ML19319D453
| ML19319D453 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 09/20/1973 |
| From: | Rodgers J FLORIDA POWER CORP. |
| To: | Anthony Giambusso US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 8003170583 | |
| Download: ML19319D453 (19) | |
Text
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AEC DISTRIBUTION FOR PART 50 DOCKET MATERIAL (TEMPORARY FORM)
CONTROL NO:
7329 yIng: k;-;
FROM:
DATE OF DOC DATE REC'D LTR MEMO RPT OTHER Florida Power Corporation St. Petersburgh, Fla. 33733 Mr. J.T. Rodgers 9-20-73 10-1-73 X
TO:
ORIG CC OTHER SENT AEC PDR Yn A. Giambusso i signed SENT LOCAL PDR XXX CLASS UNCLASS FROP INFO INPUT NO C'tS REC'D DOCKET NO:
XXX 60 50-302 DESCRIPTION:
ENCLOSURES:
Ltr trans the following.......in response to Report: Effects of High Energy Piping our 12-15-72 ler.....
System Breaks Outside Reactor Bldg.
Summary answers to the (21) questions contained in AEC's 12-15-72 ler.
(60cysenenclrec'd)
PLANT NAME:
Crystal River Nuc. Generating gy g
Pl nr FOR ACTION /INFORMATION I 10-2-R3 JB BUTLER (L)
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y Mr. A. Giambusso
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Deputy Director for Reactor Projects
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7 Washington, DC 20545 IN RE: FLORIDA POWER CORPORATION CRYSTAL RIVER NUCLEAR GENERATING PLANT DOCKET N0. 50-302
Dear Mr. Giambusso:
Enclosed are sixty (60) copies of a report entitled Effects of High Enerqy Piping System Breaks Outside Reactor Building, for the Crystal River Unit 3 Nuclear Generating Plant. This report was generated for Florida Power Corporation by Gilbert Associates Incorporated and is in response to your letter of December 15, 1972.
Also enclosed to facilitate the staff's review are sixty (60) copies of summary answers to the twenty-one (21) questions contained in your December 15, 1972, letter in question and answer format.
This report will be incorporated by reference in the Final Safety Analysis Report via Amendment #32 to the Crystal River Unit 3 Application for Licenses, to be filed on October 1,1973.
Please feel free to contact us if we can be of further assistance.
Very truly yours, 11 J. T. Rodgers
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Assistant Vice President JTR/ns
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General Office 3201 Tnirt;-tounn street soutn. P.O. Box 14042, st. Petersburg. Flonda 33733 813--866-5151
s SU) NARY RESPONSES TO TWENTY-ONE QUESTIONS DLR LETTER OF DECEMBER 15, 1972 Crystal River Unit 3 Florida Power Corporation October 1, 1973
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QUESTION 1 The system (or portions of systems) for which protection against pipe whip is required should be identified.
Protection from pipe whip need not be provided if any of the following conditions will exist:
(a) Both of the following piping system conditions are met :
(1) the service temperature is less than 200 F; and (2) ene design pressure is 275 psig or less; or (b) The piping is physically separated (or isolated) from structures, systems, or components important to safety by protective barriers, or restrained from whipping by plant design features, such as concrete encasement; or (c) Following a single break, the unrestrained pipe movement of either end of the ruptured pipe in any possible direction about a plastic hinge formed at the nearest pipe whip restraint cannot. impact any structure, system, or component important to safety; or (d) The internal energy level associated with the whipping pipe can be demonstrated to be insufficient to impair the safety function of any structure, system, or component to an unacceptable level.
RESPONSE 1 A tabulation of all systems required to bring the plant to a safe shutdown following a postulated high energy pipe break outside containment is given in Section 5.2 of Reference 1.
Those systems which are in or near postu-lated break areas are identified.
It should be noted that there are no systems requiring protection in the Control Complex Building or in the Turbine Building.
