RBG-16-354, Forwards Responses to Draft SER Open Items Re Mechanical Fracturing Analysis & Fuel Assembly Damage by External Forces.Response to Item 10 Re TMI Item II.F.2, Inadequate Core Cooling, Will Be Provided by Dec 1983

Forwards Responses to Draft SER Open Items Re Mechanical Fracturing Analysis & Fuel Assembly Damage by External Forces.Response to Item 10 Re TMI Item II.F.2, Inadequate Core Cooling, Will Be Provided by Dec 1983

ML20082A107
Person / Time
Site:	River Bend
Issue date:	11/11/1983
From:	Booker J GULF STATES UTILITIES CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
TASK-2.F.2, TASK-TM RBG-16-354, NUDOCS 8311180048
Download: ML20082A107 (29)

Text

_ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . .

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GULF STATES UTILITIES COMPANY '

POST o F FIC E B o x 2 9 51 .eEAUM7NT. TEXAS 77704 A ft iA CODE 409 838 6631 November 11, 1983 RBG - 16,354 File Code G9.5, G9.8.6.1 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Denton:

River Bend Station Units 1 and 2 Docket Nos. 50-458/50-459 Enclosed for your review are Gulf States Utilities Company's responses to Draft Safety Evaluation Report (DSER) open items identified by the Nuclear Regulatory Commission's Core Performance Branch (CPB) and positions to the Licensing Review Group - II issues 1-CPB through 11-CPB. Attachment 1 is a summary listing of the items discussed in Attachment 2. Attachment 2 provides the response and reference material for each item. Where indicated, these responses -

will be provided in a future amendment to the FSAR.

Sincerely, J. E. Booker Manager-Engineering Nuclear Fuels & Licensing River Bend Nuclear Group J2A P JEB/ ERG @TJEP Enclosures 8 DOI

/ 40 8311180048 831111 PDR ADOCK 05000458 E PDR

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ATTACHMENT 1-L DSER.

SECTION SUBJECT FSAR REVISIONS

1.  ; 4. 2'.1. 2 (8) Mechanical Fracturing Analysis June, 1984

?- pg. 4-7 Fuel Assembly Damage by External Forces Enclosure 2 4.2.1.3(4). (LRG II 2-CPB)-

Pg. 4-8 4.2.3.2(8),

pg.'4-21 4.2.3.3(4) ,

pg. 4-27 4.2.5 pg. 4-29

2. 4.2.3.1(3) Channel Box Wear and Cracking ' Enclosure 2 pg. 4-10 (LRG II 3-CPB)

• 4.2.5 i

pg. 4-29

3. 4.2.3.1(8) Control Material Leaching pg. 4-17

4. 4.2.4.3 Post Irradiation Honitoring Program December, 1983 l pg. 4-29

.t i S. 4.2.5 Fuel Rod Bowing a pg. 4.29

6. 4.2.5 Overheating of Gado11nia Fuel Pellets 4

pg. 4-29

7. 4. 4.4 . FABLE Code Decay Ratio pg. 4-43 Stability Analysis Prior to Second Cycle.

i

8. 4.4.4 Analysis for Single-Loop Operation pg. 4-44

'9. 4.4.6 Loose Parts Monitoring System Enclosure 1 pg.~ 4-44, 47

10. 4.4.7 Inadequate Core Cooling pg. 4-44, 47 (LRG II 6-CPB; TMI ITEM II.F.2)

11. LRG II Positions 1-CPB through 11-CPB Enclosure 2 1

1

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ATTACHMENT 2 Responses to DSER Open Items

1. DSER (page 4-7, 4-21, 4-29) - Mechanical Fracturing Analysis DSER (page 4-8, 4-27, 4-29) - Fuel Assembly Damage by External Forces

Response

The response to this request is provided below in Item 11 (LRG-II 2-CPB). The plant specific results from horizontal and vertical acceleration will be provided by June, 1984.

2. DSER (page 4-10, 4-29) - Channel Box Wear and Cracking pesponse The response to this request is provided below in Item 11 (LRG-II I-CPB)

3. DSER (page 4-17) - Control Material Leaching

Response

During the NRC Core Performance Branch meeting of September 14, 1983 the Staff indicated that this issue has been CLOSED as the result of review of.NEDE-24325.

