ML20111C496
| ML20111C496 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 03/12/1985 |
| From: | Woolever E DUQUESNE LIGHT CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| 2NRC-5-042, 2NRC-5-42, NUDOCS 8503150421 | |
| Download: ML20111C496 (29) | |
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Nuclear Construction Division Telecopy (412) 7sT-2629 Robinson Plera, Building 2. Suite 210 Pittsburgh, PA 15205 March 12, 1985 United States Nuclear Regulatory Cosunission Washington, DC 20555 ATTENTION:
Mr. Harold R. Denton Of fice of Nuclear Reactor Regulation
SUBJECT:
Beaver Valley Power Station - Unit No. 2 Docket No. 50-412 Elimination of Arbitrary Intermediate Pipe Breaks Centlemen:
On February 5,1985, Duquesne Light Company (DLC), along with other utilities, met with Westinghouse and the NRC to discuss future activity on Arbitrary Intermediate Pipe Breaks. At that meeting the NRC encouraged formal submittal on this issue for NRC approval.
Enclosed for NRC staf f review are the alt ernative pipe break criteria wh ich we propose to apply to Beave r Valley Power Station Unit No. 2 (BVPS-2), which would obviate the need to postulate arbitrary intermediate pipe breaks.
Arbitrary intermediate pipe breaks are those break locationn, which based on piping st res s analysis results are below the stress and fatigue limits specif led in Branch Technical Position (BTP) MEB 3-1, but which are arbitrarily selected as the two highest stress locations between the tenninal e nd s of a piping system as required by the BTP.
It has become apparent to both the NRC staf f and the nuclear industry that this particular criterion requiring the postulation of arbitrary intermediate pipe breaks can be overly j
res tric t ive and result in an exces sive number of pipe rupture protection devices which do not provide a compensating increase in the level of safety to the public.
It is for this reason, as further explained and justified in detail in the enclosure to this letter, that DLC is pursuing the application of alt e rnat ive pipe break criteria in the design of BVPS-2.
Attachment A provides a summary of the benefits derived from elimina-tion of the arbitrary breaks. Attachment B provides the technical justifica-tion for the employment of the alternative pipe break criteria. Attachment C provides a list by piping system of the ASME Class 1, 2, and 3, and NNS piping intermediate break locations which are candida tes to be eliminated.
Attachment D gives systems information for these same piping systems. Attach-ments E, F, C, H, and I provide detailed descriptions of our provisions for minimizing stress corrosion c racking in high ene rgy lines, minimizing the ef fects of thermal and vibration induced piping fatigue, minimizing steam and water hammer ef fects, environmental effects, and minimizing local stresses from welded attachments.
The pe rcent age of the total potent ial benefits that can be realized by DLC for BVPS-2 becomes a matter of timing due to the advanced stage of
Unit;;d Stcts3 Nuclocr R:gulctory Commission Mr. H;rold R. Denton Elimination of Arbitrary Intennediate Pipe Breaks Page 2 design and construction.
To make it pos sible for DLC to realize the maximum benefits af forded by this proposed change in the pipe break criteria, immedi-ate attention by the NRC is requested with a f avo rable response to the proposed change in the pipe break criteria by April 30, 1985.
Upon your concurrence, the FSAR will be appropriately revised in a future amendment.
DUQUESNE LIGHT COMPANY I
I By
'E yJ. Wooleve r Vice President RC/wjs A t t achme nt cc:
Mr. B. K. Singh, Project Manager (w/a)
Mr. G. Walton, NRC Resident Inspector (w/a)
COMMONWEALTil 0F PENNSYLVANIA
)
)
SS:
COUNTY OF ALLEGilENY
)
On this
/44 day of M t ?,"
/// I, before me, a Notary Public in and for said Commonwealth and County, personally appeared E. J. Woolever, who being duly sworn, deposed and said that (1) he is Vice Pres ident of Duquesne Light, (2) he is duly authorized to execute and file the foregoing Submittal on behalf of said Company, and (3) the statements set forth in the Submittal are true and correct to the be s t of his knowledge.
/Nr i
Notary Public ANITA ELAINE REITER, NOTARY PUGUC ROBINSON TOWNSHIP, ALLEGHENY COUNTY MY COMMfSSION EXPIRES OCTO,GEq 20,1986
ATTACHMENT A DUQUESNE LIGHT COMPANY BEAVER VALLEY POWER STATION UNIT 2 l
BENEFITS FOR ELIMINATION OF l
ARBITRARY INTERMEDIATE PIPE BREAKS Duquense Light Company (DLC) has followed closely the recent activities of the Nuclear Regulatory Commis sion (NRC) staff and the nuc le ar industry related to the treatment of design basis pipe breaks in high energy piping systems.
In particular, it is noted that the NRC staff has expressed an interest in the industry's proposal to modify the current pipe break criteria to eliminate from design cons iderat ion those intermediate breaks generally referred to as arbitrary intermediate breaks (i.e.,
those break locations wh ic h, based on stress analysis, are below the stress limits and/o r the cumulative usage factors speci f ied in the current NRC criteria) but are selected to provide a minimum of two breaks between terminal ends. Elimina-tion of the arbitrary intermediate breaks of fers considerable benefits due to the deletion of the associated pipe whip restraints and other provisions currently incorporated in plant designs to mitigate the ef fect s of such breaks.
The break selection criteria currently employed by DLC for Beaver Valley j
Power Station Unit 2 (BVPS-2) is taken from NRC Branch Technical Positions ASB 3-1 and MEB 3-1.
These documents require that pipe breaks be considered i
at terminal ends and at intermediate locations where stresses or cumulative usage factors exceed specified limits.
If two intermediate locations cannot be de termined based on the above (i.e., stresses and cumulative usage factors are below specified limits), then the two highest stress locat ions are selected.
DLC concurs with the nuclear industry in the belief that current knowledge and experience suppo rt s the conclusion that designing fo r the arbitrary intermediate breaks is not justified and that this requirement should be deleted.
This conclusion is supported by extensive operating experience in over 80 operat ing U.S.
plants and a number of similar plants overseas in which no piping failures have been known to occur that would suggest the need to design protective features to mitigate the dynamic effects of arbitrary inte rmediate breaks.
