ML20040A118

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Summary of 760806 Meeting W/B&W Re Vendor 760719 Responses to NRC 760514 Round 1 Questions.Attendance List & Related Draft & Markedup Info Encl
ML20040A118
Person / Time
Site: 05000561
Issue date: 08/03/1976
From: Cox T
Office of Nuclear Reactor Regulation
To:
Office of Nuclear Reactor Regulation
Shared Package
ML111090060 List: ... further results
References
FOIA-80-515, FOIA-80-555 760824, NUDOCS 8201200370
Download: ML20040A118 (21)
v • d • e


Text

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UNITED STATES c

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WASHINGTON, D. C. 20665 z

AUG 2 41976

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Docket No: STN 50-561 VEND 0R: Babcock & Wilcox Company (B&W)

SUMMARY

OF MEETING WITH NRC STAFF fr On Augustg,1976, members of the Nuclear Regulatory Commission (NRC) staff met with representatives of B&W to discuss B&W's July 19, 1976 responses to several of the staff Round 1 questions issued to B&W on May 14, 1976. Questions discussed included numbers 012.27, 30, 31, 32, 36 and 211.2. The staff requested the meeting to develop a more complete understanding of, and to discuss observed deficiencies in, the B&W responses. An attendance list is enclosed.

The discussion on request number 211.2 generally included a description by staff members of the evolution of the issue embodied in the request.

and a joint B&W/st'aff discussion of what should constitute an adequate response to the request. B&W concluded that while they felt they could provide at least some of what the staff was looking for in the immediate future, detailed commitments regarding their complete response would need more work and could not be scheduled at this time.

The staff stated that "RC pump auxiliaries that are an integral part of each RC pump assembly" is intended to include auxiliary components on both the pump and the driving motor, which together comprise the RC pump

" assembly." B&W stated that they would provide the clarification requested regarding Quality Group classification and Safety classifica-tion of auxiliary components and offer further justification of their contention that their designs are in conformance with Regulatory Guides 1.26 and 1.29.

B&W also stated that they will provide additional specific design criteria for RC pump (and notor) auxiliaries in the BSAR document to include structural design criteria, materials and quality class data.

B&W's July 19, 1976 response to request number 211.2 was generally intended to demonstrate to the staff that the RC pump will operate for at least 30 minutes after a loss of component cooling water without impairing the pump coastdown capability. Additional automatic pump trip signals (nonsafety grade) were proposed to be included in the bearing temperature alarm system.

Staff reviewers pointed out that the B&W response was deficient in that:

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(1) The response did not demonstrate that the component cooling

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water system design would be required to meet the requirements j

of condition 2 of 211.2 regarding single failure protection i

and instrumentation and control provisions (safety grade 1

detection and control room alann),

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(2) The response did not provide the requested analytical justi-l fication of successful RC pump operation for 30 minutes i

following loss of component cooling water, and (3) The response did not address the RC pump test that request number 211.2 stated would be required in the event that B&W elected to justify 30 minute pump operation.

The staff brought out during discussion that the full-scale, operating condition test of an RC pump assembly was required in part to verify the vendor's ability to analytically predict temperature rise in the complex system. Once confidence in the vendor's predictive analyses dnd pump performance at maximum temperature is achieved, it should not be necessary to test a given RC pump assembly over the entire range of operating temperatures, and it may not be necessary to proto-(.)

type test pump assemblies of similar design and construction but from y'

different manufacturers.

Discussion of request numbers 012.27, 29 and 32 led to a clarification of interface requirements for B&W designed equipment to be mounted by the balance-of-plant (B0P) designer on or in safety-related structures.

B&W was requested to provide preliminary design information in BSAR-205 describing static loads and the locations of these forces on the equip-ment items. Where applicable, electrical service and other B0P re-i quirements should also be specified in the interface data.

Regarding request number 012.30, B&W agreed to specify in BSAR-205, the numerical values of decay heat load resulting from the amounts of. spent fuel described in the request for information. The response to date j

is inadequate in that the heat load values were not given as requested, although a commitment was made to submit the data. directly to the B0P designer at an unspecified future time.

l Discussion of 012.31 resulted in a B&W commitment to provide, in a future BSAR-205 amendment, a specified minimum water depth over the i

spent fuel array, as requested by the staff.