4. DSER (page 4-29) - Post Irradiation Monitoring Program

Response

The response to this request will be provided by December, 1983.

5. DSER (page 4-29) - Fuel Rod Bowing

Response

During the NRC Core Performance Branch meeting of September 14, 1983 the Staff indicated that this issue has been CLOSED as the result of NEDE-24284.

6. DSER (page 4-29) - Overheating of Gadolinia Fuel Pellets

Response

During the NRC Core Performance Branch meeting of September 14, 1983 the Staff indicated that this issue would be CONFIRMATORY

.pending review of NEDE-20943-P. No GSU action is required.

i

7. DSER (page 4-43) - FABLE Code Decay Ratio DSER (page 4-43) - Stability Analysis Prior to Second Cycle

Response

The Staff indicated that they will require a License Condition for operation pending submittal and approval of a new stability analysis prior to second-cycle operation.

8. DSER (page 4-44) - Analysis for Single-Loop Operation Response ,

• As The Staff indicated that they will require a Technical Specification precluding single-loop operation until supporting analysis is provided and approved.

9. DSER (page 4-44, 4-47) - Loose Parts Monitoring System response The response to this request is provided in the revised response to Questions 492.2 and 640.1 (Enclosure 1) which will be incorporated into the FSAR in a future amendment.

Additionally, the LPMS continuously monitors the reactor for indications of loose parts. The LPMS will consist of sensors, signal conditioning equipment, recorders, alarms, a loose parts locator, and calibration equipment. The following features will be incorporated into the design of the LPMS:

a. A minimum of two sensors will be located at each anticipated loose part natural collection region.

-b. The onlin'e sensitivity of the system will be such that the system can detect a metallic loose part that weighs from 0.25 to 30 pounds and impacts with a kinetic energy of 0.5 ft-lb on the inside surface of the reactor coolant pressure boundary within 3 feet of a sensor.

! c. The instrumentation channels (e.g., cabling, amplifiers) associated with the sensors at each natural collection l region will be physically separated from each other starting at the sensor locations to the charge converter. The charge converters will be located at points in the plant that are always accessible for maintenance during full power operation.

d. The LPMS will include provisions for both automatic and nanual startup of data acquisition equipment with automatic activation in the event the altrt level is reached or exceeded. An alert condition will be both audibly and visually alarmed on the system annunciator.

L

e. A reference signal level will be incorporated into the LPMS that is indicative of the presence of a loose part.

f.. Provisions will exist for periodic online channel check and functional tests and for offline channel calibration during periods of cold shutdown or refueling.

g. ' Provisions will be made to avoid false alert signals during normal plant maneuverr.

h. The loose part detection system components located inside the containment will be designed and tested to withstand the accelerations of seismic events up to and including the OBE.

i. All LPMS components will be designed for a 40 year life expectancy and to operate in the environmental zones in which the components are installed. All components will use solid state circuitry which provides high reliability, s

j. . Sensor locations are as follows:

1. Vessel Bottom - Sensors will be mounted on the control rod drive housings as close as possible to the vessel.

2. Exact locations for additional sensors have not been determined at this tine. These additional locations will be selected to monitor the upper RPV plenum while locating the sensor in an environmentally acceptable area.

10. DSER (page 4-44, 4-47) - Inadequate Core Cooling (II.F.2)

Response

The response to this request will be provided by December, 1983.

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11. The endorsement to LRG-II positions 1-CPB through 11-CPB is provided below. This table and the discussions provided in Enclosure 2 will be incorporated into the FSAR in a future amendment.

Item Title jndorsed FSAR Discussion 1-CPB Clad Ballooning and Rupture Yes ----------

.2-CPB Seismic and LOCA Loads on Fuel Yes 4.2.3.2.15 3-CPB Channel Box Deflection Yes 4.2.1.2.1.9 4-CPB High Burnup Fission Gas Release Yes ----------

5-CPB . Cladding Water-Side Corrosion Yes 10.4.6.2 6-CPB Inadequate Core Cooling Instrumentation Yes APP 1A 7-CPB Rod' Withdrawal Transient Analyses No

• 8-CPB Mislocated or Misoriented Fuel Bundles Yes - ------

9-CPB Void Coefficient Calculation Yes- 4.3.2.4.2 m

10-C,PB , Bounding Rod Worth Analysis Yes 15.4.9.3.1 11-CPB [ Core Thermal-Hydraulic Stability Yes --

• The LRG-II 7-CPB issue is satisfied by Technical Specifications

.(STS 3/4.1.4) which prohibit rod withdrawal at indicated power levels above the low power setpoint of the Rod Control and Information System if the bypass valves are open. FSAR Section

~'g' 7.6.1.7 is_ currently being revised to address this issue and will be incorporated in a future amendment.

