ML20054C419
| ML20054C419 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 04/16/1982 |
| From: | Davidson D CLEVELAND ELECTRIC ILLUMINATING CO. |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8204210025 | |
| Download: ML20054C419 (36) | |
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Dalwyn R. Davidson VtCE PHESIDENT SYSTEM E NGINE ERING AND CONSTRUCTION April 16, 1982 Nr g
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Mr. A. Schwencer Chief, Licensing Bra ich No. 2 Division of Licensing r
U. S. Nuclear Regulatory Commission b.,
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Washington, D. C.
20555 t.
p 5-Perry Nuclear Power Plant Docket Nos. 50-440; 50 441 Response to Request for Additional Information -
Clarification of Mechanical Engineering Branch responses
Dear Mr. Schwencer:
This letter and its attachment is submitted to provide further clarification and information per your request of April 2,1982 in regards to the Mechanical Engineering Branch draft SEP.
It is our intention to incorporate these responses in a subsequent amendment to our Final Safety Analysis Report.
Very Truly Yours, Dalwyn R. Davidson Vice President System Engineering and Construction DRD: mlb cc:
Jay Silberg John Stefa lo Max Gildner l1 s'
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h4gOCK05000440 0025 820416 E
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Question 1 The applicant should clarify whether the main steam and feedwater piping and supports located in the auxiliary and turbine building are analyzed and designed for the OBE and SSE.
Response
1.
The piping and supports for the four main steam lines running through the Auxiliary and Turbine Buildings are analyzed and designed for the OBE and SSE.
2.
The piping and supports for the two feedwater lines in the Auxiliary Building are analyzed and designed for the OBE and SSE.
Some of the piping inside the Turbine Building (that nearest the piping in the Auxiliary Building) is analyzed and designed for the OBE and SSE.
3.
A significant portion of the piping described in items "1" and "2" above is not ASME piping.
The supports nearest the class break are designed for the OBE and SSE to ASME limits. The remaining supports further away from the class break are designed for the OBE and SSE but not necessarily to ASME limits.
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Question 2 Pipe whip restraints to be finalized and provided in FSAR.
(DSER 3.6-29)
Response
Pipe whip restraints will be finalized in the FSAR when the design is complete.
Question 3 The applicant to specify that the augmented ISI program is applicable to all'high-energy piping in the break l
exclusion region; not just MS and FN.
(DSER 3.6-9.)
1
Response
All high energy lines over 1 inch in diameter which have a break exclusion area will undergo an augmented inservice
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inspection. The augumented program will include a 100%
UT examination of all welds except for socket welds which will be examined using PT examination methods.
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Question 4 Applicant needs to clarify the criteria used for the design of pipe whip restraints that also function as pipe supports.
(DSER 3.6-24.)
Response
pipe whip restraints which also serve as guides are as follows:
1.
Main steam line restraint / guides inside drywell including guide numbers G101B and G10lD as shown in FSAR Figure 3.5-65 and guide numbers G101A and G101C not shown.
These restraint guides are designed and supplied by GE.
GE has classified these restraint /
guides as a plate and shell type support under Subsection NF of the ASME Code.
These restraints attach to the building steel which is classified AISC.
2.
RCIC restraint / guide inside drywell, G201A as shown in FSAR Figure 3.6-70.
This restraint consists of a structural steel frame with an insert around the pipe which serves as a guide. The insert / guide is classified as a linear type support under Subsection NF of ASME Code.
The supporting framework is classified as building structure and designed in accordance with AISC.
This approach is consistent with that of pipe supports on steel platforms.
l 3.
Main steam and feedwater restraint / support structures in the Auxiliary Building steam tunnel as shown in Figures 3.6-75 and 3.6-76.
These l
restraints consist of a structural steel frame with an insert around the pipe which serves as a guide.
The insert / guide is classified as a linear type support under Subsection NF of ASME Code. The supporting framework is classified as building structure and designed in accordance with AISC. This approach is consistent with that of pipe support on steel platforms.
Question 5 The applicant must provide additional information for the resolution of the cracking of the jet pump holddown beams.
(DSER 3.9-17)
Response
The following additional information is provided:
1.
The jet pump preload will be reduced from 30,000 lbs to 25,000 lbs.
2.
The jet pump beams will be inspected for crack initiation using ultrasonic techniques. The f 't r s t inspection will be performed after the first 10 years of operation, and subsequently, every two years.
