ML20004B923

Amend 15 to Mark II Containment Design Assessment Rept

ML20004B923
Person / Time
Site:	Zimmer
Issue date:	05/31/1981
From:	CINCINNATI GAS & ELECTRIC CO.
To:
Shared Package
ML20004B921	List: ML20004B920 ML20004B923
References
NUDOCS 8106010315
Download: ML20004B923 (30)

Text

ZPS-1-MARK II DAR AMENDMENT 15 MAY 1981

(

WM.

H.

ZIMMER POWER STATIO" INSTRUCTIONS FOR UPDATING YOUR DESIGN ASSESSMENT REPORT Changes to the MARK II DAR are identified.ay 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 iii and iv Pages iii and iv Page vi Pages vi and vi (Cont'd)

Pages xiv and xy Pages xiv through xv (Cont'd)

CHAPTER 10 Pages 1.0-1 and 1.0-2 Pages 1.0-1 and 1.0-2 CHAPTER 5.0

()

Pages 5.2-13 and 5.2-14 Pages 5.2-13 through 5.2-14a After Figure 5.2-4 Figure 5.2-5 Pages 5.3-22 and 5.3-23 Pages 5.3-22 through 5.3-23a After Figure 5.3-26 Figures 5.3-27 through 5.3-34 (1 sheet each)

Page 5.4-5 Page 5.4-5 CHAPTER 7.0 Pages 7.3-6 and 7.3-7 Pages 7.3-6 through 7.3-7a f

Appendix E l

Page E.4-1 Page E.4-1 l

[

l

[()

1

$1OGO f 03l5

ZPS-1-MARK II DAR AMENDMENT 15 MAY 1981 TABLE OF CONTENTS (Cont'd)

(~j)

~

PAGE 5.2 SAFETY / RELIEF VALVE (SRV) LOADS - PRESENT DESIGN LOADS (T-OUENCHERS) 5.2-1 5.2.1 Design-Basis SRV Loads - Rams Head 5.2-2 5.2.1.1 Conservatism in SRV - Rams Head Methods 5.2-3

5. 2.1. ?.

Safety / Relief Valve Discharge Cases 5.2-6 5.2.1.2.1 Single Valve Actuation 5.2-6 5.2.1.2.2 Asymmetric SRV Actuation 5.2-6 5.2.1.2.3 Automatic Depressurization System (ADS) 5.2-7 5.2.1.2.4 All Valve Discharge Caces 5.2-7 5.2.1.2.5 Second Actuation 5.2-9 5.2.1.3 Safety / Relief Valve Boundary Loads 5.2-10 5.2.1.4 Submerged Structure Loads 5.2-11 5.2.2 Assessment for SRV Loads - T-Quencher 5.2-11 5.2.2.1 Conservatism of the T-Ouencher Load Definition 5.2-11 5.2.2.2 SRV Quencher Discharge Cases 5.2-11 5.2.2.2.1 Single Valve 5.2-12 5.2.2.2.2 Asymmetric SRV Load 5.2-12 5.2.2.2.3 Automatic Depressurization System (ADS) 5.2-12 5.2.2.2.4 All Valve Discharge 5.2-12

/~T 5.2.2.3 Quencher Boundary Loads 5.2-12

\\#

5.2.2.4 Quencher Submerged Structure Loads 5.2-12 5.2.3 Assessment of NRC Acceptance Criteria - SRV 5.2-13 5.2.4 References 5.2-14a 15 5.3 LOSS-OF-COOLANT ACCIDENT (LOCA) LOADS 5.3-1 5.3.1 Design-Basis LOCA Loads 5.3-3 5.3.1.1 Load Definitions 5.3-3 5.3.1.1.1 LOCA Water Jet Loads 5.3-3 5.3.1.1.2 LOCA Charging Air Bubble Loads 5.3-3 5.3.1.1.3 Pool Swell 5.3-3 5.3.1.1.4 Pool Fallback 5.3-3 5.3.1.1.5 Condensation Oscillation 5.3-4 5.3.1.1.6 Chugging 5.3-4 l

5.3.1.1.7 Lateral Loads on Downcomers 5.3-5 5.3.1.2 LOCA Boundary Loads 5.3-7 5.3.1.2.1 LOCA Water Jet 5.3-7 5.3.1.2.2 Charging Air Bubble 5.3-8 5.3.1.2.3 Pool Swell 5.3-8 5.3.1.2.4 Pool Fallback 5.3-8 5.3.1.2.5 Condensation Oscillation 5.3-8 5.3.1.2.6 Chugging 5.3-8 5.3.1.3 LOCA Submerged Structure Loads 5.3-8 5.3.1.5.1 LOCA Water Jet 5.3-9 5.3.1.5.2 Charging Air Bubble 5.3-9

_/

5.3.1.5.3 Pool Swell 5.3-9 iii

~

2PS-1-MARK II DAR AMENDMENT 15 MAY 1981

)