Arbitrary intermediate breaks are of ten postulated at locations where stresses are well below the ASME Code allowables and within a few percent of the stress levels at other points in the sane system.
This results in compilcated protective features being provided for specific break locations in the piping system that provide little to enhance overall plant safety.
In pr ac t ic e, cons ide rat ion of these two arbitrary int e rmediate breaks is particularly dif ficult because the location of the high stress po int s may move several times as the seismic design and analysis of structures and piping develops. The industry recognizes that the revised HEB 3-1, which was included in the July 1981 revisions to the Standard Review Plan (NUREC-0800),
provides criteria for not having to relocat e intermediate break points when i
highest stress locations shift as a result of piping reanalysis. As a prac-6 tical matter, however, these criteria provide little relief, since the burden 1
is on the designer to prove that not postulat ing breaks at relocated highest stress points does not degrade safety. This may require extensive additional analysis of break / target interactions for the relocated break point s and could result in design, fabrication and installation of additional pipe whip restraints at the relocated break points, and elimination of previou sly -
installed restraints at abandoned break points. Early determination of exact l
break locations is quite import ant because of all the secondary ef fects of l
the pipe break to be considered.
The benefits to be realized from the elimination of the arbitrary intermedi-ate break locations center primarily around the elimination of the associated pipe whip restraints and other structural provisions to mitigate the conse-quences of these breaks.
While a substantial reduction in capital costs for these res tra int s and structures can be realized immediately, there are also significant operational benefits to be realized over the 40 year life of each plant. As ident ified in NUREG CR-2136, these ef fects are particularly in the areas of overall plant reliability and exposure of plant personnel to radia-tion when excessive pipe whip restraints are installed.
Access during plant operat ion for such activities as maintenance and inser-vice ins pe ct ion is improved due to the elimination of congestion created by j
these restraints and the support ing structural steel, and in some cases due to the need to remove some restraints to gain access to welds.
In addition I
to the decrease in maintenance effort, a significant reduc t ion in man-rem l
exposure can be realized through fewer manhours spent in rad iation areas.
Also, the need to verif y appropriate cold and hot clearances between pipes and restraints during initial heatup, which requires additional hold points during the startup phase, can be dispensed.
Recovery from unusual plant conditions would also be improved by elimination of this congestion.
In the eve nt of a rad ioac t ive release or spill ins ide the plant, decontamination operations would be much more ef fect ive if the complex shapes, represented by the structural frameworks support ing the restraints, were eliminated.
This results in decreasing man-rem exposures associated with decontamination and restoration activities. Similarly, access for control of fires within these areas of the plant would be improved, especially under low visibility conditions.
Substantial overall benefits in these areas would be realized by reduc ing the number of whip restraints required.
By design, whip restraints fit closely around the high energy piping with gaps typically being on the orde r of half an inch.
These restraints and their supporting steel increase the heat loss to the surrounding environment significantly.
Also, because thermal axivement of the piping system during a t artup and shutdown could deform the piping insulation against the fixed wh ip restraint, the insulation must be cut back in these areas, creating convection gaps adjacent to the restraint, which also increases heat loss to the environment. This is a major contributor to the tendency of many contain-monts to operate at temperatures near technical specification limi t s.
The elimination of whip restraints associated with arbitrary intermediate breaks would assist in controlling the normal environmental temperatures and improv-ing system operational ef ficiency.
2 l
For the above reasons, DLC requests NRC approval of the fo llowi ng for the application of alternative pipe break criteria which would eliminate the need to pos tulate arbitrary intermediate pipe breaks (i.e., those break locations which, based on stress analysis, are below the stress limits and the cumula-tive usage f actors specified in the current NRC criteria, but are selected to provide a minimum of two breaks between terminal ends):
ASME Section 111 Piping Inside Containment Piping systems shall be designed to accommodate pipe breaks at terminal ends and locations where the stress or usage factor criteria of MEB 3-1 are exceeded. No arbitrary intermediate breaks will be postulated when the stress and/or usage factor criterion are not exceeded.
For breaks that must be taken, the design will accommodate pipe whip, jet impingement, and compartment pressurization result ing from mecha-nistic treatment of the break.
Current acceptable methods for limiting break opening, moderate and low energy exclusions, limited duration operation, etc. may still be applied.
For flooding evaluat ion, environmental qualification of equipment, and structural design of areas traversed by high energy piping systems, breaks will continue to be postulated in accordance with the present project criteria (i.e., in each area traversed by the high energy piping system, non-mechan is t ic bre aks are pos t ulated at the location that results in the most severe environmental consequences). Therefore, elimination of the arbitrary intermediate breaks will not impact the flooding evaluation, environmental qualification program, or plant structural design.
ASME Section III and Seismically Designed Non-ASME Section III Piping Outside Containment Piping systems shall be designed to accommodate pipe breaks at terminal ends and locat ions where the stress criteria of MEB 3-1 are exceeded.
No arbitrary intermediate breaks will be po s tulat ed when the stress criterion are not exceeded.
1 For breaks that must be taken, the design will accommodate pipe whip and jet impingement ef fects resulting from mechanistic treatment of the b re ak.
Compartment pressurization and flooding effects from break s postulated in accordance with MEB 3-1 will be accommodated in the design. Current acceptable methods for limiting break opening, moderate and low energy exclusions, limited duration operation, etc. may still be applied.
For environment al quali ficat ion of equipment and structural design of areas traversed by high energy piping systems, breaks will continue to be postulated in accordance with the present proj ect criteria (i.e., in traversed by the high energy piping system, non-mechanistic e ach area breaks are postulated at the location that results in the nost severe environment al consequences).
Therefore, elimination of the arbitrary intermediate breaks will not impact the environmental qualification program or plant structural design.
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Application of the alternative pipe break criteria described above will not alter the commitment to quality -in the design of safety related structures, systems, and component s.
The quality ass ur ance program will continue to e ns ure safety related structures, systems, and component s are de s igned,
f abricated, erected, and tested to the quality standards commensurate with the safety function to be performed.