The inspection plan is based on the GE prediction, with 97.5%
probability, that no cracks will occur before 19 years of operation, and that the period from crack initiation until beam failure is expected to be greater than 2 years.
3.
Any cracked beam found during inspections will be replaced.
It is expected that any replacements would be beams with a reduced tendancy for cracking.
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Question 6 The applicant must provide additional information for the modifications to the CRD return line (NUREG-0619).
(DSER 3.9-18)
Response
CEI is committed to NUREG-0619 with respect to the CRD return line modifications, including those described in section 8.1.4.a' and 8.1.4.c'.
CEI plans to run a test to verify proper flow in accordance with the NUREG.-
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Question 7 The applicant must provide the program for the inservice testing of pumps and valves.
(DSER 3.9-39).
Response
The program for inservice testing of pumps and valves is currently under preparation. This program will be submitted for NRC review when it is completed.
Question 8 The applicant to provide a response to the leak testing of pressure isolation valves.
(DSER 3.9-20).
Response
The program for leak testing of pressure isolation valves is currently being prepared as part of the ASME Code Section XI inservice testing program. This program will be submitted for NRC review when it is completed.
Question 9 The applicant to provide a program for preservice exam and pre-operational testing of snubbers.
(DSER 3.9-37).
Response
A.
For Preservice Examination, CEI plans to conduct a program of inspection of snubbers which will include the following items.
1.
A check for visible signs of damage or impaired operability including indications of seizing or jamming.
2.
A verification that location orientation, position setting and configuration are in accordance with design drawings.
3.
An assessment that clearances are available for snubber movement.
4.
Non-mechanical snubbers will be checked for proper fluid level and for leakage.
5.
A verification that structural connections and connecting hardware is correctly installed.
This program will be conducted prior to the start-up test to help insure operability of the snubbers during start-up testing.
B.
For start-up testing, CEI plans to conduct a program for verifying proper thermal movements of snubbers during the initial heat up of the systems whose operating temperature exceeds 250 F.
A description of this program will be provided in Chapter 14 of the FSAR in a future amendment.
Question 10 The applicant to provide specific loads, load combination, and acceptance criteria for piping in the wetvell that are affected by hydrodynamic loads.
Response
The Jttached design specifications identifies the loads and load combinations for piping that are af fected by hydrodynamic loads in the wetwell (suppression pool).
For exampic, the specific loads used to design the SRV lines identified in Section 3:02.2, paragraph 2a, 2b, 3b, 4. 8, 9, are 10, 11, 12, and 13.
For other lines in the wetwell, the loads are identified in paragraphs 2a, 2b, 2c, 4, 8, 9, 10, 11, 12, and 13.
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3:02.2 Design Loadings 1.
Loading combinations:
Table 3A and B presents the detailed loading combinations 1
for~the various system analysis conditions.
2.
Sustained loadings:
a.
Deadweight loading, caused by the weight of the pipe, 1
contents, insulation (reference Gilbert Associates, Inc.
l Specification SP-353-4549-00 or SP-354-4549-00), and equipment mounted in-line, such as pumps, valves, strainers, etc.
Interface information may be obtained by referencing the certified vendor's print of the equipment.
b.
Pressure stress, produced by internal fluid pressure. The system design and operating pressure are presented in Table 1.
The system design pressures from Table I shall I
be used in the Tables 3B and 11 analyses.
c.
Other mechanical loads, such as unbalanced pressure forces, applied external loads, carth loading, vehicle loading, etc.
See GE drawing 762E525 (GAI folder number 4549-23-061).
3.
Occasional loadings:
a.
Flow transients, resultant from the rapid opening or closing of a valve or other situations resulting in fluid momentum changes.
There are no such transients identified 1
for this system.
b.
Safety valve reaction forces, resulting from the lifting or safety valves.
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4.
Earthquake loadinns, resulting from:
a.
Equipment not supported by the piping shall be considered rigid.
b.
Induced vibrations of the piping system in response to the l
time varying forces and displacements transmitted to the piping from the supporting structures. The earthquake loadings are based on two (2) earthquake models; the l
Operating Basis Earthquake (OllE) and the Safe Shutdown Earthquake (SSE).
The characteristics of the carthquakes, 1
i that af fect piping design, are provided in specification SP-750-4549-00.
Differential building movements are shown l
in the Gilbert Associates, Inc.