7.3.2.1 Downcomers 7.3-6 15 7.3.2.2 SRV Piping 7.3-7 7.3.3 References 7.3-7a 7.4 REACTOR PRESSURE VESSEL HOLDDOWN BOLTS 7.4-1 7.5 BALANCE-OF-PLANT (BOP) PIPING AND EQUIPMENT 7.5-1 7.5.1 BOP Piping 7.5-1 7.5.1.1 Evaluation of Bounded Load Combinations (Rams Head Definition) 7.5-1 7.5.1.1.1 Loads and Load Combinations Evaluated 7.5-1 7.5.1.1.2 Drywell Piping 7.5-1 7.5.1.1.3 BOP Piping Inside the Reactor Building 7.5-2 7.5.1.1.4 Wetwell Piping 7.5-3 7.5.1.2 Impact of Change to T-Ouencher Discharge Device 7.5-3 7.5.1.3 Impact of SRV T-quencher and LOCA cn Rams Head Design Basis 7.5-3 7.5.2 Balance-of-Plant Equipment 7.5-5 7.6 NUCLEAR STEAM SUPPLY SYSTEM (NSSS) EQUIPMENT 7.6-1 7.6.1 Piping and Equipment 7.6-1 7.6.1.1 Reevaluation Procedures for NSSS Piping 7.6-3 gx N/

vi

_, _ _ ~ _ - _.,

ZPS-1-MARK II DAR AMENDMENT 15 MAY 1981 TABLE OF CONTENTS (Cont' d)

PAGE 7.6.1.2 Reevaluation Procedures for NSSS Equipment 7.6-3 7.6.1.2.1 Reevaluation of Pipe-Mounted / Connected Equipment 7.6-3 7.6.1.2.2 Reevaluation for Floor / Structure-Mounted Equipment 7.6-3 i

i e

O l

l l

I l

lO 15 vi (Cont' d )

l l

2PS-1-MARK II DAR AMENDMENT 15 MAY 1981 LIST OF FIGURES (Cont'd)

O NUMBER TITLE 3.3-2

• Suppression Pool Pressure 3.3-3 Suppression Pool Temperature Sensors 3.3-4 Temperature and Pressure Sensor Locations 3.3-5 Suppression Pool Strain Gauge Locations 3.3-6 SRV Discharge Line Sensor Locations 3.3-7 HPCS Suction Line Strain Gauge Locations 3.3-8 Downcomer Strain Gauge and Accelerometer Locations 3.3-9 T-quencher Sensor Locations 3.3-10 Quencher Support Strain Ga,,e Locations 3.5-1 SRV Test Schedule 4.0-1 Primary Containment Before Modification 5.2-1 Orientation of SRV Rams Head Devices for Automatic Depressurization Actuation 5.2-2 Reactor Vessel Cycling Steam Pressure vs. Time - ATWS Pump Trip Off 5.2-3 Reactor Vessel Cycling Steam Pressure vs. Time - ATWS Pump Trip On 5.2-4 Cross Section of Suppression Pool and Definition of Suppression Chamber Walls Loading Zones 5.2-5 Comparison of Zimmer Design Basis and NUREG-0487' 15 Suoplement 1

(~')

5.3-1 Drywell/Wetwell Pressure History for Recirculation Line Break (DBA)

's 4

5.3-2 Drywell/Wetwell Temperature History for Recirculation Line Break (DBA) 5.3-3 Drywell/Wetwell Pressure History for Main Steamlinc Break 5.3-4 Drywell/Wetwell Temperature History for Main Steamline Break 5.3-5 Drywell/Wetwell Pressure History for Intermediate Size Break (IBA) 5.3-6 Drywell/Wetwell Temperature History for Intermediate i

Size Break (IBA) 5.3-7 Drywell/Wetwell Pressure History for Small Break (SBA) 5.3-8 Drywell/Wetwell Temperature History for Small Break (SBA) 5.3-9 Vent Clearing Jet Angle i

5.3-10 Jet Impingement Load 5.3-11 Chugging Load 5.3-12 Drag Load Due to Vent Clearing Jet 5.3-13 Maximum Impact Pressure on Small Structures 5.3-14 Maximum Impact Force on Pipes 5.3-15 Maximum Impact Force on I-Beams 5.3-16 Time History of Impact Load for Small Structures in Pool Swell Region 5.3-17 Lateral Loads on Groups of Downcomers at Probability s

Level 10-*

5.3-18 Time Sequence xiv

ZPS-1-MARK II DAR AMENDMENT 15 MAY'1981 LIST OF FIGURES (Cont'd) 7U NUMBER TITLE 5.3-19 Acoustic Load Illustration 5.3-20 Loading Description 5.3-21 Restricted Pipe Motion During Breakdown 5.3-22 Recirculation Line Break 5.3-23 Break Flow vs. Time - Feedwater Line Break 5.3-24 Recirculation Line areak Nodalization 5.3-25 Feedwater Line Nodalization 5.3-26 Drag Coefficients 5.3-27 Zimmer Plot Legend 5.3-28 4TCO Loads vs. Existing Design LOC 104 Vert 5.3-29 4TCO Loads vs. Existing Design LOC 129 Vert 5.3-30 4TCO Loads vs. Existing Design LOC 117 Vert 5.3-31 4TCO Loads vs. Existing Design LOC 111 Vert 15 5.3-32 4TCO Loads vs. Existing Design LOC 137 Vert 5.3-33 4TCO Loads vs. Existing Design LOC 309 Vert 5.3-34 4TCO Loads vs. Existing Design LOC 311 Vert 7.1-1 Zimmer FSI Analysis Model 7.1-2 Average Shear Strain Versus K, 7.1-3 Average Shear Stain Versus Critical Damping 7.1-4 Cross Section of Suppression Pool and Definition of Suppression Chamber Walls Loading Zones

