as the maximum allowable suppression pool temperature while maintaining the reactor critical. CG&E Company concurs with this assumption for Zimmer. TS3 for Zimmer is 110* F. In the unlikely event that a relief valve sticks open at Zimmer, the operator would be alerted to this by both primary and secondary alarms and plant parameter displays. The following primary alarms / displays would indicate an open valve immediately after the valve opened:
- a. the SRV LEAK DETECTOR alarm,
- b. the ADS /SRV OPEN position alarm, and i ( c. the ADS /SRV OPEN position indication.
The following secondary alarms / displays would identify changes in the plant due to steam discharging through the open valve:
- a. the continuous display Pool-Temperature recorder /
J indicator would show an increasing pool temperature, 1
- b. the pool temperature monitoring system would alarm at TS1 (95* F) ,
t ! c. power meters would indicate a generator load de-crease with no change in reactor power, and i
- d. steam flow, feed flow mismatch could occur.
The above primary and secondary alarms / displays are those that are representative of a stuck-open relief valve (SORV) . These alarms / displays provide the operator both immediate and unambiguous indications of a stuck-open relief valve and high pool temperature. In order to clarify what information the operator sees and actions he must perform, the following event sequence is pro-vided. This event sequence has been chosen to maximize the l () severity of the transient. Initial conditions are the same as those utilized by the Mark II Owners Group (Reference 1). l l 8C-1
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 A safety / relief valve (SRV) spuriously opens and sticks open. (~f Immediately the ADS /SRV OPEN alarm sounds, the ADS /SRV OPEN indication light for the specific valve turns on, and the SRV LEAK DETECTOR alarm sounds. (1) Since it is assumed that the pool temperature is just below TS1, at the outset of the transient, the pool temperature monitoring system will alarm immediately after the valve opens and the displayed temperature will begin to increase. Since steam is being diverted from the Turbine-Generator, the generator load will decrease while a constant reactor power is maintained. The operator at this time (<1 minute after initiation of event) has a clear set of " symptoms" that indi-cato a stuck-open relief valve has occurred. Operating pro-cedures require the following operator actions:
- a. Attempt to close the relief valve remote manually
, by placing the control switch to open, than back to auto two times, in accordance with OP.EOP.30. See Attachment 8D.
- b. SCRAM the reactor and follow OP.EOP.01. See Attachment 8D.
14
- c. Initiate pool cooling.
- d. Open bypass valves fully.
(J') After scram, the pool temperature monitor would continue to show an increasing temperature and an alarm would sound when TS3 was reached. From the above description, it can be seen that the operator has sufficient information via alarms and displays to immediately identify a SORV. Due to clear identification of a SORV and the minimal operator action required, the reactor will be scrammed prior to the bulk pool temperature reaching 110* F. CG&E Company has provided diverse primary and secondary alarms / displays and an extensive pool temperature monitoring system that will provide an immediate identification of a SORV. CG&E Company plant operating procedures provide for and require specific actions be taken in the event of a SORV. CG&E Company (1) The ADS /SRV OPEN alarm and ADS /SRV OPEN indication light are triggered by position switches which provide a positive valve position indication. The SRV LEAK DETECTOR alarm is triggered by discharge pipe thermocouples. O-8C-2
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 (f has, therefore, justified that a manual scram at TS3 (110' F)
, is a conservative assumption for the evaluation of pool temp-erature transients involving a stuck-open relief valve at power.
14 REFERENCES
- 1. Letter plus enclosure dated January 9, 1981, from Mr.
Buchholz (GE) to Mr. Kniel (NRC) on Mark II Containment Program, " Assumptions for Use in Analyzing Mark II BWR Suppression Pool Temperature Response to Plant Transients Involving Safety / Relief Valve Discharge-revision 1." i l i ) i 4 a 1 8 8C-3 e , n---,- , --- - , , - -- - . , , - -
--wa - a < -- a- . - m 6 ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 O
ATTACIIMENT 8D O I O
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 ATTACHMENT 8D O THE CINCINtJATI GAS C ELECTRIC COMPAtav l wM. H. ZI"*ER IJUCLEAR POWER STATICN l Ut4 ! T 1 E:4ERGENCY OPERATING PRCCEDURE STuCA OPEN RELIEF vALv6 PROCEDURE NUMBER: OP.60P.30 REVISION: 00 PREPARED dY:a .' A% - - - M QATE: I
/
REVIE.vE0 dY: - - -, aM dL----- [/ Af/ _ 3arg: / f.