QUESTION 2 Design basis break locations should be selected in accordance with the following pipe whip protection criteria; however, where pipes carrying high energy fluid are routed in the vicinity of structures and systems necessary for safe shutdown of the nuclear plant, supplemental protection of those structures and systems shall be provided to cope with the environmental effects (including the effects of jet impingement) of a single postulated open crack at the most adverse location (s) with regard te those essential structures and systems, the length of the crack being chosen not to exceed the critical crack size. The critical crack size is taken to be 1/2 the pipe diameter in length and 1/2 the wall thickness in width. ~~
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.s QUESTION 2 (Cont'd)
The criteria used to determine the design basis piping break locations in the piping system should be equivalent to the following:
(a) ASME Section III Code Class I piping breaks should be postulated to occur at the following locations in each piping run or branch run :
(1) the terminal ends; (2) any intermediate locations between terminal ends where the primary plus secondary stress intensities Sm (circumferential or longitu-dinal) derived on an elastically calculated basis under the loadings associated with one-half safe shutdown earthquake and operational plant conditions exceeds 2.0 S for ferritic steel, m
and 2.4 S for austenitic steel; m
(3) 'any intermediate locations between terminal ends where the cumulative usage factor (U) derived from the piping fatigue analysis and based on all normal, upset, and testing plant conditions exceeds 0.1; and (4) at intermediate locations in addition to those determined by (1) and (2) above, selected on a reasonable basis as necessary to provide protection. As a minimum, there should be two intermediate locations for each piping run or branch run.
(b) ASME Section III Code Class 2 and 3 piping breaks should be postulated to occur at the following locations in each piping run or branch run:
(1) the terminal ends; (2) any intermediate locations between terminal ends where either the circumferential or longitudinal stresses derived on an elastically calculated basis under the loadings associated with seismic events and operational plant conditions exceed 0.8 (Sh + S ) or the expansion stresses exceed 0.8 S ; and A
A (3) intermediate locations in addition to these determined by (2) above, selected on reasonable basis as necessary to provide protection. As a minimum, there should be two intermediate locations for each piping run or branch run.
RESPONSE 2 Structures and systems necessary for safe shutdown are insured to be func-tional in the event of a crack break anywhere along a high energy pipe line by investigating the effects of an open crack at adverse locations.
(a) The guidelines presented in the above question are not applicable to the design of the Crystal River Unit Three plant. The piping is designed to ANSI Code for Power Piping B 31.1.0-1967 and not,
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RESPONSE 2 (Cont'd) to ASME Section III which was issued in 1970. Ihe guidelines which are utilized are those presented in Question 2.b and in Section 4.1 of Reference 1.
(b) The guidelines presented in Question 2.b are applicable to Crystal River Unit Three and are followed to determine break locations.
(1) Terminal ends of each branch and piping run are considered as break locations regardless of the stress levels.
(2)
Intermediate locations where the combination of stresses exceed those in Question 2.b.2 are considered as break locations.
(3) As a minimum two intermediate locations for each piping run or branch run are selected at locations of highest stress. Additional breaks are postulated at locations where the consequences of a break can not be tolerated and whose stresses are not distinctly less than the highest stresses.
QUESTION 3 The criteria used to determine the pipe break orientation at the break locations as specified under 2 above should be equivalent to the following:
(a) Longitudinal breaks in piping runs and branch runs, 4 inches nominal pipe size and larger, and/or (b) Circumferential breaks in piping runs and branch runs exceeding 1 inch nominal pipe size.
RESPONFE 3 The guidelines presented in Question 3 are strictly adhered to including the qualifying footnotes.
This guideline appears in Sections 3.0 and 4.1 of Reference 1 and is adhered to explicity.
It is rephrased slightly, bo.
its intent is not altered.
QUESTION 4 A summary should be provided of the dynamic analyses applicable to the design of Category I piping and associated supports 'diich determine the resulting loadings as a result of a postulated pipe break including:
(a) The locations and number of design basis breaks on which the dynamic.
analyses are based.
(b) The postulated rupture orientation, such as circumferential and/or longitudinal break (s), for each postulated design basis break location.... _.
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QUESTION 4 (Cont'd)
(c) A description of the forcing functions used for the pipe whip dynamic analyses including the direction, rise time, magnitude, duration and initial conditions that adequately represent the jet stream dynamics and the sys**m pressure difference.
(d) Diagrams of mathematical models used for the dynamic analysis.