(")

7.1-5 Typical Resonant Sequential Symmetric Discharge

\\s Forcing Function - Zone 4 7.1-6 Typical Asymmetric SRV Discharge Forcing Function -

Zone 4 7.1-7 LOCA Vent Clearing Pressure Loads on Basemat 7.1-8 LOCA Vent Clearing Pressure Distribution 7.1-9 Structural Model Including Soil 7.1-10 Drywell/Wetwell Pressure History for Main Steamline Break 7.1-11 Pool Swell Symmetric Load 7.1-12 Pool Swell Asymmetric Load 7.1-13 LOCA Cyclic Condensation Pressure Load on Basemat Containment and Reactor Support for Rams Head 7.1-14 Spatial Distribution of LOCA Condensation Oscillation Load for T-quenchers l

7.1-15 Chugging Load - Magnitude and Spatial Distribution 7.1-16 Chugging Load - Time History l

7.1-17 Drywell Floor Analytical Model 7.1-18 Circumferential Variation of Moment in Drywell Floor Due'to Radial Moment Applied to Radius 22'-3" 7.1-19 Radial Variation of Moment in Drywell Floor Due to Concentrated Radial Moment Applied at Radius 22'-3" 7.1-20 Circumferential Variation of Moment in Drywell Floor Due to Concentrated Circumferential Moment Applied at Radius 22'-3" 7.1-21 Radial Variation of Moment in Drywell Floor Due to f)

Concentrated Circumferential Moment Applied at Radius 22'-3" xv

a ZPS-1-MARK II DAR AMENDMENT 15 MAY 1981 7.1-22 Critical Design Section 2 Variation of Radial Moment O

with Number of Loaded Dowrcomers 7.1-23 Critical Design Section 2 Variation of Radial Moment with Number of Downcomers Located per DFFR 7.1-24 Basemat Plan Top Reinforcing Layout 7.1-25 Basemat Plan Bottom Reinforcing Layout 1

i O

t i

i i

i O

15 4

xv (Cont ' d)

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

~

2PS-1-MARK II DAR AMENDMENT 15 MAY l'981~

()

CHAPTER 1.0 - INTRODUCTION The purpose of this revision of the Design Assessment Report (DAR) is to demonstrate that the Wm. H. Zimmer Nuclear Power Station, Unit 1 (ZPS-1) containment can accommodate all hydrodynamic load phenomena associated with the SRV dischargc and LOCA in the BWR Mark II containment, to provide evidence of conformance with the NRC Lead Plant Acceptance Criteria (NUREG-0487), and to pr vide a response to the formal questions prsed by the Nuclear Regulatory Commission (NRC).

In the summer of 1979, the Wm. H. Zimmer Power Station (2PS-1) design and construction status was such that additional load changes requiring plant modifications would seriously impact the construction schedule.

To avoid this situation the ZPS-1 "three-pronged" approach was adopted.

The three facets of this approach are:

a.

Expedite construction based on conservative loads and upgrade immediately containment capability where possible.

b.

Assess the plant for the Zimmer Empirical Load Design Basis which is expected to bound any future changes in pool dynamic loads.

()

c.

Confirm adequacy of design with results of the Zimmer in-plant SRV test and the long-term Mark II program.

The Zimmer empirical loads are described in Section 2.1.

This report describes the original design-basis for 2PS-1, subsequent reassessments for revised and newly identified loads, and finally the current reevaluation of the design using the Zimmer empirical loads to ensure the adequacy and conservatism of the containment structures, piping, and equipment.

This report also describes the conformance of the Zimmer design to the NRC Lead Plant Acceptance Criteria (NUREG-0487).

Sub-section 5.2.3 compares the design-basis T-quencher load with 15 the criteria of NUREG-0487 and Supplement 1 of NUREG-0487.

Subsection 5.3.2 compares the design-basis LOCA loads with the criteria in NUREG-0487.

This report provides the NRC staff l

with all information necessary to continue and complete the licensing of the Wm.

H. Zimmer Nuclear Power Station as scheduled.

All pertinent information related to loads, load specification, I

load combinations, acceptance criteria, plant modification, plant margins, and confirmation of loads that apply to ZPS-1 has been compiled in this document.

In addition, an in-plant SRV test will be performed to confirm the adequacy of loads used for design assessment.

OV 1.0-1 i

i ZPS-1-MARK II DAR AMENDMENT 15 MAY 1981 G2 In this report the individual loads and load combinations that are being utilized in the reassessment are identified and described in the first four sections.

Reports defining the individual loads and providing justification for application to the ZPS-1 containment are referenced rather than repeated.

This is consistent with the objective of this report.

The methods used in reevaluating the structures, piping systems, and equipment are described in Chapter 7.0, Fatigue analysis 15 of the downcomers and SRV lines is included in Subsection 7.3.2.

The plant modification and resultant changes that have been completed are described in Chapter 9.0.

The plant margins and conservatisms are summarized in Chapter 10.0.