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In a discussion of 012.'36 B&W pointed out that their July 19, 1976

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response did include tabular data, that described the input parameters used in, and uncertainty factors added to the equations of proposed j

standard ANS 5.1 to calculate the decay heat rates following reactor i

shutdown. They reiterated that their response did describe a method that gave results identical to the staff position APCSB BTP 9-2.

The staff reviewers stated that specific details, including the governing

. equations and assumptions used in B&W's method must be reviewed by the staff to verify that the B&W calculational method is equivalent to the method specified in the Acceptance Criteria of the steff's Standard l

Review Plan, Section 9.2.5.

B&W agreed to provide complete details in the BSAR-205 document, including a copy of the version of the proposed standard (ANS 5.1).that was used.

In general, B&W's intent is to document all required material in the BSAR-205 pages' no later than the date scheduled for B&W', responses I

to Round 2 positions, which is October 18, 1976.

Thomas H. Cox, Project Manager Light Water Reactors Branch No. 3 Division of Project Management 1

Enclosure:

Attendance List cc: Sabcock & Wilcox Company Mr. A. H. Monteith ATTN: Mr. Kenneth E. Suhrke Ohio Edison Company Manager, Licensing 47 North Main Street Nuclear Power Generation Akron, Ohio 44308 l

P. O. Box 1260 j

Lynchburg, Virginia 24505 Mr. W. E. Kessler Commonwealth Associates, Inc.

Washington Public Power Supply System ATTN: Mr. J. J. Stein 209 East Washington Managing Director Jackson, Michigan.49201 b0 G rge ashington Way Robert J. Kafin, Esq.

Richland, Washington 99352 en s

York 12801 Mr. Robert Borsum Bethesda Representative B. G. Shultz, Project Engineer Babcock & Wilcox Stone & Webster Engineering Corp.

Nuclear Power Generation Division P. O. Box 2325 Suite 5515, 7735 Old Georgetown Boston, Massachusetts 02107 Road Bethesda, Maryland 20014 ly

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AUG 2 4 576

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l ENCLOSURE 4

l ATTENDANCE LIST MEETING OF AUGUST 6, 1976 I

NRC AND B&W-1 NRC:

.B&W:

i T. H. Cox (DPM)

E. Swanson l

R. Kirkwood (P.S.)

G. Anderson L. Riani (APCSB)

J. R. Hamilton A. R. Ungaro (APCSB)

G. J. Brazill V. T. Leung (APCSB)

J. G. Newton D. L. Tibbitts (DPM)

J. J. Happell G. Mazetis (RSB)

0. D. Parr (DPM) 1 4

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NUCLEAR REGULATORY COMMISMON g,k u

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Docket No: STN 50-561 g

p VENDOR: Babcock & Wilcox Company (B&W)

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SUMMARY

OF MEETING WITH NRC STAFF lP*19 On August 9, 1976, members of the Nuclear Regulatory Commission (NRC)

A staff met with representatives of B&W to discuss B&W's July 19, 1976 responses to several of the staff Round I questions issued to B&W on May 14, 1976. Questions discussed included numbers 012.27, 30, 31, 32, 36 and 211.2. The staff requested the meeting to develop a more ccmolete understanding of, and to discuss observed deficiencies in, the ELW responses. An attendance list is enclosed.

The discussion on request number 211.2 generally included a description by staff members of the evolution of the issue embodied in the request, and a joint B&W/ staff discussion of what should constitute an acequate response to the request. B&W concluded that while they felt they could l

provid? at least some of what the staff was looking for in the immediate l

future, detailed commitments regarding their complete response would need more work and could not be scheduled at this time.

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The staff stated that "RC pump auxiliaries that are an integral part of each RC pump assembly" is intended to include auxiliary components on both the pump and the driving motor, which together' comprise the RC pump

" assembly." B&W stated that they would provide the clarification requested regarding Quality Group classification and Safety classifica-tion of auxiliary components and offer further justification of their contention that their designs are in conformance with Regulatory Guides 1.26 and 1.29.