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ENCLOSURE 1 RBS FSAR QUESTION 492.2 (4.4.6)

The vibration and loose-parts monitoring equipment to be provided in the plant should be described. The procedures to be used to detect excessive vibration and occurrence of loose parts should be discussed.

RESPONSE

The response to this request regarding vibration monitoring is provided in Sections 1.5.1.2.1 and 3.9.2.3B through 3.9.2.6B, - and in Table 1.8-1 discussions of Regulatory Guides 1.20 and 1.68.

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Amendment 7 Q&R 4.4-4 February 1983

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(Page Q&R 4.4-4)

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INSERT to Q492.2 The response to this request regarding loose parts monitoring equipment is provided in Section 4.4.6.1.

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RBS FSAR CHAPTER 4 TABLE OF CONTENTS (Cont)

Section Title Paqe

4. 4. 3. 6 Thermal-Hydraulic Characteristics Summary Table 4.4-17 4.4.4 Evaluation 4.4-17 4.4.4.1 Critical Power 4.4-17 4.4.4.2 Core Hydraulics 4.4-17 4.4.4.3 Influence of Power Distributions 4.4-17 4.4.4.4 Core Thermal Response 4.4-17 4.4.4.5 Analytical Methods 4.4-17 4.4.4.5.1 Reactor Model 4.4-18 4.4.4.5.2 System Flow Balances 4.4-19 4.4.4.5.3 System Heat Balances 4.4-20 4.4.4.6 Thermal-Hydraulic Stability Analysis 4.4-21 4.4.4.6.1 Introduction 4.4-21 4.4.4.6.2 Description 4.4-22 4.4.4.6.3 Stability Criteria 4.4-23 4.4.4.6.4 Mathematical Model 4.4-23 4.4.4.6.5 Analytical Confirmation 4.4-24 4.4.4.6.6 Analysis Results 4.4-25 4.4.5 Testing and verification 4.4-26 ( -

4.4.6 Instrumentation Requirements 4.4-26 l 4 4 fe I Loose. P.As A.libr*.n3 System u.4-zu 4.5 REACTOR MATERIALS 4.5-1 l

4.5.1 Control Rod System Structural Materials 4.5-1 l 4.5.1.1 Material Specifications 4.5-1 l 4.5.1.2 Austenitic Stainless Steel Components 4.5-3 4.5.1.3 Other Materials 4.5-4 l 4.5.1.4 Protection of Materials During l Fabrication, Shipping, and Storage 4.5-4 4.5.2 Reactor Internal Materials 4.5-5 4.5.2.1 Material Specifications 4.5-5 4.5.2.2 Controls on Welding 4.5-8 i 4 . 5. 2. 3 Nondestructive Examination of Erought Seamless Tubular Products 4.5-8 4.5.2.4 Fabrication and Processing of Austenitic Stainless Steel -

l Regulatory Guide Conformance 4.5-8 4.5.2.5 other Materials 4.5-10 4.5.3 Control Rod Drive Housing Supports 4.5-10 4-viii (s, l

RBS FSAR the following reactor vessel parameters is provided in the main control room and is discussed in Chapter 7.

1. Reactor vessel water level 2

2. Reactor vassel differential pressures

3. Reactor vessel internal pressure

4. Neutron monitoring system.

Insert new Section 4.4.6.1 here

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(j Amendment 2 4.4-27 February 1982

Insert (Page 4.4-27) 4.4.6.1 Loose Parts Monitoring System The Loose Parts Monitoring System (LPMS) is designed to detect loose parts in the reactor pressure vessel and to provide early warning to the operator so that damage to or malfunctions of safety-related primary system components may be avoided or mitigated. The LPMS is not considered to be a safety-related system and is designed to require minimum' operator interfacing during normal operation.