To fulfill the requirements of NUREG-0487, a description of the assessments 15 used to ensure functional capability of piping systems is in-cluded in Section E.4 of Appendix E.

The long-term Mark II program is expected to confirm that the plant, as presently designed and 'onstructed, is completely safe and adequate.

An interim

'ssment using loads derived from results of the 4TCO tests, usscribed in Subsection 5.3.3, 15 provides additional ensurances.

However, additional design modifications and plant changes are being implemented to utilize the full containment capability.

This ensures that the maximum possible margins are built into the plant, so that C3>

if load definitions should change later, they can be accommo-dated without plant hardware changes.

The ZPS-1 plant startup

• should, therefore, proceed as scheduled.

1.0-2

ZPS-1-MARK II DAR AMENDitENT 15 MAY 1981 5.2.3 Assessment of NRC Acceptance Criteria - SRV The original design methods and the design reassessments described in the above subsections address all Lha NRC concerns in the Lead Plant Acceptance Criteria (NUREG-0487).

An itemized list of the Zimmer Power Station response to the NRC Acceptance Criteria is contained in Section 5.4.

In order to demonstrate the adequacy of the William H.

Zimmer Station design-basis frequency range for the all valve discharge case, a comparison was made between it and the frequency range provided in NUREG-0487, Supplement 1 (September 1980).

The comparisen was accomplished by generating envelopes of the magnitude of the Fourier transforms of the sets of factored design traces produced by the two metheds.

The Zimmer Station design basis is derived from using a 1.5 amplitude multiplier on the three KWU design traces and sweeping a dominant frequency range of 2.9 to 9.9 hertz.

The NRC design basis is derived from using a 1.1 amplitude multiplier on the three KWU design traces and sweeping a dominant frequency range of 3 to 11 hertz.

The resulting design envelopes are compared in the attached figure.

Note that the Zimmer design-basis is 35% higher than the NRC design basis except in the narrow frequency range of 11 to 13 Hz.

In this narrow range, the NRC design basis is, on the average,

{)

P% higher.

In our judgement this small increase over a narrow frequency range 15 is of no design significance.

Our opinion is based on the follow-ing:

a.

The total structural and piping response has contri-butions from several frequencies.

The Zimmer design basis is 35% higher than the NRC design basis except in the 11 to 13 hertz rango where it is 8% lower.

In our opinion any increase in structural or piping response due to a higher 11 to 13 hertz input will be more than compensated by a lower input (and response) for all other frequencies.

b.

The Zimmer design is based on the simultaneous occur-ence of the SSE, LCCA, and the SRV events.

As the combined design response has contribution from all three loads, the increase in the SRV response would be more than compensated for by the conservatism Zimmer has used in other portions of its Empirical Design Basis loads.

This is illustrated in Subsection 5.3.3.

)

5.2-13 l

i ZPS-1-MARK II DAR AMENDMENT 15 MAY 1981 O

c.

ror desien of the Zimmer seation, those 1oads were combined by the abcolute sum method.

A significant reduction would result from the use of the SRSS 15 (Square Root of the Sum of the Squares) method which has been approved for Mark II use by the NRC.

Based on the above discussion, we conclude that the Zimmer design basis is more conservative than required by NUREG-0487, Supplement 1.

1 l

e J

O I,

5.2-14

2PS-1-MARK II DAR AMENDMENT 15 l

MAY 1981 0

5.2.4 References 1.

A. J. James, "The General Electric Pressure Suppression Containment Analytical Model," General Electric Report NEDO-10320, April 1971.

2.

W. J. Bilanin, "The General Electric Mark III Pressure Suppression Containment System Analytical Model," General Electric Report NEDO-20533, June 1974.

l O

I l

l l

l 1

O 15 1

5.2-14a L....

l AMENDMENT 15 COMPARISON OP ZIMMER MAY 1981 DESIGN BASIS AND NUREG-0487, SUPPLEMENT 1 O

I 6

U 4--

I h

1 5

E i

H ZIMMER DESIGN BASIS i

O a

g NUREG-0487 w

2--

l l

0 5

E0

$5 20 0

FREQUENCY (IIERTZ )

WM. H. ZlMMER NUCLEAR POWER STATION, UNIT I MARK la DESIGN ASSESSMENT REPORT FIGURE 5.2-5 O

COMPARSION OF ZIMMER DESIGN BASIS AND NUREG-0487, SUPPLEMENT 1

2PS-1-MARK II DAR AMENDMENT 15 MAY 1981 5.3.2.5 Condensation Oscillation Loads

[}

The condensation oscillation (CO) load definition originally used in the 2pS-1 design is the Mark II DFFR load definition.

However, to account for uncertainties in the load definition and to expedite licensing, a more conservative load definition has been used for reassessment.. This method, called the Zimmer Empirical Approach, is described in Chapter 1.0.

This approach is more conservative than required by NUREG-0487.

5.3.3 ZPS-1 4TCO Data Assessment In mid-1979, the state of construction and schedule of the William II. Zimmer Power Station were such that finalization of the design loads was required to prevent costly delays in completion of the plant.

These design loads are reported in the Design Assessment Report (DAR) as the Zimmer Erapirical Loads.

These loads were purposely defined with added conservatism to accomodate, if necessary, increased loads based on tests which were not complete at that time.

