ML20070M105
| ML20070M105 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 01/11/1983 |
| From: | Longenecker J ENERGY, DEPT. OF, CLINCH RIVER BREEDER REACTOR PLANT |
| To: | Check P Office of Nuclear Reactor Regulation |
| References | |
| HQ:S:83:182, NUDOCS 8301120279 | |
| Download: ML20070M105 (17) | |
Text
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O Department of Energy Washington, D.C. 20545 Docket No. 50-537 HQ:S:83:182 JAN 111983 Mr. Paul S. Check, Director CRBR Program Office Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Check:
ADDITIONAL INFORMATION ON MECHANICAL ENGINEERING BRANCH (MEB) ITEMS 4, 26, 64, 68, 69 AND 72
References:
Letter HQ:S:82:143, J. R. Longenecker to P. S. Check,
" Meeting Summary: November 22-24, 1982, Mechanical Engineering Branch / Clinch River Breeder Reactor Plant Meeting," dated December 14, 1982 Letter HQ:S:82:128, J. R. Longenecker to P. S. Check,
" Additional Information Resulting from the September 8-9, 1982, MEB/CRBRP Meeting," dated November 23, 1982 Letter HQ:S:83:181, J. R. Longenecker to P. S. Check,
" Additional Information on Steam Generator Non-Destructive Examine.tions (NDE) and Reactor Vessel (RV)
Core Support Cone Structural Integrity," dated January 11, 1983 Enclosed is additional information concerning MEB items 4, 26, 64, 68, 69, and 72 from Reference 1.
The response to MEB question 4, previously submitted in Reference 2, has been revised in response to comments from EG&G, Idaho Falls, and is enclosed.
Responses to MEB questions 26, 64, 68, and 69 are enclosed to complete actions previously committed to in Reference 2.
A response to MEB item 72 has been provided under separate cover in Reference 3.
The enclosed modifications to the Preliminary Safety Analysis Report will be included in a future amendment.
I B301120279 830111 v
PDR ADOCK 05000537 A
2 Questions regarding the enclosure may be addressed to Mr. D. Robinson (FTS 626-6098), Mr. D. Hornstra (FTS 626-6110), or Mr. D. Edmonds (FTS 626-6157) of the Project Office Oak Ridge staff.
Sincerely, m
J8bnR.Longen
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ker Acting Director, Office of Breeder Demonstration Projects Office of Nuclear Energy Enclosure cc: Service List Standard Distribution Licensing Distribution I
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l Ikm 4 NRC Question:
Do any mechanical systems and components correspond to Quality Group D requirements as contained in Regulatory i
Guide 1.267 (Item 2 pg. 3.2.2-3)
Response
A separate category of equipment equivalent to Quality Group D has not been specified for CRBRP.
However, any CRBRP equipment which are equivalent to equipment covered in Quality Group D of Regulatory Guide 1.26, have quality requirements corresponding to Reg. Guide 1.26.
This is i
accomplished by the CRBRP Quality Assurance Program, as discussed in PSAR Chapter 17, Appendix A, Section 0.3 and the imposition of appropriate industry standards.
[
A listing of industry standards being applied to non-safety
- related equipment is provided in PSAR Table 3 2-4.
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1 MEB Item 26:
The NRC expressed concern at the November 22-24, 1982, meeting at Waltz Mill that no specific criterion was identified for the evaluation of flow f
induced vibration (FIV) test results.
It was suggested that s limiting value of 50 percent of the Code endurance limit at 106 cycles would be appropriate.
Response
-The information presented at the same meeting for MEB Item 64 indicated that for load controlled conditionsthehighcycleloadgngsforCRBRPto 1010 cycles.
require evaluation at about 10 Since the endurance limit decreases by 6 to approximately a factor of 2 in going from 10 109 cycles, the CRBRP procedures are equivalent to the suggested limiting value.
In any event, the PIV results must be within the component design limits or corrective action will be required as noted in PSAR Section 3.9.1.
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MEB Item 64, Part 1:
The Applicant used modified creep-fatigue damage rules for non-Code stamped austentic stainless steel components.