Provisions are made to properly analyze and investigate potential loose parts. Since the analysis of potential loose part signals requires comparison with baseline signals, the LPMS will be operational and capable of. recording vibration signals for signature analysis during initial reactor start-up testing. The LeMS is designed to meet the requirements of Regulatory Guide 1.133.

F

RBS FSAR

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TABLE 1.8-1 (Cont)

Regulatory Guide 1.133 (Sectember 1977) (For Comment)

Loose-Part Detection Program for the Primary System of Light-Water-cooled Reactors Proiect Position -m.'_Pherc

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FSAR Section - 4.4.6.1 i

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V 175 of 193

0 , s RBS FSAR ,

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( QUESTION 640.1 (1.8) (14.2.7) ,

Regulatory Guide 1.68 - Rev. 2 (Tab'le 1.8-1).

Address the following:

1. Modify your exceptions to Appendix A, Positions 1.c and 5.g.g, to provide a commitment to include in your test program any design features to prevent or mitigate anticipated transients without scram (ATWS) that may in the future be incorporated into your plant design.

2. The staff recognizes that there is currently no Loose Parts Monitoring System at the River Bend Station. Commit to the test requirements of Appendix A, Positions 1.j(6) and 5.n, for such systeme that may be installed in the future.

3. Verify that sources of water used for long-term core cooling are tested to demonstrate adequate NPSH and the absence of vortexing over range of basin level from maximum to the minimum calculated 30 days following LOCA. (Clarification of Appendix A, Position 1.h(10).)

( RESPONSE

1. The exceptions to Regulatory Guide 1.68 Appendix A, Positicns 1.c and 5.g.g have been deleted.

Replace 2. The rerpenee te thir requ:rt is previded in with Table 1 C 1. g EuaulatesV Cuido 1 00, E;7i 3iOT- 2, Insert Itca II 0- 1 i

3. Section 9.2.5.2 dercribes the ultimate heat sink and its design to provide sufficient NPSH and vortex-free operation. Therefore, no testing is planned to demonstrate these criteria.

Amendment 5 Q&R 1.8-3 August 1982

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

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INSERT'to Q640.1-(Page Q&R 1.8-3)--

I- ~~A loose parts monitoring system .(LPMS) is provided in the. River ' end Station design. .See revised Section 4.4.6.1, Table 1.8-1 and

. 14.2.12.1.69.

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TABLE 1.8-1 (Cont) inspected for expansion during preoperational testing when auxiliary steam is introduced for RCIC turbine checks. To the extent applicable, emergency core cooling systems, such as ADS (which includes safety / relief valves), are checked for expansion (e.g. hanger or restraint movement) during initial heatup after fuel load.

RHR shutdown cooling lines are inspected during the RHR initial startup test. With regard to NSSS piping in the drywell, such as main steam line and reactor recirculation piping, expansion data is collected during initial heatup and operation and is compared against acceptance criteria.

?. Appcndin A, paragraphs 1.j 'C) 'p.1.CC 11) ard s e _ ,_ , en ,,

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4 -6'. Appendix A,_, paragraphs 1.m (4) (p.1.68-12) and 1.o(1) (p.1.68-13) ls Regulatory Guide 1.104 was withdrawn by the NRC on August 22, 1979 (reference: 44 FR 49321).

During preoperational testing, fuel handling and vessel servicing equipment are verified to function per test specifications. The controls, interlocks, and travel limits of the reactor building polar crane and of the fuel handling cranes are verified.

5 -6". Appendix A, paragraph 4.m (p. 1.68-15) s There is no startup test of the MS-PLCS (after fuel load). The preoperational test of the main steam-positive leakage control system demonstrates operability at the approximate design conditions, which exist for approximately

, Amendment 5 97 of 193 August 1982

(

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pr 7

. RBS'FSAR TABLE 1.8-1 (Cont)

'20 minutes following a LOCA (i.e., until RPV pressure is zero). Therefore, testing after fuel load _does not contribute any additional, meaningful data.

6 -Y'. Appendi:t A, paragraph 5.j (p.1.68-17) 5 At River Bend Station there is no design for rod runback or partial scram, thus no startup test is performed.