B&W also stated that they will provide additional specific design criteria for RC pump (and motor) auxiliaries in the BSAR document to include structural design criteria, materials and quality class data.

l B&W's July 19, 1976 response to request number 211.2 was generally intended to demonstrate to the staff that the RC pump will operate for at least 30 minutes after a loss of component cooling water without impairing the pump coastdown capability. Additional automatic pump trip signals (nonsafety grade) were proposed to be included in the bearing temperature alarm system. Staff reviewers gointed.out that the B&W response was deficient in that:

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j l (1) The response did not demonstrate that the component cooling water system design would be required to meet the requirements of condition 2 of 211.2 regarding single failure protection and instrumentation and control provisions (safety grade detection and control room alarm),.ame I

(2) The response did not provide the requested analytical justi-fication of successful RC pump operation for 30 minutes j

following loss of component cooling water, and (3) The response did not address the RC pump test that request i

number 211.2 stated would be required in the event that B&W elected to justify 30 minute pump operation.

The staff brought out during discussion that the full-scale, operating condition test of an RC pump assembly was required in part to verify the vendor's ability to analytically predict temperature rise in the complex system. Onr.e confidence in the vendor's predictive analyses i

and pump performance at maximum temperature is achieved, it should not be necessary to test a given RC pump assembly over the entire j

range of operating temperatures, and it may not be necessary to proto-s type test pump assemblies of similar design and construction but from different manufacturers. pp.oVIDED T8t. DESWO %RAMET=E tyt.E sufFlc(EOYt/y CoMSEE7ATNe 1 o erueodpASS A u g%

Discussion of request numbers 012.27, 29 and 32 led to a clarification 1

of interface requirements for B&W designed equipment to be mounted by the balance-of-plant (B0P) designer on or in safety-related structures.

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B&W was requested to provide preliminary design information in BSAR-205 describing static loads and the locations of these forces on the equip-A ment items. Where applicable, electrical service and other B0P re-O N

quirements should also be specified in the interface data.

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i Regarding request number 012.30, B&W agreed to specify in BSAR-205, the

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numerical values of decay heat load resulting from the amounts of spent fuel described in the request for information. The response to date j

is inadequate in that the heat load values were not given as requested, although a commitment was made to submit the data directly to the B0P designer at an unspecified future time.

Discussion of 012.31 resulted in a B&W commitment to provide, in a future BSAR-205 amendment, a specified minimum water depth over the spent fuel array, as requested by the staff.

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  • 3-In a discussion of 012.36, B&W pointed out that their July 19, 1976 response did include tabular data that described the input parameters used in, and uncertainty factors added to the equations of proposed standard ANS 5.1 to calculate the decay heat rates following reactor shutdown. They reiterated that their response did describe a method the staff that gave results identical top -- iewers stated that specific i

position APCSB BTP 9-2.

The staff rev details, including the governing equations and assumptions used in B&W's method must be reviewed by the staff to verify that the B&W cal-culational method is equivalent to the method specified in the Accep-tance Criteria of the staff's Standard view Plan, Section 9.2.5.

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B&W agreed to provide complete details in the BSAR-205 document, in-cluding a copy of the version of the proposed standard (ANS 5.1) that was used.

In general, B&W's intent is to document all required material in the BSAR-205 pages no later than the date scheduled for B&W's responses to Round 2 positions, which is October 18, 1976.

Thomas H. Cox, Project Manager l

Light Water Reactors Branch No. 3 Division of Project Management 1

Enclosure:

Attendance List cc: Babcock & Wilcox Company Mr. A. H. Monteith ATTN: Mr. Kenneth E. Suhrke Ohic Edison Company Manager, Licensing 47 North Main Street Nuclear Power Generation Akron, Ohio 44308 i

P. O. Box 1260 Lynchburg, Virginia 24505 Mr. W. E. Kessler Commonwealth Associates, Inc.

Washington Public Power Supply System i

ATTN: Mr. J. J. Stein 209 East Wasnington Managing Director Jackson, Michigan 49201 Robert J. Kafin, Esq.