(e) A summary of the analyses which demonstrates that unrestrained motion of ruptured lines will not damage to an unacceptable degree, structure, systems, or components important to safety, such as the control room.
RESPONSE 4 (a) The breaks postulated are too numerous to list here.
Each postulated break is presented in one of the Tables and Figures in Reference 1 as listed beloc:
System Table Figure Main Steam 4.1.2 4.1.1 Feedwater 4.1.3 4.1.3 4.1.4 Decay Heat 4.1.5 4.1.4 (b) The break locations which are included in the Tables and Figures above are considered for both circumferential and longitudinal breaks.
(c) The loading conditions developed on a pipe run or branch run are determined using the pipe fluid conditions associated with maximum normal operating conditions of the system. A computer program solution Flash II* is used to determine time dependent flow rates resulting in rapid blcwdown of a linked-multi-volume pressure system containing high temperature fluids. This time dependent information in turn is used to calculate thrust forces and jet impingement forces that are reacted out on piping, adjacent structures and equipment.
A more detailed description of these calculation methods can be found in sections 4.3 and 4.4 of Reference 1.
(d) The analytical procedure used for unrestrained pipe whip is a non-linear dynamic time-history analysis of lumped mass models of the whipping pipe and the structural element subject to impact. A complete example of int unrestrained pipe whip problem is presented in detail in Section 4.5.2 of Reference 1.
- Flash II - A Fortran IV Program for the Digital Simulation of a multinode Reactor plant during loss of coolant - RAPD - IM - 666 April '67, J. A. Redfield, l
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RESPONSE 4 (Cont'd)
(e) The areas of. concern for postulated pipe rupture outside containment are in the Intermediate Building and Compartments DH1, Dd2, and A14 in the Auxiliary Building.
The Control Room is protected by remote location.
In area A14 of the Auxiliary Building, all restraints required to protect required systems are provided and are described in Section 6.1.13 of Reference 1.
In the Intermediate Building, rupture restraints (presented in Sections 6.1.1 thru 6.1.12 of Reference 1) are provided for 24" Main Steam lines, and for 18", 10", and 6" Feedwater lines for all postu-lated rupture locations that could result in unacceptable damage.
Consequently, those lines are considered for restrained pipe break rather than unrestrained pipe whip.
The remaining unrestrained 6" Feedwater lines are investigated for unrestrained whip and do not produce unacceptable consequences.
The Auxiliary Building areas DH1 and DH2 contain redundant Decay Heat systems that are separated by a common 2'-6" thick reinforced concrete wall with sufficient strength to prevent a rupture in one compartment from damaging the system in the. adjacent compartment. Analyses and results presented in Sections 4.5.2 and 6.2 of Reference 1 justify the protection of one Decay Heat System from the other by the physical separation provided by the common wall.
QUESTION 5 A description should be provided of the measures, as applicable, to protect against pipe whip, blowdown jet and reactive forces including:
(a) Pipe restraint design to prevent pipe whip impact; (b) Protective provisions for structures, systems, and components required for safety against pipe whip and blowdown jet and reactive forces; (c) Separation of redundant features:
(d) Provisions to separate physically piping and other components of redundant features; and (e) A description of the typical pipe whip restraints and a summary of number and location of all restraints La each system.
RESPONSE 5 (a) Pipe whip restraints employed as energy absorbers are designed and subjected to a nonli.near time-history dynamic analysis as described in Section 4.5.1 cf Reference 1.
Included in the dynamic analysis are loading combina tions which include rupture thrusts, jet blowdown transients and applicable reactive forces as described in Section 4.5.1.6.,'
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RESPONSE 5 (Cont'd) -
(b) Ruptures are postulated at various locations in the Main Steam, Feedwater and Decay Heat lines based on the stress levels computed for the normal operating condition in accordance with the AEC criteria for the effects of a piping system break outside Containment.
Rupture locations are identified in Figures 4.1.1, 4.1.3, and 4.1.4
'of Reference 1.
At eafi postulated break location, both upstream and downstream, loading due to circumferential breaks and critical orthogonal orientations for longitudinal breaks are considered. Their worst effect on adjacent structural components and safety related systems are studied. Protection is provided as required to preserve safe shutdown capability.
