ML19260C894
| ML19260C894 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 01/17/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17174A409 | List: |
| References | |
| TASK-03-05.B, TASK-3-5.B, TASK-RR NUDOCS 8002060391 | |
| Download: ML19260C894 (19) | |
Text
.
SEP REVIEW OF PIPE BREAK OUTSIDE CONTAINMENT TOPIC III-5.8 FOR THE DRESDEN NUCLEAR POWER PLANT UNIT 2 1936 033 i'tf9E3*f3' S00206033f
INTRODUCTION The safety objective of Systematic Evaluation Program (SEP) Topic III-5.3,
" Pipe Break Outside Containment" is to assure that pipe breaks would not cause the loss of needed functions of safety-related systems, structures and components and to assure that the plant can be safely shut down in the event of such breaks.
The needed functions of safety-related systems are those functions required to mitigate the effects of the pipe break.
The current criteria for review of pipe breaks outside containment are contained in Standard Review Plan 3.6.1 and 3.6.2 including their attached Branch Technical Positions.
BACKGROUND In December 1972, the staff sent letters (Reference 1) to all power reactor licensees requesting an analysis of the effects of postulated failures of high energy lines outside of containment. A summary of the criteria and requirements in this letter is set forth below:
Protection of equipment and structures necessary to shut down the a.
reactor and maintain it in a safe shutdown condition, assuming a concurrent and unrelated single active failure of protected equipment, should be provided from all effects resulting from ructures in pipes carrying high energy fluid, wnere the temperature and pressure conditions of the fluid exceed 200*F and 275 psig, respectively, up to an including a double ended ruoutre of such pipes.
Breaks should be assumed to occur in those locations specified in the " pipe whip criteria". The rupture effects to be considered incluce pipe whio, r
structural (including the effects of jet impingement), and environmental.
I b.
In addition, protection of eouipment and structures necessary to shut down the reactor and maintain it in a safe shutdown condition, assuming a concurrent and unrelated. single active failure of protected eouipment, should be provicea from the environmental and structural
, em m %
yyc y 300 1936 034
effects (including the effects of jet imoingement) resulting from a single open crack at the most adverse location in pipes carrying fluid routed in the vicinity of this equipment.
The size of the cracks should be assumed to be 1/2 the pipe diameter in length and 1/2 the wall thickness in width.
(
In response to our letter, Commonwealth Edision submitted Dresden Station Units 2 & 3 Special Report No. 37 " Analysis of Effects of Pipe Break Outside the Primary Containment", dated January 23, 1974, and Supplement 1 and Revision 1 to this report dated March 22, 1974 and February 19, 1975, respectively (References 2, 3, 4).
These submittals were reviewed and accepted by the staff in our safety evaluation report of May 12, 1976 (Reference 5).
On Novemcer 17, 1977, the licensee submitted Appendix E to Special Report No. 37 whicn analyzed the effect of pipe breaks outside containment on electrical conduits and cable trays in Dresden 2 & 3 and identified areas requiring modifications (References 6).
The NRC staff reevaluation of the effects of pipe breaks outside containment under SEP Topic III 8..B includes the comoarison of Dresden 2 with current criteria for pipe breaks outside containment and the evaluation of Appendix E to Special Report No. 37.
The staff used an " effects oriented" approach to determine the acceptability of plant response to pipe breaks, i.e. each structure, system, comoonent, and power supply which must function to mitigate the effects of the pipe break and to safely shutdown the plant was examined to determine i
its susceptibility to the effects o.f the postulated break. Break effects considered were ccmpartment pressuri:r. tion, pipe whip, jet imoingement, spray, flooding, and environmental conditions of temoerature, pressure, and humidity.
1936 035
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(The effects of potential missiles generated by fluid system ruptures and rotating machinery were also considered and are evaluated under SEP Topic III-4.C, " Internally Generated Missiles".)
i Cata for this assessment was gathere'd during a visit to the Dresden 2 plant on August 27-30, 1979.
EVALUATION The results of the SEP reevaluation of pipe breaks outside containment of Dresden 2 are provided in Table 1.
The following paragraphs provide additional information used to evaluate certain pipe breaks listed in Taole 1.
Current criteria for pipe breaks outside containment assume the loss of offsite power if the pipe break results in a reactor or main turbine trip.
The Dresden 2 facility has never experienced a total loss of offsite power resulting from a unit trip (reactor and turoine).
The Dresden Station has experienced only one total loss of offsite pcwer and this was caused by a tornado on November 12, 1965.
Therefore, for the Drescen site, the assumed loss of offsite power resulting from a unit trip is extremely con.arvative.
The safe shutcown systems which were examined from the stancpcint of protection f
from pipe break effects are identified in the SEP Safe Shutdown Review f r Q
Dresden 2 (Reference 3).