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OPERATICNS SUPE 4vl50R: - - -y
-------- DarE: -----------
S T A T I0td SUPERIt4TE 4CEN T: _[ .y h_____JATE: __@ j5~ []__ l l o CP.eeP.>c-1 P00R ORIGINAL l REv.00 I HD-1 l l
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981
}, T UC K OPEN REL!EF VALVEJ MRV 1 10 COND!TIO'4S O SAFETY RELIEF VALVE INDICATES OPEN. g 2.0 AU T Or4 A T I C AC T ! ONS NONE o
o P00R ORIGINAL CP.EOP.30-2 964.00 8D-2
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 3.0 IMMEDIATE ACTIONS l
,- 3.L CONTROL cpERATOR 3.1 1 CLOSE THE SAFETY RELIEF VALVE. ATTEMPT THIS TdICE.
312 IF... THE 50RV CAN NOT BE CLOSED THEN... INSERT A MANUAL SCRAM AND ENTER GP.E00.01. REACTOR SCRAM CONCURRENTLY 41TH TMIS PROCEDURE. 32 SENIOR CONTROL OPERATOR 321 IF... THE SAFETY RELIEF VALVE IS STUC< OPEN, THEN...ANNOUNCE TWICE ON THE PA, "STUC< OPEN RELIEF VALVE, INSERTING REACTOR SCRAw." 3.2.2 INITIATE SUPPRESSION PCOL C0 CLING w!TH RH LCOPS A AND 6. 3.3 LICENSE 3 PLANT OPERATOR 3.3.1 REPORT TO Tr5 CONTROL 400M. 14 cococococococcccoccocosecococcocoococococcucco C 3
- CAUTION 4
)
(^J
~ o a THE 100 DEGREE F/HR C00LOOWN RATE SUST BE *
- EXCEGOE0 FOR A SORV. a a e CCCCCCCCCCCCCCCCCCCCUCCC0cCC30CC*cccccccccccCo 332 IF...THE SAFETY RELIEF VALVE IS STUC< OPEN, THEN.. 0 PEN ALL TURSINE 3YPASS VALVES FULLY.
3.5 SHIFT TECHNICAL ADVISOR _1STA1 REPORT TO THE CONTA0L ACCM. 3.o RAOLCHEw TECH 361 RETURN TO RA0aASTE CONTROL ROCM. 3.6.2 SECURE RA0nASTE PROCESSES AS NECESSaav. 3.o.3 NOTIFY THE MAIN CONTROL ROOM OF ACTIONS TA45N AND PROCEED AS DIRECTED 3Y TrE SMIFT SUPERvlS00, l'h U
'=' P00R DRIGINAL l 8D-3 l
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 3.7 1HIFI_SUPERv!SOR rh ts_) 3.7.1 REPORT TO THE CONTROL ROCM. 3.7.2 ASSUME EMERGENCY OUTV SUPERVISOR. 373 ASSESS THE SITUATION. 3.3 CAS OPERATOR 3.8 1 RESTRICT, CONTROL ROOM ACCESS. 3.d.2 PREPARE FOR NOTIFICATION. 4.0 10PPLEMENTARY ACTION 41 SENIOR CCNTROL OPERATOR 4.1 1 ENTER OP.EPG.02. " CONTAINMENT CCNTROL" 14 CONCURRENTLY wlTH THIS PROCEDURE. 4 1.2 WHEN...THE SHUT 00WN COOLING INTERLOCK 5 CLEAR. THEN... PLACE SHUTOOWN COOLING IN SERV!CE. OMITTING THE nARMUP ANJ FLUSHING RE2VIREMENT5 92 Sm!AT Si>P E A v 15CR /~ 4 2.1 OECL AR E AN UNUSUAL EVENT. V) 4.2.2 NOTIFY THE OPERATING ENGINEE4. 4.2.3 ENTER OP.IPOP.03. " PLANT SHUT 00aN" CONCUR 4E?4TLY dITH THIS PROCEDURE. CE%Do I o PDDR DRIGINA Cr.EOP.30-* CEv.00 8D-4
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 O THE CthCINNATI GAS 6 ELECTRIC COMPANY a .w . H. Z I .*
- E 1 NUCLcAP 70aEl STATION UNIT 1 E*ERGENCY CPERATING PROCEDUAE REACTGk SCEA*
O PROCEDURE NuwdER: G P . 4 G P .11 REv!sICN: 05 PREPARED av: _ __________3 ATE: _L_[f_f3L.