FSAR Sections - 3.9.2A, 7.1.2, 8.2.1, 14.2.1, 14.2.4, 14.2.10, 14.2.12 I

3 Amendment 5 78 of 193 August 1982

RBS FSAR 7 CHAPTER 14 TABLE OF CONTEN;S (Cont)

Sect' ion Title Page

. 14.2.12.1.49 Turbine Building Ventilation System Acceptance Test 14.2-103 14.2.12.1.50 Instrument and Service Air System Acceptance Test 14.2-105 14.2.12.1.51 13.8-kV and 4160-V Distribution System Preoperational Test 14.2-107 14.2.12.1.52 Drywell Leakage Test 14.2-108 14.2.12.1.53 Containment Structural Integrity Test and Containment Leak Rate Preoperational Test 14.2-109 14.2.12.1.54 Seismic Instrumentation Pre-operational Test 14.2-109 14.2.12.1.55 Communications Systems Acceptance Test 14.2-110 14.2.12.1.56 Cranes Acceptance Test 14.2-111 14.2.12.1.57 Circulating Water System Acceptance Test 14.2-111 14.2.12.1.58 ' Condensate Acceptance Test 14.2-113 14.2.12.1.59 Reactor Plant Sampling System Acceptance Test 14.2-114 20 14.2.12.1.60 Turbine Plant Sampling System

<N' Acceptance Test 14.2-115 14.2.12.1.61 Liquid Radwaste Sampling System Acceptance Test 14.2-115 14.2.12.1.62 120-V AC Power Distribution

. Preoperational Test 14.2-116 14.2.12.1.63 Main Steam System Acceptance Test 14.2-117 14.2.12.1.64 Turbine Control Acceptance Test 14.2-118~

14.2.12.1.65 Solid Radwaste Systen Acceptance Test 14.2-119 14.2.12.1.66 Emergency Lighting System i

Acceptance Test 14.2-119 l

14.2.12.1.67 Main Condenser Air Removal System Acceptance Test 14.2-121 l Vents and Drains Systems i 14.2.12.1.68 i Acceptance Test 14.2-123

[ ~~iI.2.12.2 Initial Startup Test Phase Discussion 14.2-124 14.2.12.3 Initial Startup Test Procedures 14.2-127 14.2.12.3.1 Test Number 1 - Chemical and Radiochemical 14.2-127 g 14.2.12.3.2 Test Number 2 - Radiation g Measurement 14.2-128 Amendment 5 14-v August 1982 14.2.12.1.69 Loose Parts Monitoring System-Preoperational Test j

-maJ RBS FSAR

c. Functionally demonstrate the ability of the systems to properly collect and dispose of

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drainage.

4. Acceptance Criteria

a. Interlocks, controls, and alarms performances are as specified by the system elementary diagrams.

s b. Pump performance is comparable to that shown in the manufacturer's technical instruction manual.

ID45ECT A 14.2.12.2 Initial Startup Test Phase Discuccion

1. Startup Test Procedure All those required tests comprising the initial startup test phase are discussed in Section 14.2.12.3. For each test a description is provided for test objective, test prerequisites, s test procedure, and a statement of test acceptance criteria, where applicable.

The operating power-flow map is presented as _

Fig. 14.2-4. The test conditions are marked on Fig. 14.2-4, and each test described in Section 14.2.12.3 is accomplished at the test conditions stated in Fig. 14.2-5.

The acceptance criteria section of each test has one or two sections. The following two paragraphs describe the degree of each kind of test criterion and the actions to be taken after an individual 5 criterion is not satisfied.

a. Level 1 If _ Level 1 test criterion is not satisfied, the plant is placed in a hold condition that is judged to be satisfactory and safe, based upon prior testing. Plant operating or test procedures or the Technical Specifications may guide the decision on the direction taken.

Startup tests consistent with this hold condition may be continued. Resolution of the problem is immediately pursued by appropriate equipment adjustments or through engineering support by offsite personne'l if needed.

Amendment 5 14.2-124 August 1982

(

INSERT (For Page 14.2-124) 14.2.12.1.69 Loose Parts Monitoring System TEST OBJECTIVES

1. To demonstrate proper operation of the Loose Parts Monitoring Equipment.

2. To collect data to use as baseline information during subsequent operation.

3. To verify alert level.

PREREQUISITES AND INITIAL CONDITIONS The Loose Parts Monitoring System is connected to the reactor and instrumentation and control testing is complete.