The piping systems being considered, including an estimate of the number of i
arbitrary intermediate break s eliminated and a number of pipe rupture restraints and shields deleted, are detailed in Attachment C.
Attachment D gives system information fo r these same piping systems.
A total of 245 breaks are to be eliminated.
In this submittal we are providing additional technical information to justify further that req ue s t.
Specific NRC concerns are addres sed in the attachments as follows:
t l
- 1. Technical justification for elimination of i
arbitrary intermediate breaks Attachment B
- 2. Provisions for minimizing stress corrosion cracking in high energy lines Attachment E F
- 3. Provisions for minimizing the effects of. thermal and vibration induced piping f atigue Attachment F
- 4. Provisions for minimizing water / steam hanener ef fects Attachment G i'
- 5. Environmental analysis Attachment H
- 6. Provisions for minimizing local stresses from welded attachments Attachment I The application of the proposed criteria changes will result in the deletion of approximately 245 break locations, 105 pipe wh ip restraints, and ' 56 jet impingement shields in Classes 1,
2, and 3, and NNS piping.
The breaks, restraints, and shields currently planned for elimination are listed in A t tachent C.
However, it should be noted' that piping and system design is an iterat ive process and that postulated break locations could potentially move as the system design and analysis of structures and piping develops over the course of the design process and its potential for af fecting postulated break locations, changes af fecting high energy systems are continuously monitored and evaluated to determine the impact on break location. We propose to apply l
these alternative criteria to any potential break locations in ' the systems 3
identified herein, provided the stresses at those locations are below the break selection threshold, and the operational concerns in Attachment s F j
through I'are adequately addressed.
Also, for those piping systems, or portions thereof, which are not included in this submittal, the existing guidelines in MEB 3-1 of the SRP (NUREG-0800)
Revision I will be met.
If other piping subsystems included in the systems 1
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4 l
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,_.______,._.--m
,__..__r
ident ified in Attachment D, but not specifically ident ified in this submit-tal, subsequently qualify for the conditions described herein, the impleme n-tation of. the proposed elimination of the arbitrary intermediate break criteria may be used. If this criteria is to be applied to additional systems not included in A ttachment D, those systems will be appropriately identified to the' staff.
DLC has evaluated the potent ial cost savings and operational benefits that result from the elimination of arbitrary intermediate breaks. These benefits include $5-6 million savings in analysis, design, f abricat ion, and installa-tion of associated pipe. wh ip restraints, jet impingement barriers, and man-rem in dose reductions for BVPS-2 over its 40 year plant life. A summary of the benefits realized by the elimination of the arbi trary intermediate breaks is provided in Attachment A-1.
The actual benefits that DLC will realize are expected to be higher than these due to the hidden factors and intangibles that are di f ficult to identify at this time.
It. is clear, however, that elimination of the arbitrary intermediate breaks is both safety effective and cost effective.
4 1
4 5
ATTACHMENT A-1 SUP9tARY OF BENEFITS FROM THE ELIMINATION OF ARBITRARY INTERMEDIATE PIPE BREAKS -
BEAVER VALLEY POWER STATION UNIT 2 Effect of Break Elimination Cost Savinas Operational Benefits Elimination of 162 Pipe Rupture Restraints. Design, Fabrication and Installation Costs Potential improvement in performance of inservice (PRR) and Jet Shields inspection (ISI)
Dose Reduction Costs
. Dose reduction from improved personnel access during maintenance, ISI and recovery from unusual plant conditions, e.g., radioactive spills, fires, etc.
Improved capability to recover from unusual plant condition, e.g., decontamination following radio-active spills, access for fire fighting, etc.
Reduced system heat loss resulting from improved insulation design.
Dose reduction and improved construction schedule by eliminating the need to set and maintain PRR clearance gaps.
Elimination of Equipment Relocation
. Relocation Cost Improved system layout and design for future plant modifications.
Elimination of Evaluations Associated Jet Impingement Load and Pipe Whip Analyses with the Dynamic Effects Costs TOTAL SAVINGS
$5-6 Million 277 man-rem in dose reduction over the
>=
40 year plant life.
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r1 EC5z H
I pa 1
B4-12241-7668A i
Attachment B TECHNICAL JUSTIFICATION FOR ELIMINATION OF ARBITRARY INTERMEDIATE BREAKS I
The following items provide generic technical justification for the elimination of arbitrary intermediate pipe breaks and the associated pipe whip restraints for all piping other than Primary Loop.
1.
The operating procedures and piping and sys tem designs minimize the possibility of stress co: ros ion cracking, thermal and vibration induced fatigue, and water / s t ema hamme r in these lines in which arbitrary pipe breaks are currently postulated.
Detailed de scriptions of the design provisions for these phenome na are provided in Attachments E, F,
and G, respectively.
2.
Welded at t achment s are analyzed for stress levels at every location to account fo r their contribution.to the total stress.
A survey of welded I
a t t achme nt locations indicates that very few are in close proximity to arbitrary intermed i ate breaks.
For a further de sc rip t ion, refer to Attachment I.
3.
The pipe breaks and whip restraints being retained, i.e., terminal ends and intermediate points that exceed threshold limits, provide an adequate i
level of protection in areas containing high energy lines.
i
)
4.
Pipe breaks are postulated to occur at locations where stresses are 80 percent of Code allowables (Class 2 and 3) or where the cumulative l
usage factor is' 10 percent of the allowable (Class 1).
The arbitrary breaks to be eliminated all exhibit stresses and usage factors below these l
conservatively low thresholds.
5.
Pipe rupture is recognized in branch technical position MEB 3-1 as being a
" rare event which may only occur under unant icipated conditions."
The systems have been designed to preclude many operating transients ide nt i-f ied in NUREG-0582.
Those transients that remain have been analyzed and found ac ceptab le.
Consequently, the nuniber of unanticipated transients has been significantly lowered.
6.
Arbitrary intermediate breaks are only pos tulated to provide addi tional conservatism in the design.
There is no technical basis for postulating these breaks.
(See NUREG 1061 Volume 3) 7.