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w _o OPERATIONS SUPEdvl50R: .L f ,, h_t. g _ _,, ______ O A T E : f_' '_/ - ?/____
$ TAT 10N SUPERINTE 40ENT: ____ __
__ _ _ _ _ _ _ __, J A T E : ___ __ f_[_ O
=P.EcP. ~ 1 aEv.05 P00R ORIGINAL 8D-5
Zl'S-1-MARK II DAR AMENDMENT 14 Al'RIL 1981 d!!GI9R _.1Elad O 10 (010tT10N 11 REACTOR S C R A l* CCCURS PE4 A T T ACH'4E NY 1. 20 AyT04ATIC AST10j5 21 CONTROL RCOS ARE INSERTED. Y' 22 REACTOR PRESSURE CONTRCLLEO BY DEnc AND/Ca 5 '4 / . 23 RdACTOR LEVEL CONTRJLLEO AT 18.75". 29 R4 PUMPS MAf SHIFT FROM FAST TO SLO. 576E] JH TAIP. 2.5 EMERGENCY CORE COOLING SYSTE45. A r.0 THE DId5EL GENERATORS MAY INITIATE. o. o . . , , . _ P00R ORIGINAL REv.05 8D-6
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 o 3.0 15df21111 aCI19NJ V 31 (ONTROL OPERATCR 311 A R t1 AND DEPRESS THE 4 *ANUAL SCoAu BUTTONS. 312 PLACE REACTOR MODE Sw!TCH T0 SHUT 00'aN 313 INSERT IRM' S AND SRM' S 3.1.4 SHIFT AP9M/IRM RECORDERS TO THE IRM POSITIONS TO MONITOR ANO VERIFY REACTOR POaER DiCREASE. 315 VERIFY AUTOMATIC ACTIONS. 3.1 6 MONITOR RPV PRESSURE. 317 RE-ESTABLISH RPV WATER LEVEL ABCVE THE SCRAM SETPOINT. 32 Sgy10R CCNTROL CPERATOR 3 2.1 ANNGUNCE OVER PA SYSTE*. "REACTCR sCRA*. REACTOR SCRAM". 3.2 2 *0NITOR THE P R I .* A R Y CONTAINMENT PRESSuaES. TEAPERATURES. AND SUPPRESSION PCi L L5VsL. 14 (~S 3.2.3 ANNCUNCE ANY CONTAINMENT ISCLATIONS TO CONTRGL \_) 400M PERSONNEL. 33 LICENSE 0 DLANT OPERATOR 3 3.1 VERIFY THE TURBINE GENERATOR HAS TRIdPEC. 3.3 2 VERIFY TRANSFER OF THE AutILIARY ELEC TRIC AL LOADS. 3 3.3 VERIFY AUTO START CF DIESEL GENERATORS. IF APPLICABLE. 3.4 PLANT CPERATCRS REPORT TO THE CONTROL ROOM. 3.5 SHIFT TECHNIC AL ADVISOR REPORT TO THE CONTROL ROOM. 36 RfDLC-E9 TECH 3.o.1 RETURN TO THE RAcaASTE CONTRCL HOC *. 3.o.2 SECLRE RAG 4ASTE PROCESSES AS NECESSARY.
'N (O
CP.EOP.01-3 8D-7 P00R ORIGINAL
ZPS-1-MARK 11 DAR AMENDMENT 14 APRIL 1981 363 NOTIFY THE MAIN CONTROL ROOM OF ACTIONS TAAEN. /~ ANO PROCEED AS DIRECTE] Sv THE SHIFT k-)/ SUPERVISOR. 3.T SHIFT SgPF A'f1}QR 3.7 1 REPORT TO AND REMAIN IN THE CONTROL A00.*. 3.7.2 ASSUME E.MERGENCY OUTY SUPERVIS3R. 3.7.3 ASSESS THE SITUATION. 3.3 (AS OPERATOR 3.3 1 RESTRICT CONTROL R00.9 ACCESS 382 PREDARE FOR NOTIFICATION 4.0 SUPPLEWENTARY ACTIONS 41 (ONTROL OPERATCR 4.1 1 ESTAdLISH REACTOR 4ATER LEVEL 1ETaEEN 12 2" ANO 58" w!TH THE RFP'S. o o o o o o oo o o o o o o oo o c a c o c o o s o o c a o oo.1o o o o o o o o o o o o IF...UNASLE TO RES TCRE VESSEL LE/EL. o 14 e a () '\ J
- THEN... PROCEED TO OP.EPG.01.
oooooooooooooooooooooooooooooooooooooooooooo o 412 PLACE FEE 0 WATER CONTROL IN SINGLi ELE *ENT CONTROL. 413 RE3E T THE SCRAM SET]CaN FEATURE aHEN 'JIRECTED BY 55.