These systems are:
036 3
(a) Reactor Control and Protection System (b) Pressure Relief System (4 electromatic/1 electro pneumatic relief s
valves)
(c) Low Pressure Coolant Injection System (d) High Pressure Coolant Inje_ction System (e) Containment Cooling Service Water (f) Instrumt tation of shutdown /cooldown (g) Emergency power (AC and DC) and control power for the aoove systems and components.
Table 1 provides an assessment of circulating water (CW) system breaks in the crib house.
Reference 7 contains an evaluation of the effects of a CW line break in the condenser pit of the turbine building.
To prevent a break in the condenser pit from flooding critical equipment, the licensee has installed flood doors between the conder.ser pit and the condensate pump room.
Permanent flood level switches have been installed in the pit to warn the operator of a flooding conoition and to trip the CW pumps if necessary.
The ficoding rates used in Reference 7 are much greater than would be assumed using current MEL3 criteria since the licensee assumed a complete severance of a CW line.
- Also, as part of the permanent modifications resulting from the analysis in Reference 7, tne diesel generator cooling water pumps (#2, #3, ar# #2/3) in the crib house have been replaced with submersible pumps, and containment cooling service water pumps 28 and 2C have been surrot.nded with a watartignt enclosure wnich extends from the 495' elevation to the 511' elevation.
The maximum historical flood level for the river at the crib house is 508' w
1936 037
At present, the fire protection water system is normally pressuri:ed through a cross-connect from the service water system (SWS).
A moderate energy leak from a fire protection pipe anywhere in the plant would not result in a k
noticeable decrease in fire protection system pressure, so the leak would continue until it was discovered by the plant operator or until the amount of water leaked had caused drain sump. level alarms or equipment damage indications in the control room. To remedy this situation, the licensee plans to modify the fire protection water system so that a leak would result in a noticeable decrease in fire protection system pressure and activate an alarm.
This modification is currently scheduled to be made in January or February 1980.
This modification will suostantially improve the operators aoility to detect and stop fire system leaks before essential equipment is flooded.
The control rod drive (CRD) hydraulic system return line, which penetrates primary containment on the 545' elev. of the Reactor building, is currently valved out of service as a result of ongoing studies of CR0 return line no::le cracking.
Pipe breaks associated with the CRD hydraulic control units, on the 517' elev.
of the Reactor Building, could involve 1) the drive insert and withdraw lines wnich lead through containment penentration to the drives, 2) the CRD hydraulic drive and cooling water lines, 3) the CRD charging water line, and 4) the CRD g
return line. A break in the drive withdraw line would cause its associated control rod to insert (scram). A break in any of the other lines would cause the rod to remain in position but the red could still ce inserted by the
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operator or by a reactor protection system scram signal.
Loss of electrical power to the CR0 hydraulic control unit would also result in a rod insertion.
Therefore, potential pipe break damage to the CR0 hydraulic control units d
would not
- event control rod trip (scram) by the operator or the reactor protection system.
The Containment Cooling Service Water (CCSW) system pumps, which are in the Turbine Building north side (495' elev.) above the condensate pump room, take suction from the crib house via buried pipe. No means to isolate this pipe at the crib house end exists; so a MEL8 in the condensate pump room at the CCSW pump end of the pipe would result in flooding the condensate pump room to the level of the river at the crib house.
Postulated site flood levels in excess of 511' would result in the complete loss of the CCSW system.
CONCLUSIONS Based on the information submitted by the licensee and obtained during our site visit to Dresden 2, we have cetermined that the followir.g review areas have not been addressed in previous staff safety evaluations and should be resolved with the SEP:
1.
Postulated pipe breaks outside of the primtry containment between the containment penetration and the first containment isolation valve have not been evaluated for the main steam lines, isolation condenser steam and condensate lines, and reactor water cleanup inlet line.
Currently 5
4940-00t i936 039
the staff applies the provisions of Branch Tecnnical Position ME3 3-1 (Reference 10) section 8.1.b., to the review of the postulated break
-4 areas.
The licensee will be required to ccmaare the design of the Oresden 2 systems with these current regulatory provisions.
2.
The effects of postulated pipe breaks in certain systems could result in damage to the containment isolation valves or power supply and control cables to the containment isolation valves for those systems. The combi-nation of the single active failure provision and damage to the containment isolation valve could result in an unisolable break flow path.
The systems of concern are the isolation condenser system (steam line only),
and reactor water cleanup system (inlet line).
The staff currently applies the provisions of Branch Technical Position ASB 3-1 (Reference 11), Section 8.2.c., to the review of these break areas.
The licensee will be required to compare the design of the Dresden 2 system with these current regulatory provisions.
3.