The modified rules assumed that in compressive hold, the creep damage is only 20% as damaging as that caused by the same sustained stress in tension.
Other studies indicate that this may not be conservative and so the Applicant should justify the 20% factor.
Response
Technical justification for the. modified compressive hold rules was provided at the November 22-24, 1982, meeting with the NRC MEB at Waltz Mill, Pennsylvania (see attachments to DOE letter HQ:S:82:143 dated December 14, 1982).
The enclosed changes to PSAR Section 4.2 document the procedure used to perform creep damage calculations.
Appendix 5.2A is deleted because.it duplicates the description of the modified rules given in Section 4.2.
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J-RE.Pu u wm hseav 4.2-181 I Subjict to the above limitations, the creep damage may be calculated [
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g
- accordance with F9-4T and Code Case 1592 as modified. The modi g
g) fiettion is to use a peak stress to rupture design curve based n
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the ess to rupture design curve in Code Case 1592 adjuste or i
the inf nce of a non-linear stress state caused by the
'sence f
of a geome
'c stress concentration as with the follow *
/
Step 1 - Determ the smooth speciment streis ture stren*gth curve by, ts of the same material the temperature of f
interest.
i Step 2 - Determine the stre.
ruptur strength curve with the presencea cimens of the same heat of mate of the geometric stre ncentrations under the same conditions in (1) wi with the same his ies.
lytically detemine the peak stress relative the net st s thus defining the stress rupture stre h in terms of "pe - stress" vs. time to g
rupture./
Step 3 - RatVCode Case 1592 stress to rupture de n curve by the io of Step 1 divided by Step 2.
This mus e done for at least 3 points in time with a separation in tim f at least 9 two orders of magnitude. --In cases where the stren 5 ratio Varies with lifetime, the lesser of the value extrapo ed to the component lifetime or the experimental value for s longest duration tests shall be used.
O (8) The total creep-fatigue damage is determined by adding to the creep
- s../
damage and fatigue damage calculated in accordance with T-1411
-1412,
-1413, and- -1414 of Code Case 1592.
(9) The allow'able creep-fa.tigue damage (D') i's determined from the lesser of the values from Figure T-1420-2 of Code Case 1592 (See Figure 4.2-47a)
I and an average of test values from creep-fatigue interaction tests.
of notched specimens.
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k High Cycle Strain Controlled Fatigue Limits i
For those 304 and 316 Stainless Steel components which are outside ASME Code jurisdiction, the fagigue damage for strain controlled cyclic deformations in excess of 1 10 cycles may be evaluated using allowable strain ranges obtained from Figure 4.2-47B, provided metal temperatures do 0
not exceed 1100 F.
Fatigue life reduction factors must be applied indepen-1 dently for slow strain rates and hold times, in accordance with ASME Code 57 requiremen ts.
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4.2-181 Amend. 57 Nov. 1980 t
INSERT 4.2-181 1
(7) Subject to the above limitations, the creep damage may be calculated in accordance with F9-4T and Code Case 1592 with one of the following modifications.
(A) Use a peak stress-to-rupture design curve based upon the stress-to-rupture design curve in Code Case 1592 adjusted for the influence of a non-linear stress state caused by the presence of a gr.ometric stress concentration.
(a) Determine the smooth specimen stress rupture strength curve by tests of the same material at the temperature of interest.
(b) Determine the stress rupture strength curve with the presence of the geometric stress concentrations under the same conditions in (a) with specimens of the same heat of material with the same histories. Analytically determine the peak stress relative to the net stress thus defining the stress rupture strength in terms of " peak stress" vs time to rupture.
(c) Ratio the Code Case 1592 stress to rupture design curve by the ratio of (b) divided by (a). This must be done for at least 3 points in time with a separation in time of at least two orders of magnitude.
In cases where the strength ratio varies with lifetime, the lesser of the value extrapolated to the component lifetime or the experimental value for the longest duration tests shall be used~.
(d) Use the greater of the creep damage using this modified rule and the creep damage using the stress unaltered by
.the stress concentration and the Code Case 1592 stress-to-rupture design curve.