A commitment was also made to confirm that the Zimmer Empirical Load was indeed adequately conservative when results of those tests became available.

The major uncertainty in the pool dynamic loads was the phenomena of LOCA steam condensation.

To better define these loads, the Mark II Owners' Group performed additional tests at the 4T single-()

vent test facility.

The results of this test became available in early 1980.

To confirm the adequacy of the design load defini-ti.ons for Condensation oscillation (CO) and Chugging, new boundary 15 loads were defined based upon the 4TCO data and applied to Zimmer structural models to allow response spectra comparisons.

A CO load based on the 4TCO data was submitted to the NRC in July 1980.

This load was defined as >wo sets of precsure time histor-ics from the 4TCO tests.

The first set (called Col) bounds all the 4TCO data taien under blowdown conditions which could be con-servatively predicted to occur during a LOCA in the Zimmer station.

The second set (called CO2) bounds the data taken under conditions which.could be conservatively predicted to occur coincident with the actuation of the Automatic Depressurization System (ADS).

In October 1980, after review of this load definition, the NRC con-curred that this was an appropriate approach to demonstrate the conservatism of the design basis.

After the lead plant load was reviewed, another CO load based on 4TCO data was presented to the NRC by the Mark II Owners' Group.

This generic CO was essentially the same as the lead plant load except that the elimination of the small amount of non-prototypical (unrealistically high pool

+

temperature) data was lef t to the individual plants as a plant unique application methodology.

If the Zimmer unique selection is applied to the generic load, it becomes essentially identical to tne lead plant load.

1

~

i 5.3-22 m

ZPS-1-MARK II DAR AMENDMENT 15 MAY 1981 A chugging load based on 4TCO data was also defined in July 1980 g)s and finalized in September 1980.

This load is again a set of

(_

4TCO time histories.

The amplitude is a bounding average to conservatively simulate the observed variation in amplitude between chugs.

The chugs were conservatively assumed to be in-phase at all vents.

This load was compared with data from the JAERI (Japanese Atomic Energy Research Institute) multivent test facility and found to be conservative.

After review of this lead plant chugging load, the NRC agreed that this load was adequate to proceed with construction and licensing with the provision that an assessment be made with the Mark II Generic Chugging Load Definition.

Comparison of typical response spectra envelopes are included here.

The two curves represent the Zimmer design basis and a hypothetical design basis utilizing 4TCO data and the SRSS (Square Root of the Sum of the Squares) method of load combina-tions which has been approved by the NRC for Mark II application.

Curve 1 represents an envelope of the Zimmer Empirical Load (ZEL)

Design Basis:

'OBE + SRV (Low Setpoint) + col (ZEL)

SSE + SRV (Low Setpoint) + col (ZEL) 15 envelopeJOBE + SRV (ADS)

+ CO2 (ZEL) lSSE + SRV (ADS)

+ CO2 (ZEL)

()

OBE + SRV (ADS)

+ Chugging (ZEL)

SSE + SRV (ADS)

+ Chugging (ZEL) i Each load combination is combined by Absolute Sum.

The loads are as defined in the Zimmer DAR.

Curve 2 represents an-envelope of loads based on 4TCO data combined with seismic and SRV loads:

'OBE + SRV (Low Setpoint) + col (4TCO)

SSE + SRV (Low Setpoint) + col (4"CO) envelope /OBE + SRV (ADS)

+ CO2 (4TCO) lSSE + SRV (ADS)

+ CO2 (4TCO)

OBE + SRV (ADS)

+ Chugging (4TCO)

SSE + SRV (ADS)

+ Chugging (4TCO) i Each load combination is combined by SRSS.

Although a reduction in the SRV (ADS) amplitude multiplier of about 30% (1.1 in place of 1.5) has been app.cVed by NRC, this reduction was not included in this comparison.

These comparisons show that the design basis is very conservative at frequencies below 50 hertz.

The occasional exceedences above 50 hertz are not considered significant to the design.

O 5.3-23

2PS-1-MARK II DAR AMENDMENT 15 MAY 1981

()

5.3.4 References 15 1.

General Electric Company and Sargent & Lundy, " Mark II 4

Containment Dynamic Forcing Functions Information Report,"

NEDO-21061, September 1976 (Revision 2).

2.

Final Safety Analysis Report, Wm. H. Zimmer Power Station, Chapter 6.0.

3.

General Electric Company and Sargent & Lundy Engineers,

" Mark II Containment Dynamic Forcing Functions Information Report," NEDE-21061-P, September 1976 (Revision 2).

4.

" Analytical Model for Liquid Jet Properties for Predicting Forces on Rigid submerged Structures," NEDE-21472, September 1977.

5.

S. Abramovich and A. Solan, "The Initial Development of a Submerged Laminar Round Jet," Journal of Fluid Mechanics, Vol. 59, Part 4, pp. 791-801, 1978.

6.