Structural Engineering Department Report "PNPP Seismic Displacement for Structures".
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The following maximum damping factors shall be applied to the, analysis of piping and equipment.
COMPONENT PERCENT OF CRITICAL DAMPING OBE SSE Equipment and large diameter 2
3 pipe systems, pipe diamecer greater than 12 inches Small diameter piping systems, 1
2 diameter less than or equal to 12 inches For the dynamic analysis of active components as noted in Regulatory Guide 1.61, damping values for OBE also apply to SSE.
5.
Jet Loadings, produced by impingement of a jet or fluid from adjacent piping or vessels that have ruptures postulated.
There are no jet loadings on the fuel transfer system.
6.
Pipe rupture loadings, resulting from the blowdown of the system through a crack or other openings in the system piping.
This loading differs from the jet loadings, described above, in that it results from rupture in the system under analysis, and not from other adjacent systems.
The Fuel Transfer System has 1
no pipe rupture loads.
7.
Annulus pressurization loadings, resulting from LOCA due to a circumterental break in a teedwater, mainsteam or recirculation line within or adjacent to the reactor pressure vessel / shield wall annulus region.
These inertial loads are generated by the rapid build-up of presture in the annulus between the reactor pressure vessel and the biological shield wall.
See Table 10 for load cross-reference information.
8.
Pool swell loadinns, resulting from a drywell wall vent clearing transient following a design basia LOCA. Air and steam vented from the drywell to the suppression pool causes the pool to swell in a bulk mode.
These pool swell loads include inertial (vibrational), displacements, and drag loadings. See Table 10 for cross reference information.
9.
Safe,ty/ Relief valve loadings, resulting from actuation of one or n. ore sa tety/re t ie t valves including those actuated by the Automatic Depressurization system.
The loads are caused by the expulsion of the water column in the discharge line and steam blowdown into the suppression pool.
Safety / relief valve discharge causes inertial, drag, and displacement loads. See Table 10 for cross-reference information.
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10.
Chunting loadings, resulting from water oscillating in the top row or drywell wall vents.
This oscillation is caused by low blowdown steam flows when the vents are capable of condensing the steam. Chugging causes inertial and displacement loads.
See Table 10 for cross-reference information.
11.
Condensation oscillation loadings, resulting from high steam mass tiow rates through the top drywell wall vents when drywell free space is vented to containment free space. This causes a pressure oscillation on the drywell wall.
Condensation oscillation causes inertial and displacement loadings. See Table 10 for cross-reference information.
12.
Weir swell loadings, resulting from ECCS water spilling from the pipe creak and condensing steam in the drywell. This will I
create a drywell to containment negative pressure differential causing suppression pool water to be drawn into the weir annulus and swelling over the weir wall.
Weir swell causes impact, drag, and inertial loads on piping and components in the drywell between El. 599'-0" and 613'-0".
See Table 10 for cross reference information.
13.
Thermal loads:
Thermal expansion, which produces expansion stresses:
Thermal expansion stresses are calculated from the expected temperature range of the piping system. The piping system temperature range is the temperature difference between the system at ambient temperature as defined in Table 6 and the system operating temperature indicated in a value of 1.0 should be utilized for the stress range reduction factor, f, for analysis of ASME Code Class 2 and 3 piping.
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i TABLE 3A ASME CODE ANALYSIS LOAD COMBINATIONS AND STRESS LIMITS DESIGN CRITERIA FOR CLASS 1 PIPING SYSTEMS (l)
PERRY NUCLEAR POWER PLANT UNITS 1 and 2 System Analysis Condition Primary Loads (12) (14)
Primary & Secondary Loads (11)(12)(14) 1.
Design (10)
Design pressure, deadweight, other N/A sustained mechanical loads, OBE (inertial) - Equation 9 of NB-3600 1 1.5 Sm 2.
Normal N/A Operating pressure, other sustained mechanical loads (primary and secondary effects), OBE (inertial),
a-OBE (displacement), thermal
- Equation (10) of NB-3600;< 3.0 S !
m If Equation (10) is exceeded Equation (12) $ 3.0 Sm Equation (13) $ 3.0 Sm prs 1.0 3.