(B) If tests subject to the above limitations (1 through 6) show no decrease in rupture life for prototypic notch geometries, calcu-late the component creep damage neglecting the stress concentra-tion due to the notch. No reduction in damage below the damage using the stress unaltered by the stress concentration and the Code Case 1592 stress-to-rupture design curve shall be used.
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Page 1 of 1
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TMILE 5.2-l SIMWtY OF (DOE, CDOE CASES AND RDT l.
STANDARDS APftlCMILE TO CESIGN AND MANUFACTURE OF REACTOR VESSEL, 0.05URE HEAD AND GUNID VESSEL Cintura 9
's fYessure Internels Guard Ccuponent/Crlier t e fleector Yessel Boundary (as appropriatel Vessel Section 111 Addende thru Wlater Addende thru Winter Addende thru Winter Addende thre ASE Code,
'74
'74
'74 Summer '75 1974 Edition Class 1 Class 1 Class I Class 2ee ASE Code Cases 1521-l.1592-2,1593-1682,8690 1521-1 1592-4,1593-l.1594-1 0,1594-1,8595-1, 1597-4,1593-l lf elected by sup-plier 1521-1 & 1682 1596-1 (1682,1690 Optionell RDF Standards E8-18T, 2/75 E15-2NI-T, 11/74 El5-MB-T, 11/74 EI S-le-T, 11/74 Mendefory E15-2NI-T, 11/74 Amend thru 6/75 A,*end thru 4/75 Amend thre 4/76 Amend thru I/75 b
F 2-2, 8/73 F2-2, 8/73 F2-2, S/73 F2-2, S/73 1
l Amend thru 7/75 Amend thru 7/75 Amend thru 7/75 head thru 7/75 N
F3-6T, 12/74e se F9-4, 9/74 fr34T, 10/75 F3-6T, 12/74 With Amend.1/75,
F6-5T, 8/74 F6-ST, 8/74 F6-57, 8/74 Amend thru 2/75 Amend thru 2/75 Amend thru 11/75 i
F 7-3T, 11/74 F7-3T, 6/75 F7-3T, 6/75 F9-4T, 9/74 NI-lT, 3/75 F9-4, 9/74 l
N1-2, 3/75 Amend thru 7/75 afor those reactor vessel and clostre head components Internal to the pressure boundary special purpose higin cycle f atigue curves and creep demoge rules have been developed as discussed in h_r_d:a " 2Ah gg4 4g3g k1 hot l'$
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L APPEN01X 5.2A
- 9 Modif Ications to the High Temperature Design Rules or Austenitic Stainless Steel.
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Creen-tIgue Evafuatton Creep-fat ue evaluations will be performed in accorda e with the applicable criterla e opt as modified herein.
The creep-fati e damage rules of Paragraph T-1400 f Code Case 1592 consider creep damage acc ulations resulting from stresse which are clearly compressive to be ually as damaging as creep age accumulations from tensile stresses.
e damaging effects of com essive stresses in a high temperature environme are known to very con derably from one material to another. Strain contro ed fatigue test dat of austenitic stainless steels (304 and 316 SS) consiste ly point to com assive residual stresses having little or no deleterious of ct. There I also test evidence that suggests that when subjected to 'altern e hold p lods in both tension and compression that hold in compression has a allng ffect on the damage produced by the tensile hold. Based upon these ta, he creep-fatigue damage rules are modified as described in subsequen aragraphs.
The effects of the presence of st ess ncentrations on stress rupture properties are known to vary co Iderabl with the material, geometry of the stress concentration, magnitud of the st ss level, the environment, and life.
In the case of austenitic st inless steels, test data consistently points to stress concentrations havin a less severe offact on stress rupture strength
^
than predicted using the a lytical apprcaches 'of 1592 and F9-4 criteria, and in the case of 316 SST, ere.is a consistent trhqd to significant notch strengthening for some t pas of geometries, partic04arly with a service environment and life a the upper limit of those in 4be UlS. The rules of RDT F9-4T and Code Case 1 2 require comparing the peak stress to the Co'de strength which is based upon oath specimen data. They do not he adverse ef fect on stress rupture str, cognize that peak stresses may have n ength nor do they recognize that no uniform stress states may alter the strength of the material. Based pon test data, the creep damage rules are mgdtfled as described in su equent paragraphs to allow the use of a peak' stress to rupture design curve.