*.1 4 RETURN RT TO SERV!CE.
4.1.5 DETERMINE THE CAUSE OF TmE S C R A .* . 4.1.o aHEN THE SCRAM SIGNAL HAS CLEAREO, svoASS T dE SCRAM OISCHARGE VULUME ANJ RESET T-E SC2AM VALVES.
*.1 7 '40T E ON CH AR T TRACES T*E T!wE . EVENT. -
413 COMPLETE TdE SCRAM REPOR T (ATTACHMENT 2). 4.2 }{NICA CONTRCL CPERATOR 421 IF... SUPPRESSION PCOL TEdPERATuai ExcEE;S 75 I DEG F. OR ORyaELL TE AGER A TJRE EXCEECS L35 Dii F. JR DRYaELL 20ES$uRE ExCEEJS 1.e4 PSIG. 24 ("N \- JP.EOP.01-4 ,
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 J SUPPRESSION POOL WATER LEVEL EXCEEDS 22 FT. 6 I (> 'uj IN., TPEN... ENTER OP.EPG.02, CONTAINMENT CONTROL AND EXECUTE IT CONCURRENTLY n!TH THIS PROCEDURE. 4.2.2 CLOSE ANY SORv A. IF AN SORV CANNOT BE CLOSE0 AFTER 2 ATTEMPTS, VERIFY THE REACT 0R IS SCRAMMEO. 4.2 3 IF SRV'S ARE CYCLING, MANUALLY OPEN ONE SRv AND REQUCE RPV PRES 50RE TO BEL 0m 900 PSIS. 4.2 4 RE SE T GROUP ISCLAT!0NS AS DIRECTE0 SY TME SS, IN ACCORDANCE nITH OP.PC.02, PRIMARY CONTAINMENT ISOLATION RESET. 43 LICEN110 PLANT OPERATOR 1911 HP AUTO INITIATION SIGNALS DILL CAUSE A LCAv5HED JN Sn!TCnGROUP 1C. 10, AND 16.
*.3 1 IF A LOADSHED OCCURREJ, INITIATE QP.60P.21, 34 LOSS OF aux!LIARY ELECTRICAL PonER.
f'% \ ' 4.3.2 VERIFY TNAT THE TUR4INE SUPPURT SfSTEMS S PARAMETERS ARE WITHIN LIMITS. 4.3.3 MAINTA[N AND/OR RE-ESTAdLISH dAIN CONDENSER VACUUM IN ACCORDANCE AITH OP.CA.01 AS JIdECTED Sv THE SelFT SUPERVISOR.
*.4 SHIFT SUPERVISOR 4.*.1 CHGCK THAT THE IMMEDIATE ACT!ONS OF THE REACTOR SCRAM PROCEDURE HAVE SEEN CO*PLETEO.
DIRECT SHIFT CREd MEMSE25 GN APPLICABLE SUPPLEMENTARY ACTIONS. 4.6.2 DETE4MINE THE CAUSE OF THE SCRAM AND CLEAR TnE CONDITICN, IF POSSI3LE. C00RJINATE TwE ISME0! ATE AND SUPPLE *ENTARY ACTIONS lf T-E APPROPRIATE PROCEDURES. 443 OIRECT THE CONTROL OPERATOR TO RESET SCRAM. 4 9.* IF CCNDITIONS Ex!ST dMICn ARE OESCRIdEJ IN TAdLE F.*-1 0F THE EMbRGE%CY PLAN, DECLARE: A. UNUSUAL EVENT /^N (l JP.EOP.01-5 ,
""ll ., P00R ORIGINAL.
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 I B. ALERT C. SITE EMERGENCY
- 0. GENERAL EMERGENCY 445 HAVE CAS OPERATOR INITIATE NOTIFICATIONS. AS APPLICABLE.