Elimination of pipe whip restraints associated with the arbitrary breaks l
will facilitate in-service inspection, reduce heat losses from the piping, and eliminate the po t ent ial inadvert ent restraints of : piping during thermal growth and seismic motion.
8.
Equipment environmental ^ qualification (EQ). req uirement s will not be affected by the elimination of the arbitrary breaks.
Breaks are po stu-l lated nonsechanistically for EQ purposes.
For ' a.further descrip tion, refer to Attachment H.
It is concluded that the elimination of. arbitrary-intermediate breaks is tech-nically justified, based on the reasons stated above..
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Page 1 ef +
ATTACHMENT C F05TULATED ARBITRARY INTERMEDIATE BREAKS (AIBs) l
- BREAKS / PROTECTION DEVICES ELIMINATED -
l i
Estimated Number Deleted P_iping' System ****
1.ocation*
NPS Safety Class AIBs RESTRAINTS SHIELDS i
Auxiliary Steam System OC 10 HTS OC 8
NNS OC 6
NNS OC 4
NNS 43 9
8 OC 3
NNS OC 2
NNS OC 1\\
NNS i
Steam Generator Blowdown IC 3
2 IC 25 2
6 2
2 IC 2
2 o
OC 3
2 3
3 2
OC 2
2 OC 3
NNS OC 15 NNS 11 5
4 Chemical and Volume Contrcl
- a. Charging flowpath downstream of Regen. Heat Exchanger (including IC 3
2 1
0 0
Aux. Spray)
IC 2
2 2
0 0
b_. Charging flowpath between con-IC 3
2 6
1 5
tainment Penetration and Regen.
Heat Exchanger
- c. Normal letdown flowpath upstream IC 2
1 of Regen. Heat Exchanger IC 2
2
- d. Normal letdown between Regen.
IC 3
2 20 2
5 Heat Exchanger and letdown orifices IC 2
2
- e. Normal letdown from letdown orifices IC 2
2 to containment penetration B4-12241-7565 l
3
l l
l l
i i
Page 2 of 4 l
ATTACHMENT C POSTULATED ARBITRARY INTERMEDIATE BREAKS (AIBs)
- BREAKS / PROTECTION DEVICES ELIMINATED -
Estimated Number Deleted Pipina system Location
- NPS Safety Class AIBs RESTRAINTS SHIELDS
1 8
2 0
IC 1
1 IC 2
2 2
1 0
2 No. 1 Seal Bypass I
1 11 3
0 valves
.~
i,. Excess letdown between RCPB isola-IC 2
1 2
1 0
tion valves J. Normal letdown between containment OC 2
2 penetration and Nonregen. Heat Exchanger 16 3
1 k,. Normal letdown downstream of OC 2
2 Nonregen. Heat Exchanger
- 1. Charging / Seal water injection CC 4
2 lines from charging pumps to OC 3
2 containment penetrations OC 2
2 40 7
3 OC 1\\
2 Auxiliary Feedwater into Main IC 4
2 4
0 0
Fn dwater lines Mais Feedwater
- a. Piping between containment pene-IC 16 2
6 18 3
tration and steam generators b,._ Piping outboard of break exclu-OC 16 NNS 2
6 0
sion zone OC 6
NNS B4-12241-7565
?
Page 3 of 4 ATTACHMENT C POSTULATED ARBITRARY INTERMEDIATE BREAYJi (AIBs)
- BREAKS / PROTECTION DEVICES ELIMINATED -
Estimated Number Deleted Pipina System Location
- NPS Safety Class AIBs RESTRAINTS
- SHIELDS Gaasous Nitrogen System OC 1
NNS 8
3 3
Main Steam Piping a,. Piping between containment pene-IC 32 2
6 7
3 tration and steam generators
- b. Piping outboard of break exclu-OC 32 NNS 6
21 0
sion zone Rrector Coolant System
- a. Pressurizer Spray lines IC 6
1 4
1 12 4
4 1\\
1
- b. 8 in. Bypass line including IC 8
1 6
3 9
branch connections 2
1
- c. Pressurizer Safety and Power-IC 6
1 3
1 1
Operated Relief Valve Inlet.
3 1
lines R>sidual Heat Removal
to second RCPB isolation valve 1
0 0
- b. Piping connected to accum.
IC 10 1
injection line (loops 22 & 23)
B4-12241-7565
m.
_,.m m.
Page 4 of 4 ATTAC}e!ENT C POSTULATED ARBITRARY INTERMEDIATE BREAKS (AIBs)
- BREAKS / PROTECTION DEVICES ELIMINATED -
1 Estimated Number Deleted Piping System Location
- a. SIS lines normally connected OC 4
2 8
3 3
t to NHSI pumps 3
2 4
J-b_. Low /High Head Pump Injection IC 6
1 a
lines from RCS loop connection 3
1 l
to second RCPB isolation valve 2
1 c_. Piping from RCS loop connection IC 12 2
1 0
0' to accumulator tanks 2
2 6
0 0
TOTALS 245 105 56 i
- Location code OC: lines located outside containment structure IC: lines located inside containment structure
- Stress and break locations not available
- NPS:
Nominal Pipe Size 1
- Includes branch connections although not specifically identified f
I t
I t
B4-12241-7565 t
i O
+,
Page I of 4 ATTACHMENT D POSTULATED ARBITRARY INTERMEDIATE BREAKS (AIBs)
- SYSTEMS SUltlARY -
Sys. FSAR Operating Piping System ****
Sec. No.
NPS Safety Class Pipe Class' Pipe Material Temperature ('F)
Auxiliary Steam System 10.4.10 10 NNS 151 CS 360 8
NNS 151 CS 360 6
NNS 151 CS 360 4
NNS 151, 601 CS 360 3
NNS 151 CS 360 2
NNS 151 CS 360 1
NNS 151 CS 360 Steam Generator Blowdown 10.4.8 3
2 1502 SS 547 25 2
1502 SS 547 2
2 601 CS 547 3
NNS 901 CS 547 15 NNS 901 CS 390 Chemical and wolume Control 9.3.4
- a. Charging flowpath downstream of 3
1 1502 SS 495 Regen. Heat Exchanger (including 3
2 1502 SS 495 Auxiliary Spray) 2 2
?502 SS 130
- b. Charging flowpath between con-3 2
1502 SS 130 tainment Penetration and Regen.
Heat Exchanger
- c. Normal letdown flowpath upstream 2
1 1502 SS 549 2
2 1502 SS 549
~ of Regen. Heat Exchanger
- d. Normal letdown between Regen.