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Modtficatto to Creen-Fatigue Damane Rules in cases ere, in the service life of the component, all three principal stresses are clearly compressive during a hold period, the creep-f atigde evaluat on shall be modified as described herein.
If prerequisites for \\pe use of th modified rule are not met for a portion of a component's lif e, the cree -fatigue rules of T-1400 of Code Case 1592 shall be used without f ication f or that portion of the component's lif e.
The modified rule is mo d cribed in items (1) to (7), where (1) to (5) are prerequisite conditions, 58 d item (7) is a final appilcability criteria to be satisfied.
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Amend. 58 5.2A-1
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HEF-6 6.
Large thermal stresses arise in the outer region of j
the perforated area of the steam generator tubesheet to the rim.
Creep rupture damage
' combined with fatique due to relaxation of high residual stresses limits life of the component.
i The' ASME Code does not provide acceptance criteria for the design of the perforated plotes in elevated temperature service.
Jaaponse:
i The resolution of this item consists of the following actions:
A.
Thu Applicant is committed to perform Mcchanical Properties Tests to verify and supplement ASP.E Code j
and RDT Standards methods and design information for assuring the structural adequacy of the steam generator.
Prototype Steam Generator Tests will be run to verify certain performance characteristics.
Hydraulic Tert Model, Large Leak Tests. Few Tube Tests,R$kTests(departnrefromnucleateboiling),
l Tube Support wear Tests. Modular Steam Generator Tests. Single Tube Performance. Stability and i
Interaction Tests, Tube-to-Tubesheet Weld Tests.
Scale Hydraulie Model Feature Tests, and Flow-Induced Vibration Tests will also be conducted.
i The tests are needed to confirm the structural adequacy of the tubes.
l B.
The Applicant will carry out a confirmatory program to confirm the adequacy of the methods and criteria used to ensure the structural adequacy of the i
tubesheet' for its intended lifetime.
The specific tasks involved are as follows:
1)
Develop effective properties of perforated region for use in inelastic design analyses.
l 2)
Evaluate effects of thermal gradients and equivalent material property variations on l
ligaments near periphery of perforated region.
3)
Extended existing Appendix A-8000 Code methods for calculating the rTnearized membrane, shear i
and in-plane bending # stresses in the ligamer.ts using the equivalent solid plate stresses.
Inc)ude all of these nominal stresses in the comparison with allcwable primary membrant plus brnding, and primary plus secondury allovsbles.
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4)
Develop methods of evaluating local cyclic plastic strain concentration effects based on equivalent solid plate stres.ses for use in the f atigue evaluation.
5)
Develop methods of evaluating local cyclic creep strain concentration effects based on equivalent solid plate stresses for use in the fatigue evaluation.
6)
Evolunte elastic follow-up into outermost ligaments and (1)
Reclassify portion of discontinuity stresses caused by pressure and mechanical loads as primary in accordance with the associated amount of einstic follow-up that occurs during thermal transients.
(ii)
Reclassify portion of thermal stresses os i
primary in accordance with the amount of clastic follow-up that occurs during thermal transients.
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7)
Develop ratcheting evaluation methods for outermost ligaments based on elastic equivalent solid plate stresses reclassified per Item 6 shd including nominal membrane, shear and in-plane bending # stresses.
e 8)
Develop creep rupture damage evaluation methods for outermost ligaments bseed on equivalent solid plate stresses.
The ofrects or elastic i
follow-up will reduce the amount or stress i
relaxation and increase the creep rupture damage.
9)
Perform detailed tube-to-tubesheet joint i
analysis for tubes in high radial thermal transient region at periphery of the perforated i
region including local thermal effests.
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'In-plane bending occurs on eithir side or minimum ligament i
ecction creating a " kinking" type of failure mechanism.