The consequences of a potential main feed system line break in the main feed regulatory valve area (turoine cuild 538' elev.) have not been determined with regard to possible damage to the engineered safeq.,2rds system electrical caoles in that area. The licenree will ce required to evaluate the consequences of a possible feed line break in this area and I
provide adequate protection as required.
(This area is also being reviewed uncer the NRC fire protection evaluation to determine the effects of a fire on the electrical cables in that area. )
1936 040 pyj()-002" 7
4 A main feeo sy<te If ne break in the main feed pump area (turbine building 517' elev.) could damage the bus duct connecting diesel generator #2 to the 4KV switchgear in the reactor building.
Current criteria postulate a i
loss of offsite power (caused by a unit trip on loss of feedwater) and a potential single active failare of the #2/3 diesel generator.
This scenario would result in loss of all AC power to the unit.
5.
A properly oriented moderate energy line break of the reactor building closed cooling water system could cause the flooding of the recundant 4KV switchgear 23-1 and 24-1 (on the 545' elev of the reactor building) to the level of the curo surrounding the switchgear - about 5".
Items 1, 2, 3 and 5 above constitute potential devations from current criteria for postulated pipe breaks outside containment.
Additional information will be obtained from the licensee to determine if 1) the systems identifed in items 1 and 2 conform to the provisions of References 10 and 11, 2) if the costulated flooding identified in item 5 represents a hazard to switchgear 23-1 and 24-1, and 3) if a feedwater pipe ruoture, as described in item 3, could damage electrical cables for system which are required to function to mitigate the effects of the rupture or to safely snutdown the plant.
Although item 4 above results in the total loss of AC power on the site, the likelihood of a unit trip resulting in the loss of offsite power is extremely small based on the coerational history of this plant. Thus, AC power would be available to operate the safe shutcown systems wnich were identified in the evaluation section of this recort. Nevertheless, should a loss of offsite power be 1936 041 WC00
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postulated, the isolation condenser system would be available to remove reactor decay heat and commence a plant cooldown.
In this case, makeup water to the 4
isolation condenser would be provided by the fire protection system.
The twenty minutes of shell side water in the condenser provides adequate time to start a diesel driven fire pump and assure a makeup flow path from the fire system to the isolation condenser. Decay heat removal using the isolation condenser can be maintained while repairs are made on the diesel generator which was assumed to fail (single active failure) or until offsite power is regained.
The staff is continuing its evaluation of the postulated pipe breaks identified in items 1, 2, 3 and 5 above and will update the evaluation as conclusions are reached.
5 h
T936 042
I A!!! L 1.
[ff ECIS Of PIPE ilHf AK 0U15101 C0!11Alllllfill Affected Hitigating Affected Safe Adequacy of Zone Pipe Break System Shutdown System Protection Remarks Crib llouse SWS (MELB)*
ilone DGCW pumps Adequate.
SWS leak (822 r,pm) envelopes all HELB's in upper crib W
house.
SWS function is not lost.
trJ SWS may flood DGCW ptmps in lower crib house (see following remarks).
O CW (MElu)
None OGWS pumps Adequate. CW leak envelopes all HELB's in lower crib house. DGCW U
pumps (#2, #3, and #2/3) would be flooded, but these are submersible pumps. CW system function and offsite power would not be lost since SWS pumps in upper crib house would not be
,a flooded. A control room flood alarm is activated at 3' level in lower crib house.
Heactor Build.
Fire system (MElB)
None lione Adequate.
Floor drains are adequate (613')
to remove this leakage (approx. 100 gpm).
Fire system HELil envelopes other HELB's in this zone, e.g.
demin, water, y Heactor fluild.
Isolation Condenser None lione Adequate. Pipe whip of isolation T
(509')
Steam and Condensate condenser steam or condensate lines W
(llELB) 4 may damage both divisions of ESS N
cable trays.
These cable trays contain control cable for the isola-tion condenser constant vent valves which are located on the 'ill'"
level
'T (References 9).
lhese valves fail closed on loss of electrical signal.
W ie~iast page of l'able 1 for list of abbreviations.
S
IABLE 1.
(Continued)
Affected Hitigating Affected Safe Adequacy of lone Pipe Break System Shutdown System Protection Remarks Reactor Build fire system and lione lione Adequate.
Floor drains and stairs on (b89') (Cont.)
clean and contaminated north side of zone and drains, stairs, Demin. water (MELB) and open hatch (20' x 20') on south side of zone provide adequate drainage for HELB's.
51.C5 (llf10) flone lione Adequate. A StCS HElH outsido containment would result in the containment isolation check valve inside containment seating with reactor system pressure.
This would isolate the flow path from the reactor recirculation system to the pipe break.
- Reactor Build.
SLCS (liEto) lione lione Adequate.
See above remarks.