3 2
1502 SS 287 Heat Exchanger and letdown orifices 2
2 1502 SS
- e. Normal letdown from letdown orifices 2
2 602 SS 287 to containment genetration j
-f.
', =1 ".ter I sjection to RCPs 2
1 1502 SS 130 l\\
l 1502 SS 130 2
.2 1502 SS 130 f
a H4-12241-7566 e
e
Page 2 of 4 POSTULATED ARBITRARY INTERMEDIATE BREAKS (AIBs)
- SYSTEMS
SUMMARY
Sys. FSAR Operating Piping System Sec. No.
NPS Safety Class Pipe Classt Pipe Materia 12 Temperature (*F)
- 3. Seal water return from RCPs 2
2 1502 SS No. 1 Seal Bypass
- h. Loop fill between RCPB isolation 2
1 1502 SS65-104 valves i_. Excess letdown between RCPB isola-2 1
1502 SS 85-105***
tion valves
- 1. Normal letdown between containment 2
2 602 SS 287 penetration and Nonregen. Heat Exchanger
- k. Normal letdown downstream of 2
2 602 SS 115 Nonregen. Heat Exchanger
- 1. Charging / Seal water injection 4
2 1502 SS 130 lines from charging pumps to 3
2 1502 SS 130 containment penetrations 2
2 1502 SS 130 1
2 1502 SS 130 Auxiliary Feedwater into Main 10.4.9 4
2 601 CS 443 Feedwater lines Mila Feedwater
- a. Piping between containment pene-10.4.7 16 2-601 CS 443 tration and steam generators i
B4-12241-7566
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l Page 3 of 4 ATTACHMENT D POSTULATED ARBITRARY INTERMEDIATE BREAKS (AIBs)
- SYSTEMS
SUMMARY
Sys. FSAR Operating Pipina System Sec. No.
NPS Safety Class Pipe classi Pipe Material: Temperature (*F)
- b. Piping outboard of break exclu-16 NNS 901 CS 443 sion zone 6
NNS 901 CS 443 Gawous Nitrogen System 9.5.9
- a. Normally connected to main steam 1
NNS 601 CS 547 system
- b. In other areas _
1h NNS 151,601 CS65-104 Maia Steam Piping 10.3
- a. Piping between containment pene-32 2
601 CS 547 tration and steam generators
- b. Piping outboard of break exclu-32 NNS 601 CS 547 sion zone Reretor Coolant System 5.3 a,.
Pressurizer Spray lines 6
1 1502 SS 542-656 4
1 1502 SS 542-656 1
1 1502 SS85-105
- b. 8 in. Bypass line including 8
1 2501Rf SS 542 branch connections 2
1502 SS 542
~c. Pressurizer Safety and Power-6 1
1502 SS 400/656*
Operated Relief valve Inlet 3
1 1502 SS 400/656*
lines B4-12241-7566 i
Page 4 of 4 ATTACHMENT D POSTULATED ARBITRARY INTERMEb! ATE BREAKS (AIBs)
- SYSTEMS
SUMMARY
Sys. FSAR Operating Pipina System Sec. No.
NPS Safety Class Pipe Class 3 Pipe Materiala Temperature (*F)
Re2idual Heat Removal 5.4.7
- a. Piping from RCS loeg connection 12 1
1502 SS 610**
to second RCPB isolation valve b_. Piping connected to accum.
10 1
1502 SS 85-105**
injection line (loops 22 & 23)
S2fety Injection System (SIS) 6.3
- a. SIS lines normally connected 4
2 b_. Low /High Head Pump Injection 6
1 1502 SS 610-Hot leg conn.
lines from RCS loop connection 3
1 1502 SS 542-Cold leg conn.
to second RCPB isolation valve 2
1 1502 SS
- c. Piping from RCS loop connection 12 1
1502 SS 542 to accumulator tanks 12 2
602, 1502 SS85-105 2
2 602 SS85-105
- Water solid loop seal maintained at 400*F
- Excludes temperatures for RHR system operation
- Line can see loop temperature conditions during heatup or when normal letdown path inoperable.
- Includes branch connections although not specifically identified.
1.
Pipe Classes: 150-151 Lb Carbon Steel Piping
- 2. Material: CS-Carbon Steel SA/A106, Gr. B or C 601-600 Lb Carbon Steel Piping SS-Stainless Steel: Class 602 is 602-600 Lb Type 304 Stainless Steel Piping SA/A376 or 312 Type 304. Class 901-900 Lb Carbon Steel Piping 1502 is SA/A376 or 312 Type 316.
1502-1500 Lb Type 316 Stainless Steel Piping Class 2501Rf is SA376 Type 316.
250lR#-1500 Lb Type 316 Stainless Steel Piping 2
B4-12241-7566
ATTACHMENT E DUQUESNE LIGHT COMPANY BEAVER VALLEY POWER STATION UNIT 2 PROTECTION OF HIGH ENERGY LINES FROM STRESS CORROSION CRACKING Stainless Steel Lines:
All high energy piping at BVPS-2 made with stainless steel is made of alloy Type 304 or Type 316.
High carbon. grades Type 304H and Type 316H are not used.
The piping material was furnished in the solution annealed condition and fabricated using forming and welding practices which limited the degree of sensitization consistent with Regulatory Guide 1.44 recommendations, f
While the controls used to minimize sensitization alone do not prevent the possibility of stress corrosion, expe rience and tests have shown that by maintaining a high degree of cleanliness. and controls on fluid chemistry that stress corrosion will not occur under these conditions.
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The fluid chemistry requirements specified for the Reactor Coolant System (RCS) are provided in FSA R Table 5.2-5 (A t t achment E-1).