4.4 6 NAVE OP.EPP.08, AIRBORNE AELEASE CALCULATIONS PERFORMED IF AN INADVERTENT RELEASE OF AIRBORNE RADICACT!vE MATERIAL HAS OCCURAE0 AS VERIFIED BY ANY OF THE FOLL0w!NG C]NTR0L A00M PANEL IN0! CATIONS: A. AUDIBLE ALAR 45 AND SUSTAINEO INCREAS40 REA00uTS ON: MAIN PLANT VENT STACK PA01ATION MONITOR; ANO/OR OFFGAS POST TREATMENT RA0!ATION MONITOR; AND/OR PLANT V E ?i T PLENJM RADIATICN SONITOA; AND/OR 14 REFUELING FLOOR VENT PLENUM RA0!ATION h) w/ MONITOR B. AUTOMATIC INITIATION OF THE VG SrSTEM. 4.4.7 HAVE THE COMPUTEA ANNUNCIATOR LOG REVIEdEC. 4.4.9 NOTIFY THE P0aER SUP ER /130a. 4.4 10 NOTIFY CPERATING ENGINEEA 0F REACT 14 SCRAM A'80 CAUSE. tF CAUSE IS CORRECTED. OSTAIN PERMISSION TO RESTART UNIT CR SMUTCOaN IN ACCORDANCE WITH OP.!P00 03. SHUT 00aN. Cd OP.EPG.03, C00LD0dN Gul0ELINE. 4.4 11 RE]UEST THAT RAD / CHEM TECH 08TAIN A SAMPLE JF REACTOR *ATER AND ANALvZE F04 00SE E0ulvALENT 100!NE ACTIVITY. AT THE POST LOCA SA9PLE MONITOR. 5.0 O_15C MS Sirg, THE CONDITIONS AND SETPCINTS aHICH COULO CAUSE A SCRAM AAE LISTED IN ATTACHwENT 1. IF A i4 Y OF THESE CON 0!TIONS AaE DISCOVERED TO Ex!ST. OR IS IT IS Ev!]ENT THAT AN AUTOMATIC SCRAM 15 UNAVOIDABLE. T-E 4EACTOR SHoutJ SE dANUALLY SC A A.dM E O. A V JP.ECP.01-3 AEv.05 8D-10
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981
-s THE OPERATOR SHOULD BE AWARE THAT ANY SCRAM MAY CAUSE A (d TEMPORARY low WATER LEVEL CONDITION IN THE REACTOR.
LEVEL MAY INITIATE EMERGENCY CORE COOLING SYSTEMS. THE THIS OPERATOR SHALL CONTINUQUSLY MONITOR RPv WATER LEVEL USING ALL AVAILAdLE INSTRUMENTATION AS FOLLOwS: T,gE RANGE INSTRUWENT NO. 3 A:JEL j NARROW RANGE ILR-1C34-R608 1H13-P503 (0" TO +o0") NARROW RANGE ILI-1C34-roc 6A,8.C lH13-P603 (0" TO +60") WIDE RANGE ILR-1921-R623A 1H13-P601 (+60" TO -150") WIDE RANGE I L R-1821- I o 2 3 8 1H13-Po01 14 (+60" TO -150") WIDE RANGE ILI-1821-R604 1H13-P601 (+o0" TO -150") SHUTDOWN RANGE ILI-1321-R605 1H13-Po02 (0" TO 400") FUEL ZONE RANGE ILI-la21-A610 1H13-Po01 l (+50" TG -153") FUEL ZONE RANGE ILR-1621-R615 1613-P6C1 (+50" TO -150") cEND* i I (Q> , GP.EOP.01-7 8D-ll
- ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 (V 1 AUTOMATIC REACTOR SCRAM SETPOINTS REACTOR TRIP FUP.C T I ON TRIP SETPOINT MODE S. ITCH
- 1. IRM HI Flux 120/125 OF SU FULL SCALE
- 2. APRM:
A. HI Flux 15% SU B. FL0w BIAS FLUX (0.6eW + 51)t OR RUN 113 5% max RUN C. HI FLJX (FIXED) lidt RUN
- 3. Rx VESSEL PRESSURE HI 1043 PSIG Sv/RUN
- 4. RX VESSEL LEVEL L O 'a 12 5 INCHES SJ/RUN
- 5. MSIV CLOSURE et CLOSURE RUN EITHER MSIV IN 3 OF 4 STM LINES
- 6. MSL PADIATION HI 3x FULL P0aER SJ/ dun 3ACKGROUND
- 7. PRI CONT PRESS HI 1.o9 PSIG Sv/RUN 14
- 8. SCRAM OISCH VOL 76 GALS SU/Rus4 LVL HI
- 9. TSV CLOSURE 3 0F 4 TSV CLOSE0 104 4UN (IF >30% P0aER)
- 10. TCV FAST CLOSURE 850 PSIG OIL PRESSURE AuN (IF > 30% P0aER)
- 11. *00E Sw IN SHUTD0wN N/A 1/A
- 12. MANUAL SCRAM N/A SU/RUN
[AUUA LSC,RA$
- 1. 110 DEG F SUPPRESSION POOL TEMPERATURE
- 2. SORV. (AFTER 2 ATTEMPTS TO RECLOSE).