Upset (13)(16)
Peak pressure, deadweight, other Peak pressure, thermal
~ _ _ _ _ _ - _ _ _ _ _ - -
sustained mechanical loads, SRV sustained mechanical loads (primary &
j (inertial) SRV (drag), OBE (inertial) secondary effects) OBE (inertial),
- Equation (9) of NB-3600 i the OBE (displacement), SRV (inertial),
lesser of 1.8 S or 1.5 S SRV (displacement), SRV (drag), other m
y occasional loads (primary & secondary effects) - Equation (10) of NB-3600 $ 3.0 Sm UF < l.0 l
If Equation (10) is exceeded:
Equation (12) $ 3.0 Sm Equatien (13) $ 3.0 Sm i
i TABLE 3A i
ASME CODE ANALYSIS LOAD COMBINATIONS AND STRESS LIMITS DESIGN CRITERIA FOR CLASS 1 PIPING SYSTEMS (l)
PERRY NUCLEAR POWER PLANT UNITS 1 and 2 4
System Analysis Condition Primary Loads (12) (14)
Primary & Secondary Loads (11)(12)(14) i 4.
Emergency Peak pressure, deadweight, other N/A sustained mechanical loads, other occasional loads, SRV (inertial),
chugging (inertial), or condensation oscillation (inertial)(2), and SRV (drag) or pipe rupture or jet impingement
' ~ ' Equation (9) of NB-3600 < 2.25 S
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5.
Faulted I (7)(8)(9)
Peak pressure, deadweight, other N/A sustained mechanical loads, other occasional loads, pool swell (inertial) and pipe rupture or jet impingement or pool swell (drag) or pool swell (impact)(4)(6) - Equation (9) of NB-3600 <:,2.25 S or 3.0 Sm (18) m 6.
Faulted II Peak pressure, deadweight, other N/A sustained mechanical loads, other occasional loads, SSE (inertial), SRV (inertial), and chugging (inertial) or i
condensation oscillation (inertial)(2) andpiperugtureorjet impingement or SRV (drag)( ) - Equation (9) of NB-3600 < 2.25 S or 3.0 Sm (18) m
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TABLE 3A ASME CODE ANALYSIS LOAD COMBINATIONS AND STRESS LIMITS DESIGN CRITERIA FOR CLASS 1 PIPING SYSTEMS (I)
)
PERRY NUCLEAR POWER PLANT UNITS 1 and 2 System Analysis Condition Primary Loads (12) (14)
Primary & Secondary Loads (11)(12)(14) 7.
Faulted III Peak pressure, deadweight, other N/A
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sustained mechanical loads, other i
occasional loads, SSE (inertial), and chugging (inertial) or condensation i
oscillation (inertial)(2),(and pipe rupture or jet impingement 5) _
l Equation (9) of NB-3600 1 2.25 Sm or 3.0 Sm (18) o 8.
Faulted IV (7)(8)(9)
Peak pressure, deadweight, other N/A sustained mechanical loads, other occasional loads,'SSE (inertial), pool swell (inertial), and pipe rupture or jet impingement or pool swell (drag) or pool swell (impact)(9)(6) - Equation (9) of NB-360012.25 Sm or 3.0 Sm (l')
1 9.
Faulted V Peak pressure, deadweight, other N/A sustained mechanical loads, other occasional loads, SSE (inertial), annulus pressurization (inertial) - Equation (9) of NB-3600 1 2.25 S or 3.0 Sm (18) m
- 10. Faulted VI (19)
Peak pressure, deadweight, other N/A sustained mechanical loads, other occasional loads, SSE (inertial), SRV (inertial), SRV (drag)(17), weir swell (inertial) and weir swell (impact) or weir swell (drag) - Equation (9) of NB-360012.25 Sm or 3.0 Sm (18) v
TABLE 3A ASME CODE ANALYSIS LOAD COMBINATIONS AND STRESS LIMITS DESIGN CRITERIA FOR CLASS 1 PIPING SYSTEMS (I)
PERRY NUCLEAR POWER PLANT UNITS 1 and 2 System Analysis Condition Primary Loads (12) (14)
Primary & Secondary Loads (11)(12)(14)
- 11. Ilydrotest (10)
Ilydrostatic test pressure (per Table 2),
N/A deadweight shall be evaluated in accordance with subsection NB-3226.
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TABLE 3B ASME CODE ANALYSIS LOAD COMBINATION AND STRESS LIMIT DESIGN CRITERIA FOR CLASS 2 AND 3 PIPING SYSTEMS (1)
PERRY NUCLEAR POWER PLANT UNITS 1 AND 2 System Analysis Secondary Loads Primary and Secondary condition (10)
Primary Loads (12)(14)
(11)(12)(14)
(11 )(12 )(14 )
1.