The strict i
limitations on halogens and oxygen have proven to be ef fective in prevent-i ing stress corrosion cracking.
Makeup water to the RCS and the remaining secondary side safety related high energy lines, with the exception of 4
steam generator blowdown, are subject to the same strict halogen require-ments.
These strict controls have proven to be ef fective in preventing j
stress corrosion cracking in these systems which operate at lower tempera-i tures than the RCS.
The stean ge nerato r blowdown system chemistry is controlled by the feedwater system conditions described below, thus j
preventing stress corrosion cracking. The steam generator blowdown system j
is continuously monitored to ensure correct water chemistry is maintained.
Caustic stress corrosion is prevented because of the lack of additives that can generate caustics.
i Stress corrosion cracking initiated from the outside surfaces of the f
piping is prevented by the use of strict cleaning procedures followed by swipe testing to ens ure low chloride and fluoride levels and the use of thermal insulation supplied in accordance with the recommend at ions of Regulatory Guide 1.36.
a Carbon Steel Lines:
1 Carbon steel high energy piping are protected from stress corrosion crack-ing due to the water chemistry controls specified in FSAR Table 10.4-13
( A t tachme nt E-2).
All-volatile chemistry treatment in the feedwater system maintains oxygen concentration and pH within desired levels.
The f eed vate r system is also continuously monitored to ensure correct water chemistry.
No caustic is present in this environment, precluding the possibility of caustic stress corrosion cracking of the carbon steels.
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ATTACHMENT E-1 l
BVPS-2 FSAR TABLE 5.2-5 REACTOR COOLANT CHEMISTRY SPECIFICATION Electrical conductivity Determined by the concentration of boric acid and alkali present.
Expected range is <1 to 40 pMhos/cm at 258C.
Solution pH Determined by the concentration of boric acid and alkali present.
Expected values range between 4.2 (high boric acid concentration) and 10.5 (low boric acid concentration) at 250C.
Oxygen (1) 0.005 ppm, maximum Chloride <2>
0.15 ppm, maximum Fluoride (2>
0.15 ppm, maximum Hydrogen (3) 25-50 cc (STP)/kg H O 2
Suspended solids (4) 1.0 ppm, maximum pH control agent (Li70H)(5) 0.7-2.2 ppm as Li Boric acid Variable from 0-4000 ppm as B Silica (8) 0.2 ppm, maximum i
0.05 ppm, maximum Calcium (88 0.05 ppm, maximum Magnesium 8 0.05 ppm, maximum NOTES:
1.
Oxygen concentration must be controlled to less than 0.1 ppm in the reactor coolant at temperatures above 180 F by scavenging with hydrazine or by maintaining the proper hydrogen concentration.
During power operation with the specified hydrogen concentration maintained in the coolant, the residual oxygen concentration must not exceed 0.005 ppm.
2.
Halogen concentrations must be maintained below the specified values at all times regardless of system temperature.
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BVPS-2 FSAR
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TABLE 5.2-5 (Cont)
NOTES:
(Cont) 3.
Hydrogen must be maintained in the reactor coolant for all plant operations with nuclear power above 1 MWt.
The normal operating range should be 30-40 cc/kg H 0.
3 4.
Solids concentration determined by filtration through filter having 0.45 micron pore size.
5.
The specified limits for lithium hydroxide must be established for prestart-up testing prior to heatup beyond 150'F.
During cold hydrostatic testing and hot functional testing, in the absence of boric acid, the reactor coolant limits for lithium hydroxide must be maintained to provide inhibition of halogen stress corrosion j
cracking. Upon plant restart, the lithium hydroxide limits should be established at 180*F.
I 6.
These limits are included as recommended standards for monitoring coolant purity. Establishing coolant purity within the limits shown for the species is judged desirable with the current data base to minimize fuel clad crud deposition which affects the corrosion resistance and heat transfer of the clad.
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i 8vPS-2 FSAR TABLE 10.4-13 STEAM GENERATOR STEAM SIDE AND FEEDWATER CHEMISTRY SPECIFICATIONS Cold Hydro /
Cold Hot Functional /
Wet Hot Shutdown /
Layup Hot Standbv Start-up From Hot Standby Normal Power Operatfon Ctw.m i s t ry Blowdown Blowdown Feedwater
_ 810wdown Feedwater Blowdown Perameter.
Control Control Expected Expected Control Control Expected Expected Control Control Expected pH e 25*C 10.0 -
8.8 -
8.8 -
8.8 -
NA 8.5 -
8.5 -
8.8 -
NA 8.5 -
8.5 -
10.5 9.2 9.2 10.0 10.0 10.0 9.2 9.0 9.0 Free hydroxide NO O.15
<O.15 NA NA O.15 (O.15 NA NA O.15
<O.15 cs ppm CaCOs Catton conduc-NA 2.0
<2.0 NA NA 7
<7 NA NA 2.0
<2.0 ttvlty mhos/cm
- 25*C Total conduc-NA NA NA NA NA NA NA 54 NA NA NA ttvity phos /cm P 25*C Sodlum, ppm NA NA
<O.1 NA NA NA
<O.5 NA NA NA
<O.1
-Chloride, ppm
<O.5 NA
<O.15 NA NA NA
<O.5 NA NA NA
<0.15 NHs, ppm As pH
. NA 50.5 NA NA NA
<10.0 50.5 NA NA 50.25 requires
[Os ] +
NA NA
[On] +
[On] +
NA NA Hydrazine, ppm 75 -
NA NA
[0,j
+
150 0.005 0.005 0.005 0.005 Dissolved
<100 NA
<5
<100
<100 NA
<5
<5
<5 NA
<5 oxygen, ppb Stue, ppm NA NA
<1.0 NA NA NA
<5 NA NA NA
<1.0 H
Fe, ppb NA NA NA
<100 NA NA NA
<10 NA NA NA H>
Cu, ppb NA NA NA
<50 NA NA NA
<5 NA NA NA Suspended NA NA
<1 NA NA.