(570')
Isolation conde..-
Isolation conden-480V Switch gear Potentially inadequate. An steam & condenser ser steam isola-Div. I.
isolation condenser (IC) steam lines (llELB) tion valve or condensate line break may damage the IC steam line isolation valve and a failure of the IC steam line isolation valve inside containment would result in a LOCA cutside containment. Also, the IC steam or condensate line break may f
dama0e the Div. I 480V switchgear; however, this is acceptable because c4 w
the Div. 11 480V switchgear would c2 u
be available.
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TABLE 1.
(Continued)
Altected filtigating Affected Safe Adeqisacy of lone Pipe fireak System Shutdown System Protection Remarks Reactor fluild.
Fire system tione lione Adequate. State, floor drains, (S70') (cont.)
(llELB) anti open hatches provide adequate drainage to prevent equipment flooding.
Reactor Build.
RWCU (llElB)
RWCU isolation Div. II Potentially inadequate. Dicak may valves Cable fray valves damage cable tray with control &
power for RWCU valves 1201-2, or the valve itself.
This may pravent closure of the valve.
CR0 Supply line lione Div. II.
Adequate. Analyzed in Reference 6.
(llELB)
Cable Tray Isolation Condenser tione Div. II Adequate. Only one division of Vent (llELB)
Cable Tray ESS cable trays in afrected.
These trays were analyzed in Reference 6.
RDCCW (NELB) flone 4KV switchgear Potentially inadequate.
Spray from (both division)
ROCCW leak could inpinge on either 4KV switchgear 23-1 or 24-1 and flood both 23-1 and 24-1 to level of S" causing loss of both ESS divisions.
Reactor Build.
CRD return ilone lione Adequate. Sufficient distance (S45') (Cont.)
line (llELB) separates line from Div. 11 cable trays overhead.
Reactor Guild.
RBCCW or lione Reactor protection Adequate. Spray from fl[Ill af fects (545')-(Cont.)
SWS (flElB)
(RPS) instrument only one of two redundant racks.
U, racks 2202-5 or 2202-6.
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lABLE 1.
(Continued)
Affected Hitigating Affected Safe Adequacy of lone Pipe Break System Shutdown System Protection Remarks (508') would still be below the vent opening at the top of the CCSW pump 28 & 2C waterti ht 0
enclosure (511')
lurbine Unild.
Condensate Dooster tione CCSW pumps Adequate.
A sin 0 e break cmid l
(495')
pump discharge (llElin) affect Div. I power cables to pumps 2C & 20 or Div. 11 cables to pimips 2C & 20.
In either case the remaining pumps could provide the safe shutdown function.
CR0 water pumps None CCSW pumps Adequate.
See remarks above for and discharge lines CCSW pump power cables.
(llElB)
Various HELD systems None CCSW pumps Adequate. See remarks for HElB's on Turbine Build 469' level. HELB's on 495' level would drain to 469' level.
Turbine Build.
Main feed (llELD)
Emer0ency AC Emergency AC Adequate. A main feed line break in (517')
power power the turbine building could damage the diesel generator #2 bus duct.
Current pipe break criteria would assume loss of of fsite power (cause by turbine trip) and single active failure leading to loss of emergency
_g AC power.
Reliability ut offsite g
power at Dresden provides assurance d
that emergency AC power would not p
be lost.
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References a
1.
NRC letter, A. Giambusso to CE Co., dated Decemoer 18, 1972.
2.
Analysis of Effects of Pipe Break Outside Containment, Dresden Station Unit 2 & 3 Special Report No. 37, dated Januar/ 23, 1974.
3.
Supplement 1 to Dresden Station Unit 2 & 3 Special Report No. 37, dated March 22, 1974.
4 Revision 1 to Dresden Station Unit 2 & 3 Special Report No. 37, dated Februarj 18, 1975.
5.
NRC letter to CE Co., dated May 12, 1976, transmitting License Amendment No. 16 and 14 to License Nos. OPR-19 and 25.
5.
Accendix E to Dresden Station Unit 2 & 3 Special Report No. 37 dated November 17, 1977.
7.
CE Co. letter J. Abel to 0. Ziemann dated August 23, 1973 concerning Floodir.g of Critical Equipment at Oresden Units 1, 2, and 3.
3.
SEP Review of Safe Shutdown Systems for the Dresden Unit 2 iuclear Power Plant.
9.
CE Co letter M. Turcak to G. Lear dated June 5, 1973 transmitting Drescen Station Units 2 and 3 Fire Protection Safe Shutdown Analysis.
10.
Brancn Technical Position MES 3-1, appended to Standard Review Plan 3.5.2.
11.
Branch Tecnnical Position ASB 3-1, appenced to Standard Review Plan 3.6.1.
i 1936 051
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