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MED Item 69:
Applicant should review the current version of the MEB 3-1 (Revision 1) to assure that other documents used for specifying pipe break locations provide an equivalent level of conservatism.
Response
MEB 3-1, Rev. 1, is now being applied to CRBRP systems with the exception of the main PHTS and IHTS piping.
The piping integrity analysis addresses the main piping.
The enclosed change pages for PSAR Section 3.6 provide the appropriate references to BTP MEB 3-1 and to BTP ASB 3-1.
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I 3.6 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCfATED WITH THE POSTUl_ATED F
RUPTURE OF PIPING u.Q n.
CRBRP systems and components important to safety will be appropriately protected against dynamic offects, including the offects of missiles, pipe whipping, and discharging fluids that may result.from equipment.f ailures or other events.
3.6.1
. Systems In Which Ploe Breaks are Postulated 3.6.1.1 Svetems inside Containmani Spontaneous ruptures of heat transport system (HTS) and auxillary sodium piping inside containment are not considered credible because of the high quality of the piping, operating temperature and pressure conditions for this piping, the inert environment provided for it, and the capability of the leak detection system to provide an early warning of any breach in the piping boundary. ' As a result, massive f ailures of this sodium HTS piping have not been included in the design bases for CRBRP systes Inside containment. A four inch crack, which leads to 8 gpm PHTS leak rates, has been chosen as the design basis. A description of the analyses and test r.esults to support this position are presented in the Piping integrity Status Report (Reference 2, Section 1.6).
A similar detailed evaluation has not been performed for the sodium piping in the auxillary liquid metal syst as. The piping parmeters (e.g. t/D ratio),
service conditions (temperature, pressure, duty cycle), monitoring and inservice inspection techniques are similar to those for the heat transport
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system. Based on this, the maximum credible crack length is not larger than the 4 inch crack specified for the PHTS. The cell liner design is proceeding on the basis of containing the 4. Inch crack with pressures and temperatures that are characteristic of the assogiated sodium systm.
A3h The chit led water system piping in-ontainment is moderate energy piping according to the definition in B 3-1: Pipe leaks are postulated in the piping and mitigated by the features discussed in Section 9.7.3.
3.6.1.2 Systems Outside containment For systems outside o(y MfiS S-l containment, the Intent of the guidelines in AppxL 0
- 59e> Branch Technical PositionMPON ? I ' J. ". 0".xc, ': 5.- 7/12/72 will be used as a basis for leak evaluations. Where seamless. pipe is used longitudinal 31 breaks will not be postulated if all stresses are below 0.8 (1.2 Ss + S )-
A Separation and Isolation of equipment by arrangement as shown in the figures in Section 1.2, atmosphere separation as described in Section 3A, and equipment enclosure are provided to protect safety systems and components required to shutdown the reactor saf ely. The high and moderate energy piping systems outside containment are listed in Tables 3.6-1 and 3.6-2, with the PSAR Section which discusses the system and the potential results of pipe leaks. Chapter 27 15.0 also contains analyses of postulated pipe leaks.
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Amend. 61 3.6-1 Sept. 1981
3.6.1.2.1. Water / Steam Sy' stems The definitions contained in Appendix'A to the Branch Technical Positions are considered to be applicable to the water and steam piping outside containment. The following is a tabulation of the high and moderate energy systems together with a discussion of the design features which protect the essential systems necessary to shut the reactor down and to mitigate the consequences of a postulated pipe break.
r 3.6.1.2.1.1 -Steam Generator Auxiliary Heat Removal System (SGAHRS) j The elements of this system are described in Section 5.6.1.
r A58 3-I The piping fromJthe auxiliary feedwater isolation valves to the steamdrum,thepipinghetween the steam drum and the Protected Air Cooled Condenser (PACC), and t piping from the steam drum to the Auxiliary Feed-water Pump (AFWP) turbi drive isolation valve are high pressure as defined in Appendix A to e fBT and will be evaluated for postulated ruptures.
Because these pipes are ocated outside of the cells containing the major auxiliary feedwater components, a continued supply of auxiliary feedwater will be available after a postulated rupture.