NA NA NA NA NA
<1.0 Zd zolids, ppm tv1 Q10wdown NA NA As NA NA Maximum Maximum NA NA As
<1.0 E
rate, gpm/SG required required to main-tain control parameters 1 of 1
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4 ATTACHMENT F DUQUESNE LIGHT COMPANY l
BEAVER VALLEY POWER STATION UNIT 2 i
PROVISIONS FOR MINIMIZING THE EFFECTS OF THERMAL AND VIBRATION INDUCED PIPING FATIGUE i
I.
GENERAL FATIGURE DESIGN CONSIDERATIONS i
For Class 1 lines, fatigue considerations are addressed by the cumulat-ive usage factor (CUF).
In order to ens ure that piping will not fail due to fatigue, the ASME Code has set the CUF limit at 1.0.
By defini-tion, all arbitrary intermediate break locations have CUFs below 0.1.
i For Class 2 and 3 lines, fatigue is considered in the allowable stress s-range check for thermal expansion stresses.
This stress is included in i
the total stress value used to determine postulated break locations.
All arbitrary break locations exhibit stresses less than 80 percent of I
l the code allowables.
If the number of-thermal cycles is expected to be greater than 7,000, then the allowable stresses are further reduced by
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an amount dependent on the number of cycles.
j II.
VIBRATION DESIGN CONSIDERATIONS i
f BVPS-2 piping systems are designed and supported to minimize transient and steady state vibrations.
Vibration levels are observed or measured during preoperational testing j
for both steady state and transient vibration conditions.
The program 1
used to monitor these conditions is described below.
1 Steady-State Vibrations:
Visual observations are used for judging acceptability of steady-state vibration.
Visual observations may be aided by hand-held instruments (e.g., vibrometers) when considered appropriate by engi-neers experienced in piping design.
i A screening velocity or displacement will be established for use with l
hand-held instrument results.
If the measurement indicates that the velocity or displacement limit is exceeded, the measured values are l
reconciled with the respective analyses by considering the speci fic i
piping configuration, velocity or displacement amplitude me as ur ed,
i stress indices, and the endurance strength of the material properly accounting for high cycle effects.
If system modifications are 4
required, the applicable ASME design calculations are reconciled to assure acceptable system characteristics for all applicable design conditions.
j The maximum alternating stress intensity will be used to est ablish the acceptance stress criteria for steady state vibrations.
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Transient Vibrations:
Trans ient vibration condit ions are subjected to visual and instru-me nted observations.
When ins t rument ed observations are t ake n, the acceptance criteria are based on the applicable fluid system tran-sient analysis (stress, de flect ion, etc.) results.
Instrumented observations are conside red acceptable if they are within the tran-s ient analysis results acceptance criteria, the results are recon-ciled with the design analysis.
When sys t em modifications are required to achieve accept ab le levels of transient vibration, the ASME design calculations are reviewed and modified as necessary to assure acceptable system characteristics.
III. THERMAL DESIGN CONSIDERATIONS 4
4 During normal operation, mixing (backflow from the steam generator into the feedwate r line) of hot water in the stean generator with the lower t emperature feedwater is localized to the area of the feedwater inlet nozzle.
This is accomplished by both the piping arrangement and the i
l main and bypass feedwater control valves.
The feedwater piping is arranged with a downward 90 degree elbow welded 1
to the stean generator inlet nozzle coupled with a vertical run of pipe.
A thermal protective liner extending from the nozzle back through the 90 1
j degree elbow will minimize thermally induced cyclic stresses on the nozzle and inlet piping.
1 Flow surges due to main and bypass feedwater control valve modul at ion i
have been minimized due to control valve design (trim / flow coef ficient l
design) coupled with a valve positioner wh ich adds to control valve i
stability.
4 Mixing is further reduced by the presence of inverted "J" tubes located i
on the top of the feedwater ring in the steam generator.
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ATTACHMENT G DUQUESNE LIGHT COMPANY BEAVER VALLEY POWER STATION UNIT 2 PROVISIONS FOR MINIMIZING STEAM / WATER HAMMER EFFECTS Systems within Westinghouse scope of supply are not in general susceptible to water hammer.
The reactor coolant, chemical and volume control and residual heat removal systems have been specifically designed to preclude water hammer.
Operating experience at other plants with Westinghouse systems have verified this design approach.
Westinghouse has conducted a number of inves t igat ions into the causes and consequences of water hammer events.
The res ult s of these inves tigations h ave been reported to Westinghouse operating plant custome rs and have been refle ct ed in des ign interface requirements to the BOP designer for plant s under cons truct ion, to assure that water hammer events initiated in the secondary systems do not compromise the performance of the Westinghouse-supplied safety-related systems and components.
In general, the approaches taken, individually or _in combination, to address water hammer concerns were to prevent / minimize water hammer ef fects through system design features / operating procedures.
Potential water hanmer sources to be considered were based on industry experience and the concerns presented in NUREG-0582 " Water Hammer in Nuclear Power Plants." The following discusses in more detail the potential water hanmer sources, if any, that were consid-e red in the design of the subject systems and the actions taken to minimize and prevent water hammer effects.
There is a very low po tent ial for water hanmer in the subcooled water solid portions of RCS since these portions of RCS are designed to preclude void formation.
Safety Injection System (SIS):
As discussed below, it is considered unlikely that water hammer induced pipe ruptures could occur in the Safety Injection System.
The low temperature SIS lines,- which are normally water solid, have a very low probability of stean void formation.
Proper initial fill and venting ensures that low and high head safety inj ect ion system piping remains f illed.
In addit ion, the head of water provided by the RWST provides a continuous mech anism for ens uring that the low' head safety injection system lines remain full.
For the SIS lines, which are part of the Residual Heat Removal System return flowpath, operating procedures for RHRS minimize the potential fo r water hammer in these lines.
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For the SIS lines, which are part of the Reactor Coolant Pressure Boundary to the first isolation valve, there is a very low potential fo r water hammer as indicated in the above RCS discussion.
Residual Heat Removal System (RHRS):
Portions of the RHRS piping is high energy because it is normally pressur-ized by RCS or SIS during normal plant operating conditions. As discussed below, it is unlikely that water hammer induced pipe ruptures could occur in the Residual Heat Removal System.