TEST PROCEDURE

1. Indication and alarm functions will be demonstrated.

2.- The reactor recirculation pumps will be operational for portions of the test.

3. Test results will be reviewed and corresponding alert and alarm setpoints established.

ACCEPTANCE CRITERIA Alert and alarm functions will perform within design tolerances.

. a t ENCLOSURE 2 ,

RBS FSAR

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of energy that has been released at 0.1 sec is 0.4 percent of. the total energy that has been released by tLa bundle (6 MWx0.1 sec) . Note that the fractional energy release decreases rapidly with time even though a constant temperature is maintained. This is because the reaction is self-limiting as previously discussed.

The amount of energy released is dependent on the -

temperature transient and the surface area that has.

experienced heatup. This, of course, is dependent on the -

initiating transient. For example, if boiling transition , , ,

were to occur during steady-state operating conditions, the #

. cladding surface temperature would range from 1,000 to 1,5000F depending on the heat fluxes and heat transfer coefficient. Even assuming all rods experience boiling transition instantan eously, the magnitude of the energy release is seen to be insignificant. Significant boiling transition is not possible at normal operating conditions or under conditions of abnormal operational transients because of the thermal margins at which the fuel is operated. It can, therefore, be concluded that the energy release and potential for- a chemical reaction is not an important consideration during normal operation or abnormal

,, transients.

ac.

+99 4.2.3.2.12 Fuel Rod Bchavior Effects from Coolant Flow Blockage The behavier of fuel rods in the' event of coolant flow blockage is covered in Reference 19.

4.2.3.2.13 channel Evaluation j ' Channel analytical models and evaluation results are

• contained in Reference 20. ,

l 4.2.3.2.14 Fuel Shipping and Handling Analyses of the major handling loads have been performed and the resulting fuel assembly component stresses are within design limits. Additional information on fuel handling and shipping loads is presented in Reference 33.

4.273.2.15 Fuel Assembly - SSE and LOCA Loadings An evaluation of combined SSE and LOCA loadings are contained in Reference 32. g

Inser,t ,j l

c 4.2-37 ,

9

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~ INSERT'(Page 4.2-37) for 2-CPB An additional evaluation to calculate the fuel' assembly dynamic

-responses during a combined SSE.and LOCA event are contained.in Amendment 3 to Reference.32. The results conclude that the GE BWR 4,

, 5, and 6 fuel responses are within acceptable limits and the dynamic and impact loading capability is in excess'of.the predicted loads.

These.results bound plant unique results which will be.provided in a

. future amendment.

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7 RBS FSAR above 3,0000F. The above basis has been demonstrated by experiment to be conservative over (

the complete range of design temperature and expostre conditions ( 7) .

3. Plenum Creepdown and Creep Collapse Creepdown and creep collapse of the plenum are not 9

conside, red because significant creep in the plenum

~ region is not expected. The fuel rod is-designed ,

to be free-standing throughout its lifetime. The -

temperature and neutron flux in the plenum region ,. ~ ,, ,

are considerably lower than in the fueled region, '

thus the margin to creep collapse is substantially greater in the plenum. Direct measurements of irradiated fuel rods have given no indication of significant creepdown of the plenum.

4.2.1.2.1.9 Deflection The operational ' fuel rod deflections considered are the deflections due to:

1. Manufacturing tolerances

2. Flow-induced vibration c 3. Thermal effects

4. Axial load.

There are two criteria that limit the magnitude of these -

deflections. One criterion is that the cladding stress limits must be satisfied; the other is that the fuel rod-to- ,

fuel -rod and fuel rod-to-channel clearances must be 1 sufficient to allow free passage of coolant water to all, heat transfer surfaces.

The fuel rod-to-fuel rod spacing limit of 0.060 in and fuel ,

rod-to-channel spacing limit of 0.030 in are based upon the range of clearances that have in the past been used in l

boiling transition testing. More recent testing to clearances below these values would indicate that a Lower limit is acceptable.