Normal Internal design pressure, deadweight, Thermal expansion, other All primary plus all secondary other sustained mechanical loads -
sustained mechanical loads - Equation 11 of Equation 8 of NC/ND 3652 ji 1.0Sh loads (displacement)
NC/ND 3652 j[ (SA+S) h Equation 10 of NC/ND 3652 j[ 1.0SA 2.
Upset (16)
Peak pressure, deadweight, other Thermal expansion, other Peak pressure, deadweight', all sustained mechanical loads, other sustained mechanical secondary loads - Equation 11 occasional loads, SRV (inertial),
loads (displacement),
of NC/ND 3652 ji 1.2 (SA+S) h SRV (drag), OBE (inertial) -
OBE (displacement), SRV Equation 9 of NC/ND 3652 with -
(displacement), other sum of longitudinal stress occasional loads j[ 1.2 Sh (secondary effect's) -
Equation 10 of NC/ND 3652 j[ 1.2SA 3.
Emergency Peak pressure, deadweight, other N/A N/A sustained mechanical loads, other occasional loads, SRV (inertial),
chugging (inertial), or condensa-tion oscillation (inertial) (2),
and SRV (drag) or pipe rupture or jet impingement (3) - Equation 9 of NC/ND 3652 with sum of longitudinal stressji 1.8 Sh h%,_
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I TABLE 3B ASME CODE ANALYSIS LOAD COMBINATION AND STRESS LIMIT DESIGN CRITERIA j
FOR CLASS 2 AND 3 PIPING SYSTEMS PERRY NUCLEAR POWER PLANT UNITS 1 AND 2 j
System Analysis Secondary Loads Primary and Secondary Condition (10)
Primary Loads (12)(14)
(11)(12)(14)
(11)(12)(14) 4.
Faulted I Peak pressure, deadweight, other N/A N/A sustained mechanical loads, other occasional loads, pool swell (inertial) and pipe rupture or jet (7)(8)(9) impingement or pool swell (drag) or pool swell (impact)(4)(6) _
j Equation 9 of NC/ND 3652 with sumoflong8)itudinal stress < l.8 Sh or 2.4 Sh
- 5. Faulted II Peak pressure, deadweight, other N/A N/A sustained mechanical loads, other occasional loads, SSE (inertial),
SRV (inertial), and chugging (inertial) or condensation oscillation (insertial)(2), and pipe rupture or impingement or SRV (drag)(3) jet
< Equation 9 of NC/ND 3652 with sum of longitudinal stress < l.8 S or 2.4 Sh
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Faulted III Peak pressure, deadweight, other N/A N/A sustained mechanical loads, other occasional loads, and chugging (inertial) or condensation oscillation (inertial)(2), and pipe rupture or jet impingement (5)
< Equation 9 of NC/ND 3652 with sum of longitudinal stress
, < l.8 Sh or 2.4 Sh I
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TABLE 3B ASP.E CODE ANALYSIS LOAD COMBINATION AND STRESS LIMIT DESIGN CRITERIA FOR CLASS 2 AND 3 PIPING SYSTEMS PERRY NUCLEAR POWER PLANT UNITS 1 AND 2 System Analysis
~~
Secondary Loads Primary and Secondary Condition (10)
Primary Loads (12)(14)
(11)(12)(14)
(11)(12)(14) 7.
Faulted IV Peak pressure, deadweight, other N/A N/A j
sustained mechanical loads, other occasional loads, SSE (inertial),
pool swell (inertial), and pipe 3
1 (7)(8)(9) rupture or jet impingement or pool swell (drag ()6) pool swell e
or (impact)(4)
Equation 9 of NC/ND 3652 with sum of lon itudinal stress < 1.8 Sh or 2.4 S (
h 8.
Faulted V Peak Pressure, deadweight, other N/A N/A i
sustained mechanical loads, other occasional loads, SSE (inertial),
annulus pressurization (inertial),
- Equation 9 of NC/ND 3652 with sum of longitudinal stress II8) j
< 1.8 Sh or 2.4 Sh 9.