" Mark I Containment Program 1/4 Scale Test Report Loads on Submerged Structures Due to LOCA Air Bubbles and Water Jets,"

NEDE-23817-P, September 1978.

n v

O 15 5.3-23a

AMENDMENT 15 MAY 1981 l

3

\\

11 7 l

1 is6 c

O 12 9 13 7 309 l

i

" 10 4 WM. H. ZIMMER NUCLEAR POWER STATION, UNIT 1 MARK 18 DESIGN ASSESSMENT REPORT FIGURE 5.3-27 ZIMMER PLOT LEGEND l

AMENDMENT 15 rateuCNCY IN CPS MAY 1981 500.0 r00 0 100.0 50.0 20 0 10.0 5.0 C.0 10 0.5

.00 ' ' ' " ' " " ' ' ' ' ' '

'''"'""''''t

""*'J (3.,

U 7 00 6.00 5.00 o

w 7-o s

Ca: 4.00 us I -t s u

l l11 o

,g C

(

F

o v

3.00

d ^.

'F i

2.00 I

/

\\

/

\\

/-, \\

\\

/

g

'\\

/

)

\\

g/

N

/

/

\\ \\.

ii 5

_/

\\

il 1.00 v,

.-v s

c y

l_. )

b

/

\\ \\

l

-~/

\\\\

l

.f -

s l

i -.J 0.00 i..

u 1 u

. '. a. ' ' ' u -

'"r*''"".2"""

c.cac c.cn om a Ic3 0 '00 0 f 3 3C

.0 :

.: :.::: c to a.ce PERIOD IN TZC2!.33 WM. H. If MMER NUCLEAA POWER STATION. UNIT 1 MARK li DESIGN ASSESSMENT REPORT l

FIGURE 5.3-28

-3 (V

h

{

s 4TC0 LOADS VS. EXISTING DESIGN LOC 104 VERT

AMENDMENT 15 MAY 1981 ratcutscY IN CPS 50J.O 200.0 100.0 50.0 20.0 10.0 5.0 2.0 10 C.

} $,QQ, m a int ei s t i i

it *,

t e t t

' tis e t e t t it t

• t t

a t t 1. t t o!!!s t f u.LLl1Ll J f

v 14.00 12.00--

,10.00 e

H x= 8.00 I

h u

l u

C l

6.00 O

I 4.00

-~.

r J

2.00 I

/

(

i,

/

j V'%J m

/

N l

x

"; ~

0.00 s u n xa.ua

>w.

unu n=

aw....

.....w...

,,,...a

. 0 0.*

0 04 0.0M 0.C13 C.0M 0. *:40 C.0UJ 0.1LO

3. ?O 0.403 0.GM 1,.

2.*

PER

• 00 11;.LCC;.0J WM. H. ZIMMER NUCLEAR POWER STATION, UNIT 1 MARK 11 DESIGN ASSESSMENT REPORT FIGURE 5.3-29 m

4TC0 LOADS VS. EXISTING DESIGN LOC 129 VERT I

i

AMENDMENT 15 FREQUENCY IN CPS MAY 1981 503.0 200.0 100.0 50.0 20.0 10.0 5.0 E.0 10 0.

".00 o

"'tu"*'*

(v~'t 7 00 6.00 5.00 o

w 2.

o e-.

~

Ta: 4.00 ta tu a

o C

t^

V) 3 00

-~

N x

/s e

x

.JI 2.00

\\,

I s

i

[}

\\_

/

\\

r~ s 1.00 f~

q\\ _1

\\

^

/

\\

/

-m I~.

8 0.00 at.L w '.i J u D G * ' ' UuG_t.u:MuU.1. * ' *

• 'm u_' M Jn' * ' t ' n n t L U lD ' ' * * ' '

.to:

c.;;. c.c.,

t. cia c..c; a.;.a c.tua c.1:c L.;La c..;a 3.t;:

1.u 2.

PER;C3 Ill SEC2::DS WM. H. ZIMMER NUCLEAR POWER STATION UNIT 1 i

MARK 11 DESIGN ASS ESSM ENT REPORT FIGURE 5.3-30 n

N.]

4TC0 LOADS VS. EXISTING DESIGN LOC 117 VERT

AMENDMENT 15 MAY 1981 FREQUENCY IN CPS E30.0 200.0 100.0 50.0 29.0 10.0 5.0 2.0 1.0 0.

8.00. > ' ', i "i n v i t i iii i.

i> i,,

,,n n erii nV 7.00 i

G.00 5.00 o

z O

~

cco:4 00 wJu

~

t-L cr.

9'O l

2-l 3.00 1

s 2.00

/

l

~

l O

/

T \\

1.00 1

- /

V

\\

i

\\

i L

I.'-

1,.

0.00 un iii ii' iiiiin n..

. >.*> i..,,.

=111 1...

.1 4 1u

<>1ii ii1 002 0.C'J 4 0.005 0.0*O G.C O3 0.040 0.053 0.1C0 0.RUC 0 4'JO 0.603 1.0 2.';

PER I O't IN CECCNCS WM. H.ZIMMER NUCLEAR POWER STATION, UNIT 1 MARK II DESIGN ASSESSMENT REPORT iIGURE 5.3-31 Q

4TC0 LOADS VS. EXISTING DESIGN LOC 111 VERT

AMENDMENT 15 MAY 1981 FREQUENCY IN CPS 500.0 200.0 100 0 50.0 20.0 10.0 5.0 2.0 10 0.6 16.00 '

ii,mi,,.i

(

s._

14 00 12 00 10.00

s

")

fm 2 8.00

/

I J

I J

J J

"J T.

m.