2: - Inser't ,

4.2.1.2.1.10 Fretting Wear and Corroulon

Fretting- wear and cocrosion have been considered in establishing the fuel mechanical design basi ?. Individual rods in- the fuel assembly are held in position by spacers 4.2-12

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A e e: s' INSERT-(Page 4.2-12) for 3-CPB The following general guidelin3s minimize the potential for and detect the onset of channel bowing:

A. Records will be kept of channel location and exposure for each operating cycle.

B. Channels do not reside in the outer row of the core for more than two operating cycles.

C. Channels that reside in the periphery (outer row) for more than one cycle are situated each successive peripheral cycle which rotates the channel so that a different side faces the core edge.

D. At the beginning of each fuel cycle, the combined outer row residence time for any two channels in any control rod cell do not exceed four peripheral cycles.

Af ter core alterations (i.e., reload) and before reaching 40 percent thermal power, a control rod drive friction test is performed for those cells exceeding the above general guidelines or containing fuel channel with exposures greater than 30,000 mwd /T (associated fuel bundle exposures) to provide adequate assurance of the scram function.

.y .

RBS PSAR O more stringent quality control requirements during fuel fabrication. Excessive deposition of corrosion products has also been virtually elimina ted through improved control of corrosion product impurities in the reactor feedwater_ Insert  ;

Cladding hydriding is the result of excessive amounts of hydrogenous impurities (moisture and/or hydrogenous material) inadvertently introduced into the rod during the fuel fabrication process. The fuel fabrication process currently includes the following steps to minimize possible failures from this mechanism: 1) drying of components and pellets prior to rod loading, 2) hot vacuum outgassing of 7-all loaded fuel rods prior to the final end-plug weld, and

3) strict control of hydrogenous materials during fabrication. In addition, as noted in Section 4.2.1.2.1.12, every fuel rod contains supplementary protection in the form of a hydrogenous impurity getter which is placed in the plenum.

In early 1972, GE made design changes in the 7x7 fuel to reduce the incidence of pellet-cladding interaction (PCI) in future production. The improved 7x7 design incorporated a

, reduced pellet length-to-diameter ratio, chamfered pellet ends, and the elimination of pellet dishing to reduce the

.m. magnitude of pellet distortions contributing to local

'53 cladding strains. This design also employed an increased cladding heat treatment temperature to reduce the statistical variability in cladding mechanical properties.

Additional information regarding this cladding material is provided in Section 4 of Reference 4. These short-term design changes have been coupled with the longer term design effort which culminated in the 8x8 design, which was i

introduced into operating reactors in the spring of 1974.

With the 8x8 fuel, peak linear power is reduced by more than '

25 percent relative to the 7x7 fuel design to address the l strong dependence of PCI failures on bundle power. The favorable fuel performance of both early GE BWR fuel designs with low linear heat rates, and current 8x8 reload fuel provides assurance of the improved reliability of the 8x8 fuel design.

4.2.1.2.1.15 Design Basis for Fuel Assembly Surveillance t

GE maintains an active fuel assembly surveillance program specifically -intended to monitor performance in operating reactors to identify and characterize unexpected phenomena

which can influence fuel integrity and performance. Outage-oriented examinations are performed contingent on fuel availability as influenced by plant operation. Typically, peak duty fuel assemblies (with respect to exposure, LHGR, 4.2-15 t

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INSERT.(Page 4.2-15) for 5-CPB

...and condensate systems as explained'in Section 10.4.6.2. During GE surveillance of. irradiated fuel. rods,. visual inspections,'as-T: described in Section 4.2.4.3, will be performed.

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RBS FSAR L.w CTC kggg, for various cold, xenon-free conditions. For the purposes of this table middle-of-cycle (MOC) is defined as the most reactive point in the cycle. The reactivity and control fraction (CF) values for a variety of operating conditicns are listed in Table 4.3-3. The worth of various reactivity effects can be. estimated by taking the differences between reactivity states with all but one variable constant. Estimates of the temperature defect, the power defect, the xenon defect, and the excess reactivity can be inferred.