Faulted VI Peak pressure, deadweight, other N/A N/A 4
sustained mechanical loads, other (19) occasional loads, SSE (inertial),
SRV (inertial), SRV (drag)(I7),
and weir l
swell (impact) or weir swell (drag) - Equation 9 of NC/ND 3652 with sum of longitudinal stress
< 1.8 S h or 2.4 S (18) h e
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l NOTES TO TABLE 3A AND 3B f
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Notes:
(1) Refer to Section 3:02 for identification and definition of individual loads.
1
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(2) Chugging and condensation oscillation do not occur simultaneously.
4 (3) Pipe rupture, jet impingement, and SRV (drag) do not occur, simultaneously.
(4) Pipe rupture, jet impingement, pool swell (impact), and pool swell (drag) l do not occur simultaneously.
(5) Pipe rupture and jet impingement do not occur simultaneously except where Noted in Tables 7 and 8.
(6) Pool swell (impact) cannot occur in the pool.
(7)
Pool swell (impact) occurs before pool swell (drag) or pool swell
.(inertial).
(8)
Pool swell (inertial) may occur with pool swell (drag).
]
(9) Pool swell (drag) includes Fallback (out of pool) loads.
e (10) There is no Design or Ilydrotest System analysis condition required for i
Class 2 or 3 analysis.
These apply to* Class 1 only.
(11) Thermal - loads due to expansion and thermal stresses caused by rapid temperature fluctuation.
Refer to Section 3:02.2 Subitem 13.
(12) Peak pressure - coincident pressure associated with that load combination resulting in maximum calculated stress.
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(13) The 1979 ASME code through winter 1975 addenda does not require a primary i
stress intensity evaluation for upset condition.for class 1.
This check is made for comparison purposes only.
(14) Earthquake loads shall be assumed to occur where the plant is in the normal operating mode at full power.
Earthquake cycles to be considered are 5 OBE and 1 SSE with 10 stress cycles per earthquake.
l (15) For the type and number of cycles of SRV events, refer to SP-660-4549-00, Appendix A, Section 2, " Induced Loads on Pool Boundary".
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NOTES TO TABLE 3A AND 3B (16) Upset includes system test with the same stress limits. Use the indicated load combination with test pressure per Table 1 in lieu of peak pressure and delete all OBE loads.
(17) Include SRV loads for SBA/IBA only.
(18) Functional capability of piping components which are required to deliver rated flow for any plant condition will be demonstrated on a case-by-case basis (by analysis and/or testing) when Service Limit D'
applies.
(19) The occurrence of weir swell loads and the relationship of these loads to other occasional loadings are the same as those defined for pool swell loads in Notes (4), (6), (7), (8), and (9).'
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Question 11 The applicant to document the annulus pressurization study.
(DSER 3.9-34)
Response
Section 3.6.2.2.3 Titled "Asymetrical Dynamic Loading from Postulated Pipe Break at the RPV Nozzle Safe-End" will be added to the FSAR.
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- and fluid blowdown forces were the same in both analyses. However, a linear l
i approximation was made by NSC for the restraint load-deflection curve supplied by i
I GE.
This approximation is illustrated by Figure 3.6-48.
The effect of this approximation is to give lower energy absorption of a given restraint deflection.
i Typically, this yields higher restraint deflections and lower restraint to structure loads than the GE analysis. The deflection limit used by NSC is the design deflection at one-half the ultimate uniform strain for the GE restraint i
design. The restraint properties used for both analyses are presented by Table 3.6-4.
Break locations and restraints analyzed are shown by Figure 3.6-49.
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A comparison of the NSC analysis with the PDA analysis, as presented by Table 3.6-5, shows that PDA predicts higher loads in 15 of the 18 restraints analyzed. This is due to the NSC model including energy absorbing effects in 1
secondary pipe elements and structural members. However, PDA predicts higher restraint deflections in 50 percent of the restraints. The higher deflections predicted by NSC for the lower loads are caused by the linear approximation used for the force-deflection curve, rather than by differences in computer 1
techniques. This comparison demonstrates that the simplified modeling system used in PDA is adequate for pipe rupture loading, restraint performance, and pipe
]
movement predictions within the meaningful design requirements for these low probability postulated accidents.
i 3.6.2.2.3 Asymmetric Dynamic Loading from Postulated Pipe Break at the RPV i
Nozzle Safe-End a.