J 6.00 i : _n 7

\\

[

4.00 l

iy

/

8

/

/

]

2.00:

/

/

\\

/

1 i )

\\

)'

L

\\

s "A

k 0.00

...........,,.,,,,1,

.03*

0.004 0. COG 0.C10 0.C"3 0.C40 0.053 0 137 0 2CD 0.403 0.503 10 2.3 PCR1C3 D: E ^Cel:3 WM. H. ZIMMER NUCLEAR POWER STATION UNIT 1 MARK 61 DESIGN ASSESSMENT REPORT o

P00R.0RIGINAL

"- ' 3-32 4TC0 LOADS VS. EXISTING DESIGN LOC 137 VERT

AMENDMENT 15 ratouENCY IN CP3 MAY 1981 500.0 200.0 1C0 0 50.0 20.0 10.0 50 f.0 10 0.!

8.QQ si a i a i renve e n a a i i e e e e ie i e i intain ieeei e e i eei t e e et t e t t f f * * ' ?

e t.

l

%)

7 00 6.00 5.00

./

y H

C i

z 4.00 t* 3

\\

1 Y

l l

/

\\ '/

O j.

C 3.00$_

l I h

1 I

i

/I 2.00 l

1 s

o 1.00

'J

\\

/

h

/

/

'/

V h

l

//

\\\\

l d

.l 0.00

.....m.

m.

.00 f 0.C34 0.079 0.0.0 0.003 0.Ce* 0.C:?

C.1C3 0.*03 0.403 0.003 10 a.

fE6100 IN LECLNJ3 WM. H. ZIMMER NUCLEAR POWER STATION. UNIT 1 MARK 11 DESIGN ASSESSMENT REPORT o_

P00R ORIGINAL "eveE e.2->>

4TC0 LOADS VS. EXISTING DESIGN LOC 309 VERT

AMENDMENT 15 MAY 1981 FREQUENCY IN CPS 500.0 EUU.0 100.0 50.0 20.0 10.0 5.0 2.0 10 C.

8 00 u1_1_s i n.iin r i i ie u n sii i i.gnig gi_,_i_ur w _,,,1uinn,i q.

, i r

il Og 7 00 6.00 5 00 e

w 2

3 H

B C"4 00 e

us h

'A v

l a' -

3.00_

l I l

/Q l

I r

I if i

(

2.00

\\ \\

k

\\

7 r

\\

/

\\t l

1.00_.

l

//

v

\\\\

l

---g/

g l

\\ N l

l 3

l ',

0.00

, i,, 2 a

....c.,i.:i n.,,:,iii. inn

.,nii,:

,,ii

iii, in,

. Cot 0.C04 0.CC6 0.010 0.000 0.U40 0.C30 C.1CO 0.tc3 0.403 0.C 3 1.C E.'

PERIOJ lN SEC3N33 WM. H. ZIMMER NUCLEAR POWER STATION, UNIT 1 MARK li DESIGN ASSESSMENT REPORT FIGURE 5.3-34

,{}

4TC0 LOADS VS. EXISTING DESIGN LOC 311 VERT

l i

(

(

/

s-TABLE 5.4-1 (Lont'd)

MARK 11 OWNERS GROUP LOAJ OR P3F;NOMENON

__ LOAD SPECIFICATIGN NJLC REVIEW STATUS ZIMMER _ POSITION ON ACCEPT. ICE CRITERI A.

- asymmetric loading Maximum amplitude uniform below condition vent exit-linear attenuation to pool surface. 20 psi maximum overpressure. -14 psi maximum underpressure, 20-30 Hz f re-quency, peripheral variation of amplitude follows observed statistical distribution with maximum and minimum dia-metrically opposed.

II. SRV-Related Hydrodynamic Loads A.

Pool Temperature Limits for None specified NRL Criteria 11.1 Acceptable FWU and CE four-arm quencher and 11.3 (1)

Quencher Air Cleasing Loads The methodology documented in NRC Criteria in The Ziumer station is being assessed for the T-quencher l15 the Susquehanna DAR is used Section II.B.5 of loads. These loads are considered to be conservative for the T-quencher locd Supplement I to and demonstrate the adequacy of the Zimmer design. A definition.

NUREG-0487 (2) presentatioa on the impact of modifications to the SRV frequency range was given in the February 13, 1979 g

meeting. Results of an assessment of the SRV T-quencher 4

frequency range was presented at the July 26, 1979 h'

v.

meeting. The conservatism of the T-quencher load in both "s-amplitude and frequency is described in Subsection 5.2.2.1 M

r. etappr a.cn of the f requency range with the criteria r.

in SUktG-0407 is included in Subsection 5.2.3.

15 o

A further demonstration of the conservatism of the lead plant approach has been documented by Lorg Island Lighting Co. (SNRC-374, March 30, 1979, Mr. Novarro

[LILC0] to Mr. S. A. Varga [NRC] transmAtting a report entitled " Justification of Mark 11 Lead Plant SRV Load maqmg Definition.")

gu'"yg 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

4) Four-Arm Quencher Vertical and lateral ara loads Acceptable Acceptable amm'a developed on the basis of bound-Ing assumptions for air / water discharge from the quencher and

([]"$)

conservative combinations of maximum / minimum bubble pressere gg mape;p acting on the quencher.

dM 5

b) KWU T-Quencher T-quencher arm loads, as specified Acceptable Acceptable. These loads have been calculated using the

$m l'd'

methodology and assumptions described in DFFR for four-arm N

for SSES. (3) quenchers, as recommended in the Acceptance Criteria. KWU T-quencher rethods were used to verify conservatism.