Section 6.4 of Reference 1 gives details where jet shields are provided for required equipment, piping, valves, electrical cable and raceways which are in the path of jets resulting from postulated pipe breaks.
Section 6.5 of Reference 1 refers to the provision of openings in walls c f the Intermediate Building to reduce the potential flooding problem in the case of postulated Feedwat'r breaks in the Intermediate Building above Elevation 119'.
Floor openings are protected with metal enclosures extending above the predicted water levels to prevent flooding below.
(c) and
( d) Pedundant Engineered Safeguard purps and t exchangers that are in buildings with high energy piping are located in separate cubicles.
The Emergency Feedwater pumps are located in a common compartment, but there are no high energy pipes in this compartment.
The valves, piping, and support equipment required for Engineered Safeguard and Emergency Feedwater systems are sufficiently separated so that a postulated break cannot cause a loss of required redundancy.
(S ee Reference 2, Section 6.0).
(e) Sections 6.1.1 thru 6.1.13 of Reference 1, identify the location and number of pipe whip restraints necessary to preclude undesirable consequences of postulated pipe breaks.
Typical configurations of restraints are described in those sections.
QUESTION 6 The procedures that will be used to evaluate the structural adequacy of 2
Category I structures and to design new seismic Category I structures should be provided including:
(a) The method of evaluating stresses, e.g., the working stress method and/or the ultimate strength method that will be used; (b) The allowable design stresses and/or strains; and (c) 1he load factors and the load combirc tions.. _ _.
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m RESP 0 HSE 6 (a) The Ultimate Strength Design (USD) Method as set forth in ACI 318-71 is used to determine the strength of reinforced concrete structures.
(b) Maximum values as controlled by section strength are calculated in accordance with ACI 318-71.
(c) The capacities of existing designs are investigated using the load combinations as reco= mended in Document (B) - Structural Design Criteria for Category I Structures Outside the Containment, issued by the Structural Engineering Branch of Directorate of Licensing, AEC.
Section 4.9.3 of Reference 1 presents the details of the investigation performed.
QUESTION 7 The structural design loads, including the pressure and temperature transients, the dead, live and equipment loads; and the pipe and equipment static, thermal, and dynamic reactions should be provided.
RESPONSE 7 Pressurization loads are calculated using the results of the analysis for those breaks which produce maximum differential pressures on structural framing systems. Sections 4.5.1 and 4.5.3 of Reference 1 illustrate the manner in which pressure differentials are considered as design loads.
Dead load is taken as the weight of the structural framing system, including any equipment load permanently mounted. Full live load is considered or completely ignored for arriving at the capacity of the structural system depending on whether it adds to or decreases the loading effect on a given element in combination with loads due to pipe break. Sections 4.5.1 through 4.5.4 cover all loadings and load combinations considered.
The reactions from piping and equipment static, thermal and dynamic (seismic) loadings are considered in all restraint dynamic models where their effect adds to the effect of the rupture thr.:st.
Refer to Section 4.5.3.2 of Reference 1 for the load combinations considered.
QUESTION 8 Seismic Category I structural elements such as floors, interior walls, exterior walls, building penetrations and the buildings as a whole should be analyzed for eventual reversal of loads due to tLe postulated accident.
RESPONSE 8 All structural concrete elements are typically reinforced in both faces and can sustain any expected load reversal.,
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QUESTION 9 If new openings are to be provided in existing structures, the capabilities of the modified structures to carry the design loads should be demonstrated.
RESPONSE 9 Additional vent areas are provided to reduce peak differential pressures subsequent to postulated pipe break.
In all cases, the modified structural elements are capable of sustaining the design loads.
QUESTION 10 Verification that failure of any structure, including nonseismic Category I structures, caused by the accident, will not cause failure of any other structure in a manner to adversely affect:
(a) Mitigation of the consequences of the accidents; and (b) Capability to bring the unit (s) to a cold shutdown condition.
RESFONSE 10 Pipe whip resulting from postulated breaks in the Turbine Building (only non-safety class structure of concern) will not damage adjacent safety class structure to such an extent as to prevent:
(a) Mitigation of consequences or (b) Capability to achieve a cold shutdown condition.