Separation of the HTS loops l
and their respective cells and Steam Generator Building flooding protection 45 provisions (Section 7.6.5) prevent propagation of a pipe rupture event to adjacent loops and thus the essential systems to mitigate the consequences of the rupture are maintained.
-l The piping runs from the AFWP to the AFW isolation valves and from the turbine drive steam supply isolation valve to the turbine drive are low-pressure and low-temperature lines during normal plant operation.
Both lines are subjected to high-pressure during AFW operating periods and the turbine supply line is also subjected to high temperature conditions during the time the turbine is operational. However, the SGAHRS operating time is anticipated to be less than 2% of the plant operating time since the auxiliary feedwater portion of SGAHRS wil1~ not be utilized unless the Normal Heat Rejection ystem (main condenser) or Feedwater Supply System j
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has been lost. Therefore, this piping will be e. valuated as moderate energy piping for through the wall cracks during normal plant operations.
The primary concern for a crack in this pi' ping is for protection of the major auxiliary feedwater components from the accumulated water that has i
leaked. The major components are elevated to provide this protection and i
to prevent the propagation of event consequences.
Other essential systems 4
for reactor shutdown are not impacted by low temperature and low pressure j
leaks from this piping.
i No piping breaks will be postulated for the low pressure and low l
27 temperature piping run from the protected Water Storage Tank to the AFWP.
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I 3.6-la Amend. 45 (I
July 1978
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3.6.1.2.1.2 Steam Generator System jg The elements of this system are described in Section 5.5.
The majority of piping in this system is hig pressure as defined. in Appen-dix A of Branch Technical Position J-1.
The feedwater piping between the FW control valve and the steam drum, the recirculation and main steam piping joining the evaporators and steam drum, and the steam piping from the steam drum to superheated steam isolation valves will be evaluated for postulated ruptures. Because this equipmant is separated by loop into building cells and Steam Generator Building 45 flooding protection is provided (Section 7.6.5), the effects of these ruptures will not propagate to adjacent cells and thus essential systems for non-ruptured HTS loops necessary to mitigate the rupture consequences will remain operable.
3.6.1.2.1.3 Main Steam, Condensate and Feedwater Systems The elements of these systems are discussed in Chapter 10.0.
These systems are located in the Turbine Generator Building.
Safety related equipment in the Steam Generator Building is protected by the hardened wall of the Steam Generator Building.
.49l 3. 6.1. 2.1.4 Chilled and Treated Water Systems The elements of these systems are discussed in Sections 9.7 and C
9.9.
They are classified as moderate energy systems.
The primary pro-tection against pipe leaks in'those systems are floor drains, isolation provisions and separation of redundant equipment.
Separation of equip-ment by loops assures operation of redundant loops of essential systems 27 in tne event of a piping leak.
3.6-lb Amend. 49 Apr. 1979
3.6.1.2.2 Sodium and NaK Systems
' Sodium and NaK are unique relative to water in that their boiling
{-
temperatures are significantly higher than their normal operating temperatures in CRBRP applications. Therefore, in comparison with conventional water systems, no CRBRP sodium or NaK systems operate with any significant amount of internal fluid stored energy.
The highest temperature sodium system outside 8
the CRBRP containment operates at 965 F vs. the sodium saturation temperature of 1630*F at atmospheric pressure. Typical NaK operating temperatures are in the range of 500*F vs. the NaK saturation temperature of *1450*F at atmospheric pressure. Therefore, from the standpoint of fluid stored energy the CRBRP sodium and NaK systems outside containment contain essentially n,o stored energy and are therefore moderate energy systems.
In addition to fluid considerations, total system operating modes have been evaluated relative to their impacts upon failure.
In spite of fluid conditions, systems can have high internal energy, e.g., subcooled liquid systems with saturated liquid pressurizers. The cperating pressure of essentially all of the CRBRP systems is solely dependent on the developed discharge head of the system pump. This energy source is generated only to accommodate the resistance of the system to fluid flow. When considering system failures, this resistance to flow immediately diminishes resulting I
in immediate loss of system pressure.
There is therefore, very little energy available in a failed system to induce large reaction forces at the failure point.