3b0 George ashington Way Richland, Washington 99352 len s te York 12801 Mr. Robert Borsum Bethesda Representative B. G. Shultz, Project Engineer i

Babcock & Wilcox Stone & Webster Engineering Corp.

Nuclear Power Generation Division P. O. Box 2325 i

Suite 5515, 7735 Old Georgetown Boston, Massachusetts 02107 Road

-O Bethesda, Maryland 20014

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4 ENCLCSURE ATTENDANCE LIST MEETING 0F AUGUST 6, 1976 NRC AND B&W i

NRC:

B&W:

T. H. Cox (DPM)

E.fsk Swanson R. Kirkwood (P.S.)

es, % Anderson L. Riani (APCSB)

J. R. Hamilton A. R. Ungaro (APCSB)

G. J. Brazill V. T. Leung (APCSB)

J. G. Newton x 0. L. Tibbitts (SPM)

J. J. Happell-

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4 Jewey Mazetis (RSB)

0. D. Parr (DPM) 1 x.-

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6/28/76 B-SAR-205 Amend. 2

(

211.2 (Section 3.2)

In Tables 3.2-4 and 5.1-2 and in Figure 5.5-8, your

(( )

Quality Group D classification and Non-Seismic Category I classification of the RC pump auxiliaries that are an integral part of each RC pump assembly is not in ionformance with Regulatory Guides 1.26 and 1.29 and is unacceptable.

In Table 3.2-4 and in section 5.5.1 of the B-SAR provide a detailed list and description of those components that are generally identified in Table 3.2-4 as RC pump auxiliaries.

It is our position that those components that are an integral part of the RC pump, such as the oil lubrication system, oil coolers, seal coolers and cooling water lines to the interface point with the com-ponent cooling water (CCW) system, should be classified Quality Group C, constructed to ASME Section III, Class 3, and designed to Seismic Category I requirements.

It is also our position th t the following criteria shall apply to the a

portion of the CCW system that interfaces with the RC pumps to sapply cooling water to pump seals and bearings during normal operation, anti-cipated transients, and the following accidents:

acdtit.

1.

A moderate energy leakage crack or a single failure in the component 3

cooling water system shall not result __in_ fuel damag tor damage to th.e RCPB caused by an extended loss of cooling to the RC pumps. Single failures include operator error, spurious actuation of motor-operated

" elves, and loss of CCW pumps. Moderate leakage cracks should be determined in accordance with the guidelines of Branch Technical

\\

Position APCSB 3-1.

2.

An acciden s initiated from a in the CCW system piping shall not result in eJcessive fuel damage or a breach of the RCPB when an extended loss of cooling to the RC pumps occurs. A single active failure shall be considered when evaluating the consequences of this accident. Moderate leakage cracks should be determined in accordance with Branch Technical Position APCSB 3-1.

In order to meet the criteria established above, a B-SAR-205 interface requirement should be imposed on the balance of plant CCW syscem that provides cooling water to the RC pump seals and bearings, so that the system will meet the following conditions:

1.

A period of 20 minutes is considered acceptable, within which an operator can trip the RC pumps and initiate a safe plant shutdown.

Therefore, in order to provide an adequate margin of safety, unless it'can be demonstrated that the RC pumps are designed so that they can operate without cooling water for a minimum period of 30 minutes without loss of function or the need for reactor operator action, the CCW system should be designed to meet the following:

a.

Seismic classification:

Seismic Category I

(

b.

Quality classification:

Quality Group C Component code:

ASME Section III, Class 3 c.

Single failure:

Should be capable of withstanding a single active failure and should be capable of withstanding a moderate energy leakage crack in accordance with Branch Technical diti 4..

s-+

A

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e 9

1 l

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)

6/28/76 B-SAR-205 Amend. 2 l

I r

1

('

Position APCSB 3-1, with respect to cooling RC pumps or else, item I

Id must be implemented. A single failure includes malfunctioning of any valve or pump in the CCW lines to the RC pumps.

d.

Instrumentation and controls:

Safety grade instrumentation is required to detect loss of component cooling water and initiate automatic protection of the plant.

I e.

Containment isolation of systems: Only when RC pumps are not functioning.

j i

2.