QUESTION 11 Verification that rupture of a pipe carrying high energy fluid will not directly or indirectly result in:
(a) Loss of required redundancy in any portion of the protection system
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(as defined in IEEE-279), Class IE electric system (as defined in IEEE-308), engineered safety feature equipment, cable penetrations, or their interconnecting cables required to mitigate dhe consequences of that accident and place the reactor (s) in a cold shutdown condition; or (b)
Environmentally induced failures caused by a leak or rupture of the pipe which would not of itself result in protective action but does disable protection functions.
In this regard, a loss of redundancy is permitted but a loss of function is not permitted. For such situations plant shutdown is required.
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RESPONSE 11 The objective of the pipe rupture review is to assure that no postnisted rupture can lead to a situation where the reactor cannot be safely shutdown. To accomplish this objective, all components required to bring the reactor to a cold shutdown are tabulated (see Section 5.0 of Reference 1) and the effects of postulated pipe ruptures have been investigated. Required components and related circuits are either located a sufficient distance from high energy lines or are protected such thct a break or crack in a line does not impair the operation of the system.
QUESTION 12 Assurance should be provided that the control room will be habitable and its equipment functional af ter a steam line or feedwater line break or that the capability for shutdown and cooldown of the unit (s) will be available in another habitable area.
RESPONSE 12 The Control Complex does not experience adverse effects from postulated pipe breaks. Outside air to the Control Complex is ducted to the Control Room from a remote roof intake terminal and is not adversely affected by postulated pipe breaks. Steam leakage from postulated break locations to the Control Complex is minimized as it must travel through doors before entering the Control Complex.
(See Reference 1, Section 6.6.1 for more detail).
QUESTION 13 Environmental qualification should be demonstrated by test for that electrical equipment required to function in the steam-air environment resulting from a high energy fluid line break. The information required for our review should include the following:
(a) Identification of all electrical equipment necessary to meet requirements of 11 above.
The time after the accident in which they are required to operate should be given.
(b) The test conditions and the results of tese data showing that the systems will perform their intended function in tie environment resulting from the postulated accident and time interval of the accident. Environmental conditions used for the tests should be sc lected from a co servative evaluation of accident conditions.
(c) The results of a study of steam systems ideatifying locations where barriers will be required to prevaat steam jet impingement from disabling a protection system.
The design criteria for the barriers should be stated and the capability of the equipment to survive within the protected environment should be described..-
QUESTION 13 (Cont'd)
(d) An evaluation of the capability for safety related electrical equipment in the control room to function in the environment that may exist fellowing a pipe break accident should be provided. Environmental ccnditions used for the evaluation should be selected from conservative ca.lculations of accident conditions.
(e) An evaluation to assure that the onsite power distribution system and onsite sources (diesels and batteries) will remain operable throughout the evert.
RESPONSE 13 Equipnent required for safe reactor shutdown following a high energy line break is determined. Manufacturer investigations have been conducted to assess equipment integrity in a temperature, pressure, and humidity environ-ment as determined by c 3nservative calculations. Where required, environmental protection is provided.
(See Sections 5,2, 6.4, and 6.6 of Reference 1 for a more detailed analysis).
QUESTION 14 Design diagr..s and drawings of the steam and feedwater lines including branch lines showing the routing from containment to the turbine building should be provided.
The drawings should show elevationt and include the locatior
- lative to the piping runs of safety related equipment including ventila6_.a equipment, intakes, and ducts.
RESPONSE 14 Isometrics are included in Reference 1 which show the Main Steam lines (Figure 4.1.1) and the two Feedwater lines (Figure 4.1.2 and 4.1.3), as well as the Emergency Feedwater System as they run from the Reactor Building, through the Intermediate Building, and into the Turbine Building with the elevations of each line as shown.
These isometrics show the relative location of safety related equipment.
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Ihe location of safety related structures has been previous.y identified in Figures 1-5 and 1-7 of Section 1, Reference 2.
Composites of the Decay Heat Lines in the Auxiliary Building are presented (Figure 4.1.4) in Reference 1.