The initial reaction forces diminish within milliseconds following a large failure.
The only sodium or NaK system within CRBRP which has an external pressure source is the intermediate heat transport system which utilizes an expansion
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tanh that contains pressurized co'ver gas.
Even though this system does contain a pressure source, the to.tal system pressure is well below 275 psig (s225psig).
j From the foregoing discussion, it is~ concluded that sodium and NaK systems are moderate energy systems within the inten of the definitions given in Appendix A of Branch Technical Position 3-1. They will all be evaluated in accordance with the intent of the guidelines cf th ?. T. 0"a.., 23 Jhttr r# '/!?/71. The following is a tabulation of these systems and a (E
discussion of the design features which protect the systems necessary to shut the reactor down and to mitiga~te the consequences of a postulated MES.3 pipe leak.
~3.6.1.2.2.1 Intermediate Heat Transport System (Outside Containment)
STP MGS 3-/
The intermediate heat transport system piping outside containment will be evaluated in accorda e with the intent of the requirements for moderate energy piping established by "-
The primary
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protection provided for postulated leaks in this piping is separation of the
'27 IHTS loops by building cells.
Sodium catch pans are included in the design I
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.3.6-1c
3.6.2 Pipe Break Criteria y ggg/
For sodium and NaK lines, the approach and criteria escribed in Section 3.6 will be followed. For piping of other fluid ystems inside the containment, the intent of Re"h+r.,1..i; ' a will be usdd as a guide for postulating break locations, break sizes and orientations; and pipe restraints will be provided as necessary.
There is no high-pressure water and steam piping inside the con-tainment of the CRBRP. For water / steam piping outside the containment.
OP pipe breaks will be postulated in accordance with the bases described in M88 Id Sections 3.6.1.1 and 3.6.1.2.
For those piping runs for which pipe breaks
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are to be postulated, the criteria as set forth in ?;,.
L " t: ?. ". 0 ". : :..,
A.l" "-- ' '"" will be follcwed with respset to: pipe break locations, 1
break sizes and orientations. However, since the SGS and SGAHRS steam / water piping is of seamless construction and longitudinal pipe welds are not used 29 at component connections, longitudinal pipe breaks will only be postulated in nominal pipe sizes of 4 inches and larger at terminal ends and inter-mediate locations where the stresses exceed 0.8 (1.2 Sh + S ) based on the A
loadings specified in @..? S ^ +- tk O L..y
'.m L, unless the axial stress is greater than 1.5 times the circumferentia stress.
B TP M 68 B-l.
The essence of steam / water pipe break criteria is that for a water / steam pipe break in any loop, including the steam / water release, the effects rnst be restricted to that loop and must not impair operability of the other two loops SGAHRS equipment in the other two loops, or SGAHRS equipment in the lower cells of the auxiliary bay. The effects of a water / steam ipe break
-j-shall not lead to an uncontrolled sodium-water reaction and resu t in hydrogen gas release into a cell of the affected loop unless the consequences are shown to be within acce Protection of critical components must be provided by (1)ptable limits.
physicalseparation,(2)pipewhiprestraints,or (3) impingement shields. This protection will consider pipe moveunt caused by the reaction force of the jet and impingement of steam / water on adjacent components and piping.
3.6.3 Design loading Combinations 3
In all locations where piping breaks are postulated to occur, analysis will be performed on the unrestrained piping system, except where it has been 1
shown that pipe dynamics resulting from a pipe break can be neglected. This analysis is required to insure against possible damage to neighboring reactor coolant boundary components and all essential equipment located within contain-ment and the SG cells. Consideration of damage propagation to adjacent piping will be given as appropriate, consistent with the intent of Regulatory Guide
.l.46.
For the pipe supports and structures, the load combination as defined in Section 3.8.3.3.10 will be used.
For jet impingement interactions, piping subject to jet impingement shall be assumed to fail if: a) the jet impingement load exceeds the load carrying capacities of the pipe support system; or b) the pipe stress exceeds the allowables for the faulted loading condition as defined for the applicable ASME-III classifications.
Amend. 34 3.6-2 Feb.1977
.