For nn RC pump design that can operate without cooling water for j

longer than 30 minutes without loss of function or the need for.re-actor operator action, the CCW system should be designed to meet the

[

following:

t a.

Seismic classification:

Non-Seismic Category I except for the l

portion of the CCW system that forms an extension of the con-tainment boundary.

b.

Quality classification:

Quality Group D except for the portion I

of the CCW system that forms an extension of the containment boundary.

c.

Single failure: The system should be capable of withstanding a single active failure and must be capable of withstanding a mod-i erate energy leakage crack in accordance with Branch Technical Position APCSB 3-1, with respect,to cooling RC pumps or else, l

item 2d must be implemented. A single failure includes malfunc -

i tioning of any valves or pumps in the CCW lines to the RC pumps.

d.

Instrumentation and controls:

Safety grade instrumentation is l

required to detect the loss of component cooling water and pro-vide an alarm in the control room to satisfy item 2c.

j e.

Containment isolation of systems: Only when RC pumps are not l

functioning.

i i

The RC pumps are within the B-SAR-205 scope of supply; therefore, in order to demonstrate that an RC pump design can operate with loss of component cooling water for longer than 30 minutes without-loss of fune-3 tion or the need for reactor operator action, provide the following:

i 1.

A detailed description of the events following the loss of component cooling water to the RC pumps and an analysis demonstrating that no consequences important to safety may result from this event.

Include-a discussion of the effect that the loss of cooling water to the seal-l coolers has on the RC pump seals.

Shcm that the loss of cooling water does not result in a LOCA due to seal failure.

i 2.

A detailed analysis to show that Joss of cooling water to che RC pumps and motors will not cause a loss of the flow coastdown characteristic

[ _

or cause seizure of the pumps aspuming no administrative action is taken. The response should include a' detailed description of the

(

)

calculation procedura including:

v

. a.

The equations used.

t b.

The parameters used in the equations, such no the design.

g-

- v._

l

^

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e aw*

l k

6/28/76 B-SAR-205 Amend. 2 parameters for the motor bearings, motor, pump and any other equ.ipment entering into the calculation, and material property values for the oil and metal parts.

c.

A discussion of the effects of possible variations in part di-mensions and material properties, such as bearing clearance tolerances and misalignment.

d.

A description of the cooling and lubricating systems (with appro-priate figures) associated with the RC pump and motor and its design criteria and standards.

e.

Information to verify the applicability of the equations and j

material properties chosen for the analysis (i.

e., references j

should be listed, and if empirical relation are used, provide l

a comparison of their range of application to the range used in l

gp) the analysis).

p Should an analysis be provided to demonstrate that loss of component cooling water to the RC pumps and motor assembly is acceptable, we will require certain modifications to the plant Technical Specifications.a u j

/ /

M dn RC pump test conducted under operating ccnditions and with component cooling water terminated for a specified period of time to verify the W a A O as % +ka da,ep analysis. % de.d sh*M 64.'

  • N-BS AR.-tof d^c

^#5

Response

l B&W RC pumps and motors are adequately protected to prevent a seizure, which M

could result from a loss of component cooling water. The RC pump is protected by the cooling provided by seal injection water from a safety grade system, which is seismically qualified and not subject to a single active failure.

The system includes a standby pump which is automatically actuated on low seal injection flow (see the response to question 212.147).

a the component cooling water to the seal coolers serves only as a secondary back-up system for added protection of the seals.

e l gad y The seal coolers are Safety Class 1 on the primary side and Safety Class 3 on the secondary side, Seismic category I, and meet the requirements of B&W's QA program described in topical repor,t BAW-10096A, which is in compliance with 10 CFR 50, Appendix B.

The motor oil coolers are designed to TDIA standards anyd factured to standardsandproceduresequivalenttoAQECode-requiremen)t. These components I

are seismically analyzed to remain functional through the seismic event and sub-(

)

sequent coastdown of the pump and motor. Motor vendors are qualif %d to the B&W

' QA program described in topical report BAW-10096A, which complies wu.h 10 CFR 50,

[

Appendix B.

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6/28/76 B-SAR-205

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Amend. 2 b

s

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The motors are protected by high-temperature alarms at each of the thrust and radial bearings.