- 3. TRIP OF SOTH RR PUwPS TO ZERO SPEEJ.
- 4. LOSS OF IA ANO ROD ORIFT.
- 5. OG PRETREATMENT 'tCN I TC A >5 CI/SEC.
q J ATTACHMENT 1
"=
P00R ORIGINAL 8D-12
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 (O V GRAM REPORT
- 1. SCRAM P4 UMBER: _ _ _ _ _ _ _
- 2. SCRAM TIME: DATE:
- 3. FRESCRAM CONDITIONS:
4 REACTOR: CRITICAL SUBCRITICAL S. PLANT EVOLUT10N: 5 TART UP SHUTTING 00a*4 5TEADY 00 C. P0wER LEVEL: ____________% _________ _________ w'a T
- 0. MODE SdITCH: RUN , _ _ _ _ _ _ _ _ _ _ _ SU/STSY ,_ _ __ _ _ _ _ _ _
E. TURSINE-GEN: SYNC _ LOAO MwE RPM F. RECIRC FLCd _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ NO OF PUMPS ,,__ _ _____ CONTROLLER POSITION: LOCAL MAN. G. TURB IST STAGE PRESS.: __,,__,___________,,,_,,,_SIG p 14 H. REACTOR 57629 FLCa: ____ ____ ____________________cS/Ha
!. FEE 0 WATER ALoa: _ _ _ _ _ ___ ___ ____ _ _ _ _ ____ _t.3/MP
\") J. CORE DIFF PRESS AND FLOW:________ _ _ __ PSI ______________. K. RPV LEVEL: _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ________________IN L. RPV PRESS: , _ _ _ _ _ _ _ _ _ _ _ _ _ , _ _ _ , _ _ _ _ _ _ , _ _ _ _ , , 351,
*. RELIEF VALVE OPERATION
SUMMARY
OPERATED _S_ R _V YE54NO ::C P E N I NG S MANUAL AUTO ClwvENTS A 3 C D E F G H L p R S A T T ACHa4ENT 2 )
'=r P00R ORIGINAL 8D-13
ZPS-1-MARK II DAR AMTNDMENT 14 APRIL 1981 ("3>
~
- 5. MAXIMUM RPV PRESSURE __________ PSIG us
- o. ITEMS ATTACHED:
A. CONTRCL RCOM LOG (CDPY) B. COMPUTER LOG C. SEQUENCE OF EVENTS (COMPUTER)
- 3. ANNUNCIATOR LOG E. RECORDER CHARTS - APPLICABLE SECTIONS OF:
M LIN PLANT VENT STACK RA0!ATION MONITOR OFFGAS POST TREATMENT RADIATION *0NITOR PiANT VENT PLENUM RADIATION MONITOR REACTOR LEVEL FEEDWATER FLOW STEAM FLOW REACTOR PRESSURE T. CAgsi_3F AND DESCRIPTION OF SCRAM 14 DESCRIBE THE SE;UENCE OF EVENTS LEADING UP TO AND FULL 0 DING THE SCRAM. DESCRIBE OR SKETCM ANY A3 NORMAL OBSERVE 0 TR ANS IENTS. OETERMINE T-E PROBABLE CAUSE ANJ ATTACH T-E APPROPRIATE SCP AN0 NSS POST TRIP LOGS AND REC 0dDER CHARTS {(m)x TO THE SCRAM REPORT. _____._____________f.___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ TIME DATE SHIFT 5UAEavl5JR > /~
\_)) ATTACH *ENT 2 JP.EOP.01-10
! REV.05 l 8D_14
1 1
\
I ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 O i i 1 l i 1 l l 1 i ATTACllMENT 8E O - l 1 r i i j i l O
TAntE E-1 FSAR CllAPTER 15 EVENTS VERSUS LONG TERH POOL TEftPERATURE FVEN15 SIMIl AR LOHG TERM POOL TEMPERAluRE EVENTS CASES la,1b CASES 2a,2b CASES 3a 3b SORV SHAL L fl0 POOL 14 AT 150tATION/ 111 TEAK 1LHPERAlliRE FSAR CilAPTER 15 EVENIS POWER SCRAM ACCIDENT INCREASE 15.1.1 Loss of feedwater lleating X 2 feedwater Controller failure - Maximum Demand X s 3 Pressure Regulator failure - Open X 3' 4 Inadvertent Safety Relief Valve Opening X ,, 5 PWR Steam Piping Break N/A ' 6 Inadvertent IlllR Shutdown Cooling Operation X 15.2.1 Pressure Regulator failure - Closed X Generator Load Reject 2 3 4 lurbine Trip MSIV Closures X X f X 5 Loss of Condenser Vacuim X 6 Loss of AC Power X 7 1oss of feedwater flow X 8 feedwater Line llreak X 9 f ailitre of R!ill Shutdown Cooling X 15.3.1 Recirculation Pimp Trip X 3E 2 Recirculation Flow Control Failure - X [g llecreasing Flow 3 itecirculation Purnp Seizure X eg 4 liecirculation Pump Shaf t Dreak X Z
A' 0 ,e$ gmlm$ wk - 4