When RHRS is not operating, the normally pressurized portions of the system are water solid and are either at a low temperature or subcooled.
When RHRS is operating (i.e., the short operational period), valve closure times and operating procedures minimize the po tent ial fo r wat e r hamme r.
Proper fill and venting will initially ensure that air does not become trapped in any part of the RHRS during start-up. Additionally, just prior to RHRS Initiation, the RHRS will be cross-connected with the CVCS (via 2RHS*MOV750A, and B and 2CHS*HCV142). The valves are shown on FSAR Figures 5.4-4 and 9.3-24, res pect ively.
This action utilizes the pressure head in the CVCS to collapse any voids (should they remain) prior to opening the RHRS suction (2RHS*MOV701A and B and 2RHS*MOV702A and B) valves from the RCS.
Steam Generator Blowdown (SGB):
Fluid flow in the SGB lines is normally two phase flow from the steam generators to the feedwater injection taps downstream of the containment isolation valves outside containment.
There is very little probability of water hammer occurring in this po rt ion of piping because of the very low percent quality (approximately 10 percent).
t Because of the greater susceptibility for water slugging to occur in the SGB lines outboard of the contaiment isolation valves, a connection from the feedwater system was provided to inject feedwat er into the blowdown flow. This will subcool the blowdown flow and prevent water slugging.
In addition to the above, operating procedures provide precautions to prevent thermal shock and water hammer of system components while starting up the system or when changing loads.
Auxiliary Steam Systems (ASS):
The pot ent i al for water hammer has been minimized by means of operating procedures (e.g., warming up lines prior to continuous operation to remove any existing condens at io n), sloping of piping to eliminate pockets of condensate and providing drains / steam traps to continuously remove any condensate accumulating during heat-up and operation, i
Gaseous Nitrogen System (GNS):
There are no potential water hammers in gaseous lines once the system is vented and charged with nitrogen.
The GNS lines normally connected to 2
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MSS lines are part of the MSS pre ss ure bounda ry ( i.e., to the first normally closed isolation valve off MSS line).
Chemical and Volume Control (CVCS):
The low. temperature CVCS lines, which are normally water so lid, have a very small probability of steam void fo rma t ion.
Accordingly, no water hammer would be expected.
In addition, for those lines which are part of the charging flowpath, operating procedures should prevent any water hammer effects associated with the starting of a charging pump.
In the high temperature CVCS lines system de s ign, a pressure regula ting valve downstream of the letdown restriction orifices prevents flashing and maintains water solid conditions during normal plant operation so water hammer would ot be expected.
Valve interlocks, as well as operating procedures, have been provided to prevent flashing when starting up or securing the letdown flowpath.
Main Feedwater System (FWS):
The potential for water hammer in the FWS caused by rapid condensation of a steam bubble in the steam generator feedring (NUREG-0582, Item A-8) has been minimizing by incorporation of the following features:
- 1. The steam generator feedring is provided with J-tubes to prevent drain-age of water during low steam generator water level.
- 2. The feedwater piping connections of the steam generators are made with a 90 degree elbow arrangement which does not present a horizontal pipe run immediately upstream of the feedwater nozzles.
This configuration should preve nt fo rma tion of steam po cke t s unde r low steam gene rator water level conditions and minimize the volume of water external to the steam generator which could pocket a steam bubble. This piping arrange-ment follows Westinghouse design guidelines.
Industry experience has shown that the above de sign fe atures have mini-mized, if not eliminated this type of water hamme r.
During start-up testing at BVPS-2, tests will be conduct ed to demonstrate the effective-ness of this design.
In addi t ion, the main feedwater control valves, supplied by W have been s ubj ect ed to extensive design considerations and iterations to ensure the valve size and trim is compatib le with the remainder of the fe edwate r system design.
To ensure feedwater flow control at low powe r levels (approximately 15 percent), a bypass control valve is provided around each main feedwater control valve.
Auxiliary Feedwater System (AFWS):
The AFWS does not operate during normal operation.
The majority of the A FWS is moderate energy as only that port ion which is part of the FWS up to the isolation valve in containment is considered high energy.
- Thus, for this 4-inch section of piping, the discussion fo r the FWS regarding water hammer applies.
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Operating procedures (e.g.,
filling and venting procedures) minimize the potential for water hammer associated with pump starting.
Main Steam System (MSS):
i.
The potent ial of water hanme r in the MSS caused by water entrained in stean lines has been minimized due to the piping arrangement. MSS piping is sloped to eliminate pockets where condensate could collect, and conden-sate is continuously removed via the steam drains system.
Heating up the MSS lines downstream of the main steam isolation valves (MSIV) prior to opening the MSIVs is accomplished using the 2-inch bypass line. This removes any condensate which may have accumulated in the lines prior to opening the MSIV's.
Ope rat ing procedures ident ify specific actions to minimize / prevent water hammer.
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ATTACHMENT H DUQUESNE LIGHT COMPANY BEAVER VALLEY POWER STATION UNIT 2 i
ELIMINATION OF ARBITRARY INTERMEDIATE BREAKS ENVIRONMENTAL ANALYSIS There will be no change in the result s of Beaver Valley Power Station Unit 2 environment al analysis due to elimination of arbitrary intermediate breaks.
The break postulation for environmental ef fects, as indicated in FSAR Section 3.6 B.1. 3.4. 3, is performed independently of break postulation for dynanic (i.e., whip / jet) ef fect s.
When postulating breaks for environmental ef fect s, a break is assumed to occur nonmechanistically anywhere along the piping run.
The break location chosen for purposes of equipment qualification is the one which causes the worst environmental effect on safety-related equipment.
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ATTACHMENT I DUQUESNE LIGHT COMPANY BEAVER VALLEY POWER STATION UNIT 2 PROVISIONS FOR MINIMIZING LOCAL STRESSES FROM WELDED ATTACHMENTS All of the arbitrary intermediate break locations to be eliminated have been reviewed, and it has been de termined that in only six cases are welded attachment s close to the postulated breaks.
In each of these caser, the at t achme nt stresses have been appropriately added to the arbitrary break point. stresses and the resultant stresses were less than the break postula-tion stress limit.
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