In addition, the reactor diagnostic system will provide an alarm on high unit vibration. The motors are designed to rus with normal bearing temperatures of approximately 170F with the maximum temperature com-

- ponent cooling water. The alarm point for bearing temperature and operator manual shutdown of the motor is conservatively set at 185-190F.

Based on cal-calations from the motor manufacturers, th'e average temperature rise of the bo.arings is 5F per minute withcot component cooling water, which in turn would s

give 3L4 minutes from the loss of cooling water alarm signal to the high bear-ing temperature alarm signal. Motors of this type have been operated with bearing temperatures as high as 360F without substantial damage, which caused nc loss of coastdown capability.

In fact, the bearings were inspected and only required minor hand scraping to be reused in the same motor.

Based on this ex-perience and allowing the bearings to reach 360F', the operator would have an additional 34 minutes to react to the bearing high temperature alarm or a total of 37 minutes to reinitiate component cooling water flow or perform an orderly

's shutdown of the pumps without substantial damage to the bearings. This time element exceeds the 30-minute minimum requirement per the question, although we do believe the 30 minutes time allowable to be arbitrary and unsupported by published regulations or human engineering data.

In addition we propose that the following automatic pump trip signals be added or incorporated as part of 3

the bearing temperature alarms.

Since each bearing'has two> temperature ele-

~

ments one active and one spare - we propose to utiliz'e a two-out-of-thr.e i

-~

ure signal from the upper bearing assembly a'nd a ne-out-of-two e

signal from the lower bearing assembly to trip the pump motor. This redund-

,,.Q

['ancyinsignalswouldprotecttheplantfromanydowntimeforbearingmainte-nance after a loss of component cooling water.

Since B&W recommends that the essential CCW supply be provided by two loops, only two RC pumps would be affected by a single failure in any loop. Although B&W requires in our equip-ment specification that the pump and motor be seismically analyzed to provide

/

coastdowncapabilityaftertheseismiceventjwedonotrecommendcontinued l

operation until af ter an inspection has been/ performed to determine that no

/

damage has occurred.

f

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2&

b i e

W M' e'

a

.....c.

. ~.

N M MAf R.t9/. t y

k 15.X.X Event Evaluation 1.

Identification of causes and frequency classification. For each event evaluated, include a description of the occurrences that laad to the initiating event under consideration. The probability of the initiating

'p event should be estimated and the initiating event should be assigned to de

& pg.,g(St#dM one of the following groups:

t' a.

Incidents of moderate frequency - these are incidents, any one of which may occur during a calendar year for a particular plant.

  • ,u

/ $o e

b.

Infrequent incidents - these are incidents, any one of wh ch M

may occur during the lifetime of a particular plant.

h c.

Limiting faults - these are occurrences that are not expected to occur but are postulated because'their consequences would include the potential for the release of significant amounts of radioactive material.

2.

Sequence of events and systems operation. The following should be discussed for each initiating event:

a.

The step-by-step sequence of events from event initiation to the final stabilized condition. This listing should identify each signi-ficant occurrence on a time scale, e.g.,

flux moniter trip, insertion of m

[

contrc,1 rods begin, primary coolant pressure reaches safety valve set l

g point, safety valves open, safety valves close, containment isolation signal i

initiated, and containment isolated. All require'd operator actions should also be identified.

1 b.

The extent to which normally operating plant instrumentation and controls are assumed to function.

c.

The extent to which plant and reactor protection systems are required to function.

d.

The credit taken for the functioning of norma'lly operating plant systems, e.

The operation of engineered safety systems that is required.

The effect of single failures in each of the above areas and the effect of operator errors should be discussed and evaluated. The discussion should provide enough detail to permit an independent evaluation of the adequacy of the system as related to the event under study. One method of system-atically investigating single _ failures is the use of a plant operational t

analysis or a failure mode and effects analysis. The results of these types of analyses can be used to determine which functions, systems, inter-locks, and controls.are safety related and what readouts are required by the operator under anticipated operational occurrence and accident

')

conditions.

$4 15-5

- _ _ _ _ - _ _

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