ZPS-1-MARK II DAR AMENDMENT 15 MAY 1981

(])

The inertia effects of the water surrounding the submerged portion of the structural elements were simulated by the addition of a water mass (Reference 1) equivalent to the displaced volume of the structural elements.

The mass of water inside the submerged portion of the downcomers was also considered in the model for all dynamic loadings except the loads associated with the LOCA.

For these loads, the water had been vented from the downcomer and, therefore, it was not included in the model.

Depending upon the form of the loading function, both response spectrum and forced vibration methods were used to obtain the structural response.

7.3.2 Fatigue Evaluation of SRV Discharge Piping and Downcomer Vents in the Wetwell In response to NRC concerns on bypass capability and cyclic loading on SRV piping and downcomers in the wetwell, the Mark II Owners' Group have addressed the fatigue analysis of these com-ponents.

The design bases for these components are defined in the appropriate FSAR sections.

Detailed evaluations have shown these design bases to be appropriate and conservative.

The Mark II Subcommittee on SRSS and Load Combinations / Acceptance Criteria has coordinated a consistent approach to the evaluations.

Code Class 1 piping fatigue rules and methods have been employed.

O' As required, mill certification material test results have been used in order to define better allowable stresses.

The fatigue analyses have included all cyclic loading due to SRV discharge, chugging, and condensation oscillation.

The 40-year plant life 15 was divided into sub-events, 1 postulated LOCA, 1 SSE, and 5 OBE's.

The cycle ( 'mination method was used in calculating a cumulative fatigue usage factor.

The bRSS method is used for combining dynamic responses.

Detailed analyses for all significant thermal transient effects were also performed and considered in the fatigue evaluations.

The fatigue evaluations for the Zimmer SRV lines and downcomerr are outlined as follows.

7.3.2.1 Downcomers Downcomers are subjected to cyclic hydrodynamic loading, but they experience no significant thermal transients.

Without thermal transient effects, the code class 2/3 piping rules (downcomer design basis) have been shown to be conservative for fatigue considerations.

7.3-6

ZPS-1-MARK II DAR AMENDMENT 15 MAY 1981 The two governing locations on the downcomer were determined

(])

to be the drywell floor anchor point and the downcomer bracing elevation.

The appropriate numbers of sub-event load cycles were determined at each of these locations.

Assuming the downcomer Class 2/3 design-basis stress levels to be at the maximum allowable limits, cumulative fatigue usage factors were calculated.

The results are shown below.

l = 0.530 (drywell floor anchor point) u 2 = 0.124 (bracing elevation) u 7.3.2.2 SRV Piping The two low set-point SRV discharge lines were determined to be governing, due to their much greater number of actuations over the 40-year plant life.

The number of expected SRV cycles is obtained from the DFFR, Revision 3.

The five critical locations on the piping were determined:

a.

drywell floor penetration anchor, b.

reinforced tee connection at top of vertical riser, 15 c.

12 inch tc 10 inch transition joint below drywell

floor, 7-V d.

highest stressed elbow, and vertical riser pipe at high water level.

e.

Fatigue requirements have been incorporated into the recently revised drywell floor penetration design (item a).

The pene-tration with governing piping reaction loads has been analyzed, and the acceptable fatigue results are documented in the penetration stress report.

Item b. was found to be the governing location, due to high peak stresses from internal pressure and thermal effects.

The cumulative fatigue usage factor for the tee connection of the governing low set-point line is 0.438 (allowable = 1.0).

Usage

[

factors for piping locations 3, 4, and 5 were shown to be l

negligible ( u<0.05).

The SRV line fatigue evaluation will be included in the formal piping stress analyses and verification of as-built conditions.

No significant changes from the current fatigue results are expected.

I

~T (V

7.3-7

~

n-4 n

-m--,,

w 2PS-1-MARK II DAR AMENDMENT 15 MAY 1981 O

7.3.3 References 15 1.

Sir Herace Lamb, " Hydrodynamics," Sixth Edition, Dover Press, New York, 1945.

2.

ASME Boiler and Pressure Vessel Code,Section II, Subsections NC, NF, and appendices of Division 1, 1978 edition, including the Winter Addenda.

i 1

O i

l O

7. 3 - -7a

.-..-~-

=_

.. _ = _.

ZPS-1-MARK II DAR AMENDMENT 15 MAY 1981 sm E.4 FUNCTIONAL CAPABILITY 1

The piping analysis on the Wm. H. Zimmer Nuclear Power Plant l

complies with the GE Report, "Functiornal Capability Criteria 15 for Essential Mark II Piping", NEDO-21985, September 1978, as described in the topical report evaluation (1) by the NRC

,^

dated July 19, 1980.

This is applied to all essential lines for which functional capability must be ensured.

i i

i l

.i s

i

!lO 1

i 1

l I

i O

15 (1) Memorandum from J.

P. Knight of NRC to R. L. Tedesco of NRC.

l l

E.4-1

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