This figure indicates the relative location of safety related equipment.
QUESTION 15 A discussion should be provided of the potential for flooding of safety related equipment in the event of failure of a feedwater line or any other line carrying high energy fluid. !
.m RESPONSE 15 The effect of flooding in the event of the failure of a pipe line con-veying high energy fluids is investigated for the Intermediate and Auxiliary Buildings. The maximum flood level is established for both buildings and necessary structural provisions are made to prevent the flooding of safety related equipment in these buildings. Flooding caused by the rupture of a pipe carrying high energy fluids is dis-cussed in Reference 1, Sections 4.3.3. and 6.5.
QUESTION 16 A description should be provided of the quality control and inspection programs that will be required or have been utilized for piping systems outside containment.
RESPONSE 16 Quality Control and Inspection of Piping (a) The contractor is furnishing the engineer with a q. ality control manual which covers the overall quality control methods and procedures to be used for piping and fittings for nuclear systers.
The quality control methods and procedures includes how the contractor is controlling such items as :
(1) Control of Raw Materials - Including the procedures for the receiving, inspection, identification and certification of incoming raw naterials.
(2) Control of the Quality of Purchased Parts - Including inspection by the manufacturer and recent inspection and testing by the contractor.
(3) Control of Fabrication Processes - Including qualification of fabrication processes, control and checking of tools and fixtures, heat treatment and cleanliness procedures.
(4) Control of Inspection and Test Equipment - Including calibration standards and recalibration frequency.
(5) Control of Packaging and Shipping - Including final inspection releases, inspection of packages, and maintenance of cleanliness.
(6) Control of Changes to Documents Affecting Quality - Including drawings, specifications, procedures and other related documents.
(7) Material Identification - Including means used to positively identify all material and to indentify the inspection status of the material.,
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RESPONSE 16 (Cont'd)
(8) Disposition of Non-Conforming Items - Including repair, rework, and retest procedures and the steps taken to assure that non-conforming material is positively identified and not inadvertently used.
(9) Control and Storage of Inspection and Test Records - Including the listing of inspection and test records which the contractor turns over to the owner at the completion of the work or those which are maintained by the contractor for the life of the plant (i.e., 40 years).
This portion of the quality control plan or procedure describes what steps will be taken by the contractor to ensure that the records he retains will be maintained and readily retrievable for 40 years.
(b) The owner provides the facilities and equipment necessary to receive and inspen t pf ning and fittings at the jcb site.
(c) All compen
.pected for cleanliness in accordance with standards set forth aess requirements.
This cleanliness is also maintained L.
. stage, QUESTION 17 If leak detection equipment is to be used in the proposed modifications, a discussion of its capabilities should be provided.
RESPONSE 17 Leak detection capabilities are present in current system designs. A rupture of a Main Steam line is detected by a pressure switch matrix which activates an alarm and closes the Feedwater valves. A Feedwater rupture is indicated to the operator by low pressure in the Feedwater lines showing up on multiple pressure indicators on the main control board.
In addition, an alarm sounds on low level in the steam generator. A makeup and purification system rupture is indicated by an alarm signifying low level in the makeup tank, as well as a high flow alarm for the makeup line itself. A break in the suction side of the Decay Heat pumps is indicated by a low flow alarm in the Decay Heat System. There is also a high flow alarm and flow indication for low pressure injection. Other possible breaks are indicated by a low reactor coolant pressure alarm or a high reactor coolant seal flow alarm. These alarms and indications are available to the Control Room operator in the event of a rupture in a high energy pipe.
QUESTION 18 A summary should be provided of the emergency procedures that would be followed af ter a pipe break accident, including the automatic and manual operations required to place the reactor unit (s) in a cold shutdown condition. The estimated times following the accident for all equipment and personnel operational actions should be included in the procedure summary.
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1 RESPONSE 18 The Main Steam pipe break emergency procedures below are general in nature since it is deemed appropriate to allow for assessment of the incident prior to ultimately bringing the reactor to a cold shutdown.
TNe times for equipment and personnel actions for the most severe accident conditions have been calculatea and are presented in Section 14.2.2.1 of Referenca 2.
1.