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R A T i t u H s u n o n S G f e C i ae m c l p a F N l r e i a O e t 0 r uu t l M d n 1 er cl s b e i o we ri y m m v s C ow i a S e pe l t P o P cf e s s ut a eu e l t s V rO d w Rl o A ry u i ot o rt t aS f l k ss LAner t nnn t e i att ol t neoe Sl i acnu idn odid o l F i eO tI o Citi S r ek Dd rra C ccc Ct R ae ik oorf ecec P n eb gca rreow rrp o mAjA u E l i o yrince i
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uf wol do Ce f erit l l at oVdor P m anoPne aaMrnl nRD I! u Sit an 5 ww aoFdu l L aml i 1 aadti nDf d t o t reoL N rroSt ga oo nV nt ct o E ddR an d R e ennsC r V E hh l l il ttl aus acul em tl ra t eey- e omcacaro rmCSf t 5 i i ec eu oa 1 WWrrreil t r vi vrm-n - w t oirmpct d m dt aasd R ddnncces en ae aseese ooob enhi po nh nntt ne E T RRCARI CMSC I C I I SSI f P A l l o C 1 2345 6/89 12 123456 _. R A 4 5. 6. S 5 5 5 F 1 1 1
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ZPS-1-MARK II DAR AMENDMENT 14 j APRIL 1981 O l ATTACIIMENT 8F O O
v l ZPS-1-MARK II DAR AMENDMENT 14 ! APRIL 1981 l O TSAT = 235 F @ QUENCHER 235 230 / / o
/ 20 Subcooling 225 /
/ /
30 F / 220 , g
' . Subcooling 215 / ' / /
210 ! / 205 ' l' l
$ 200 l
I 8 l o ! O i z m 2 i 1 l l I l ee I I ! s8 I t" i I I I a l l l l 1 1 I I I I I i l 1 42 59 77 94 (160.8) (225.9) (294.8) (359.9) MASS FLUX lbm/ft sec.
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P00R ORIGINAL WM. H.IIMMER NUCLEAR POWER STATION. UNIT 1 MARK 18 DESIGN ASSESSMENT REPORT FIGURE 8F-1 i SUPPRESSION POOL TDiPERATURE vs. MASS FLUX
2PS-1-MARK II DAR AMENDMENT 14 APRIL 1981 INDEX TO NRC 00ESTIONS (Cont'd) AMENDMENT QUESTIONS KEYWORD INDEX TO QUESTIONS NUMBER PAGE l 14 020.62 Suppression Pool Temperature Monitoring 13 B-36 020.63 Limiting the Suppression Pool Temperature 14 B-38a 14 020.64 Lateral Loads on Downcomers I During Vent Clearing 13 B-39 020.65 Vent Exit Flanges 13 B-40 020.66 Static Equivalent Loads for Down-comers with Diameter < 24 Inches 13 B-40 020.67 Multiple'Downcomer Loading 13 B-41 020.68 Maximum Pool Swell Elevation 13 B-42 020.69 Upward AP 13 B-42 020.70 Drag Loads on Submerged Structures 13 B-44 020.71 Calculated Drywell Pressure 13 B-44 020.72 Impact Pressures 13 B-52 020.73 Pool Swell Velocity 13 B-53 020.74 Chugging Loads 13 B-54 020.75 Main Vent Condensation Loads on Submerged Structures 13 B-54 (]) SERIES 040 OUESTIONS - 041.54 Suppression Pool Hydrodynamic Loads 13 B-59 041.55 Loads Following Pool Swell or Seismic Slosh 13 B-59 041.56 Multiple Downcomer Lateral Loading Combinations 13 B-60 041.57 Support Column Loads and Downcomer Vent Loads 13 B-61 041.58 SRV Load Calculation 13 B-61 041.59 Loads Due to Subsequent SRV Actuation 13 B-64 SERIES 130 OUESTIONS 130.1 SRV Loads 13 B-65 130.2 Load Combination Time History 13 B-65 130.3 Load Combination Probabilities 13 B-66 130.4 Soil Modeling 13 B-66 130.5 Liner and Anchoring 13 B-66 130.6 Asymmetric SRV Loads 13 B-67 130.7 Combining SRV and Pool Loads 13 B-67 130.14 Breakdown of Contributing Member Forces 13 B-67 130.15 LOCA Cyclic Condensation Load 13 B-68 p(,/ 130.16 Time Phasing of Loads 13 B-69 130.17 Force and Moment Diagrams 13 B-69 130.18 Maximum Force Value Plots 13 B-70 B-iii
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981 r'g OUESTION'020.63 (L
"For limiting the suppression pool temperature, provide e the following additional information:
"(1) Present the temperature transient of the suppression pool starting from the specified temperature limits for the following transients:
(a) Stuck open relief valve (b) Primary system isolation (c) Initiation of auto depressurization system l
"(2) Describe the instrumentation which will alert the operator to take action to prevent the pool tempera-ture limit to be exceeded.
"(3) Describe the operator actions and operational se-quence for those transients stated in Item 1 above.
Provide and justify the assumptions of time for , initiating each action and the corresponding pool temperature." () RESPONSE The information requested is included in Chapter 8.0 of the Design Assessment Report. 4 1 O B-38a
1
-ZPS-1-MARK-II DAR AMENDMENT 14 APRIL 1981, /~ - b] G.5 FLOW BLOCKAGE EFFECTS OF DOWNCOMER BRACING G.S.1 Correction Factor For Blockage-The downcomer bracing is a flow restriction which increases the. fluid velocity and acceleration during pool swell.
As a result, the standard drag, acceleration drag and lift loads on structures in the pool swell zone are higher than those which would exist if no downcomer bracing was present. Although the pool swell loads are not design-controlling criteria, the load calculations were adjusted for blockage effects by introducing a multiplicative factor to the fluid velocity and acceleration. A method of correction has been developed based on References 1 and 2. A.multiplicative factor has been determined based upon Maskell's paper (Reference 2): C D - C Dg f_ (1) D f: . where: ' C D* is the modified drag coefficient of structures in the pool swell zone
- C Dg is the steady flow, free stream drag coefficient 14 n is the blockage factor S is the total blocked area C is the unrestricted flow area.
The value of the blockage factor, n, depends upon the structure's geometry. The blockage ratio varies from 0.96 to 2.77 for . structures with aspect ratios, AR, from = to 1.0, respectively. . Maskell (1963) recommends a blockage factor of 2.5 for l bluff bodies and the value is considered conservative for
- the suppression pool bracing system.
The unrestricted flow area, C, is the p;o1 surface area ! minus the area of'the columns, downcomers and MSRV lines. The blocked area, S, includes all the-flanges and members of the bracing system. The drag coefficient used on the right side of equation (1) is the steady flow, free stream
' drag coefficient of the particular structure being analyzed.
( a G.5-1
ZPS-1-MARK II DAR AMENDMENT 14 APRIL 1981
\
Equation (1) can be rearranged to obtain (V'T D m
= 1+nC Dp Dg (
= fC D f
where f is defined as the blockage correction factor. Since f is proportional to the drag coefficients, and hence to the square of the velocity, the fluid velocity and acceleration are multiplied by the square root of f. For consistency, the same multiplicative factor is used for the velocity and acceleration. As a result, the pool swell loads are calculated by the following equations: Standard Drag Load: F = p (/T V) C D A X Acceleration Drag Load: F = p ( /T a)C V 3 m s () Lift Load: F = p AV CAg 14 where: p is the fluid density f is the blockage correction factor V is the fluid velocity C D is standard drag coefficient which accounts for any interference effects A is the projected cross-sectional area of the structure C7 is the lift coefficient a is the fluid acceleration C, is the inertial coefficient Vg is the structure's volume. G.5-2 _ -}}