INSERT >

4.3.2.5 Control Rod Patterns and Reactivity Worths A detailed core simulation study for a typical BWR 6 is provided in Appendix 4A showing that the BWR core meets design performance cr.lteria. Typical BWR control rod positions are utilized in the study. The rod patterns described represent only one feasible sequence which results in power distributions well within design limits. Actual operating reactor rod patterns are based upon the measured distributions in the plant and together with the rod worth minimizing systems, limit the amount and rate of reactivity insertion in the event of a control rod drop accident in 2 such a way that the peak fuel pellet enthalpy is less than 9, the 280 cal /gm design limit. Rod worth minimizing systems

'[i also assure that the 170 cal /gm fuel enthalpy limit is not exceeded for any cold rod withdrawal error event.

For BWR plants, control rod patterns are not uniquely specified in advance; rather, during normal operation the control rod patterns are selected based on the measured core power distributions, within the constraints imposed by the systems indicated in the following sections. Typical control rod patterns are calculated during the design phase to ensure that all safety and performance criteria are satisfied. Control rod patterns and the associated power distributions for a typical BWR are presented in Appendix 4A. These control rod patterns are calculated with the BWR core simulator'3'. The ability of this model to predict centrol rod worth can be inferred from the detailed reactivity data presented in Reference 6. The comparisons of calculated and measured reactivity for the cold condition in both an in-sequence critical, where roughly 25 percent of the control rods are withdrawn, and the stuck rod measurement, where only one or two rods are withdrawn, show the ability of the model to predict rod worth. The data presented in Table 7 of Reference 6 show that no significant bias exists between these two configurations; therefore, it is concluded that the worth of the rods is accurately

( Amendment 2 4.3-15 February 1982

n a = ,

INSERT (Page 4.3-15) for 9-CPB Figure 4.3-24 is based on point model calculations. Nuclear parameters are volume weighted and power shape is not assuemd to change with void fraction and leakage effects. Integrating this curve over large void fraction changes is not representative of true change.in reactivity because the basic assumption of the point model basis is violated (e.g., power shape changes).

.,w ,

RBS ESAR ri-iy 15.4.9.2.3 Effect of Single Failures and Operator Errors Systems mitigating the consequences of this event are RCIS and APRM scram. The RCIS is designed as a redundant system network and therefore provides single failure protection.

The APRM scram system is designed to single failure criteria. Therefore, termination of this transient within the limiting results discussed below is assured.

No operator error (in addition to the one that initiates ,

this event) can result in a more limiting case since the RPS _

automatically terminates the transient.

Appendix 15A provides a detailed discussion on this subject.

15.4.9.3 Core and System Performance 15.4.9.3.1 Mathematical Model The analytical methods, assumptions, and conditions for evaluating the excursion aspects of the CRDA bro described in detail in References 4, 5, and 6. They are considered to provide a realistic yet conservative assessment of the

. associated consequences. The data presented in Reference 1

, _shows that the RCIS banked positicn mode reduces the control rod worths to the degree that the detailed analyses f(e,?

^ *.

presented in References 4, 5, and 6 or the bounding analyses _ INSERT presented in Reference 7 are not necessary.T Compliance checks are instead made to verify that the maximum rod worth does not exceed 1 percent ak.

If this criterion is not met, then the bounding analysis is performed. The rod worths are determined using the EWR .'

simulator model(2). Detailed evaluations, if necessary, are imade using the methods described in References 4, 5, and 6.

_ l-15.4.9.3.2 Input Parameters and Initial Conditions ,

-The core at the time of CRDA is assumed to be at the point in cycle which results in the highest incremental rod worth, to contain no xenon, to be in a hot-startup condition, and to have the control rods in sequence A at 50 percent rod

, density (groups 1-4 withdrawn). Removing xenon, which

! competes well for neutron absorptions, increases the l fractional absorptions, or worth, of the centrcl rods. The 50 percent control rod density (ablack and white" rod pattern) , which ncminally cccurs at the hot-startup condition, ensures that withdrawal of a rod results in the l maximum increment of reactivity.

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15.4-19 I

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INSERT-(Page 15.4-19) for 10-CPB References 1,4,5 and 6 provide sensitivity studies indicating large margins in peak enthalpy for rod worths below 1 percent Ak.. Since this margin is sufficiently large that changes in Doppler coefficients, scram curves, reactivity insertion shape, etc. will not significantly reduce this margin, no unique bounding analysis is needed.