Symptoms 1.1 Rapid decrease of steam pressure, OTSG level, and/or unit electrical load dependent upon integrated control system mode of operation.
- 1. 2 Rapid decrease in pressurizer level, RC pressure, and RC cold leg temperature in the affected loop.
- 1. 3 Reactor trip initiated by high flux (negative moderator coefficient) or low coolant pressure.
- 1. 4 For a rupture inside the reactor ouilding:
Increasing building pressure and temperature.
1.5 For a rupture outside the reactor building: Noise may be heard in the Control Room or reports made from personnel outside the Control Room.
2.
Immediate Action 2.1 Verify the reactor / turbine have tripped. If not, trip manually.
2.2 Verify the affected OTSG feedwater control and block valves are closed.
If not, close manually.
2.3 Initiate high pressure injection if preasurizer level falls below recorder scale, reactor pressure continua.s to decrease, or flux is rising.
2.4 Stop one RC pump in each loop.
3.
Follow-Up Action 3.1 Select the unaffected loop for header pressure control.
3.2 If pressure-temperature relationship is such that it would cause cavitation, stop all RC pumps. Refer to the attached curve for pressure-temperature limits.
- 3. 3 Verify minimum level control is established in the unaffected OTSG. _
RESPONSE 18 (Cont'd) 3.4 Place control switches for the Main Feedwater block valve, low load Feedwater block valve, startup Feedwater isolation valve, and Emergency Feedwater isolation velve to the affected OTSG in the " Closed" position.
3.5 Continta normal plant cooldown.
- 3. 6 Using only the unaffected OTSG, close the affe :ted OTSG turbine Feedwater pump and steam bypass root valve MSV-53 or MSV-54.
- 3. 7 Close auxiliary steam supply valve on the affected loop, MSV-55 or MS V-56.
- 3. 8 When the unaffected steam line pressure reaches 725 psig, place steam line rupture matrix switches "A" and "B" in the " Bypass" pcsition.
3.9 When pressurizer level is restored, the emergency injection may be bypassed and normal makeup and baron addition established.
3.10 Locate and isolate leak if possible.
QUESTION 19 A description should be provided of the seismic and quality classification of the high energy fluid piping systems including the steam and feedwater piping that run near structures, systems, or ce=ponents important to safety.
RESPONSE 19 High energy piping, assumed for postulated failure in the Intermediate and Auxiliary Buildings, is investigated to assure that a pipe failure does not 4
impair the operation of equipment for safe operation. Steam and Feedwater piping is included in the high energy fluid piping system located near structures and/or components. Reference 1 includes Table 4.1.1 showing these lines with postulated failures in buildings tontaining safety equip-ment.
The table lists the pipe runs.with the applicable seismic and pipe classifications.
QUESTION 20 A description should be provided of the assumptions, methods, and results of analyses, including steam generator blowoown, used to calculate the pressure and temperature transients in compartments, pipe tunnels, intermediate buildings, and the turbine building followiag a pipe rupture in these areas.
The equipment assumed to function in the analyses should be identified and the capability of systems required to function to meet a single active component failure should be described.
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RESPONSE 20 Fluid systems with postulated pipe ruptures are evaluated v th multi-node computer si.ulations. The FLASH II computer code is used to determine system response and compartment pressures. Output flow rates and enthalpies are then used as input data for the CONIEMPT (3) computer code to determine the long term temperature history within the Intermediate Building. A more complete description of these analyses (including the equipment assured to function) is contained in Section 4.2 of Reference 1.
QUESTION 21 A description should be provided of the methods or analyses performed to demonstrate that there will be no adverse effects on the primary and/or secondary containment structures due to a pipe rupture outside these struetures.
RESPONSE 21 l
The existing containment penetrations are designed for breaks at the terminal ends just inside and outside the containment. Therefore, additional analyses are not required. All postulated breaks outside containment that could result in the secondary containment being impacted are restrained.
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1 REFERENCES 1.
Gilbert Associates, Inc., " Effects of High Energy Piping System Breaks Outside the Reactor Building", Crystal River Unit 3, October 1,1973.
t 2.
Florida Power Corporation, " Final Safety Analysis Report, Crystal River i
Unit 3".
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