Introduction i
In conformance with the intent of Regulatory Guide 1.46, PNPP is designed to 1
accommodate the asymmetric loading resulting from a postulated high energy i
pipe break at the RPV nozzle safe-end. The analyses of pipe breaks of the reactor nozzle safe-end and subsequent transient effects on reactor components and other piping and equipment are generally termed " Asymmetric l
Loading Analyses". The transient loads associated with an instantaneous j
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full circumferential break at the reactor nozzle safe-end can be characterized by:
1.
Annulus oressurization (when break is in the biological shield wall annulus).
2.
Jet impingement on the shield wall and/or on the RPV.
3.
Jet reaction (thrust force) at the RPV broken nozzle.
t 4.
Impact load on the biological shield wall at the pipe whip restraint anchor.
The methodology developed to evaluate the effects of asymmetric loading consists of the following steps:
1.
Determination of the time history mass and energy flow into the annular region and the resultant pressure time history in the annulus.
2.
Determination of jet reaction and, jet impingement forces.
3.
Determination of pipe reaction force time history at the pipe whip restraint anchor.
The forces and pressure mentioned above are employed in a structural dynamic analysis to evaluate the dynamic responses of the RPV, RPV pedestal, RPV internals, the biological shield wall and attached piping systems, and components, b.
Mass Energy Release in Annulus - Annulus Pressurization Section 6.2.1.2, " Containment Subcompartments", gives a description of the break locations, the time history of the mass releases and the resulting subcompartment pressurization. The computer code RELAP 4/ MOD 3 was used in the analysis of the blowdown and pressurization. RELAP 4/ MOD 3 is a general 3.6-22a
=_ _
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computer program used to analyze the thermal hydraulic transient behavior of a water cooled nuclear reactor under postulated loss of coolant accident conditions. The program simultaneously solves the fluid flow, heat transfer, and reactor kinetics equations.
The annulus is divided into zones or subcompartments and RELAP 4 generates a pressure time history for each zone. Further detailed information is provided in Section 6.2.
c.
Jet Forces I
1.
Jet Impingement - Time History The time history of the jet impingement load on the reactor pressure vessel and shield wall is conservatively calculated and the results are illustrated in Figure 3.6-99.
j 2.
Jet Reaction (Thrust) Forces The jet reacticn or thrust acting on the pressure vessel from either a feedwater, recirculation or main steam line break was determined from:
F = K Po A where K = Thrust coefficient related to thermodynamic state of fluid at the exit Po = RPV internal reservoir pressure A = Cross sectional area of the break 3.
Impact Force at Pipe Whip Restraint Anchor l
The impact force at the pipe whip restraint anchor (of the biological shield wall) can be computed from the normal pipe whip analysis with a given break and the details of the pipe whip restraint. Typical pipe whip restraint forces at the anchor are illustrated in Figure 3.6-100.
3.6-22b
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Dynamic Structural Analysis 1.
Models The pressure and jet related force-time histories are combined in a dynamic structural analysis to provide forces, moments, accelerations, and displacements for the pedestal, RPV, RPV internals, and shield wall.
I 2.
Analysis and Results The dynamic structural analysis will produce resultant forces (or stresses) at the nodes of the analytical model.
This analysis can also produce acceleration time history for each node.
Further analysis based on these acceleration time histories can then be performed for any component which is attached to the model. The results of this asymmetric loading analysis can be combined with other loads on the components to ensure structural integrity is maintained in accordance with applicable codes and standards. The final results from the appropriate load condition are summarized in Table 3.9-2.
3.6.2.3 Dynamic Analysis Methods to Verify Integrity and Operability 3.6.2.3.1 Jet Impingement Analyses and Effects on Safety Related Components Criteria used for evaluating the effects of fluid jets on safety related structures, systems, and components are as follows:
Safety related structures, systems, and components should not be so impaired a.
as to preclude essential functions.
b.
Safety related structures, systems, and components which are not necessary to safely shut down the plant for a given postulated pipe break need not be protected from the consequences of the fluid jet.
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i Question 12 The applicant to provide assurance that the FW check valves can function following a line break outside containment.
Response
The feedwater check valves are designed to function following a line bre-outside containment.
The closure rate of the valves is designed to control the deceleration rate of the water in the reverse flow direction when the initial upstream pressure is 1100 psia and downstream pressure is 14.7 psia, d
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Question 13 The staff to review submittal justifying the use of SRSS of dynamic responses for BWR Mark III loads.
/
Response
No itction er quired by applicant.
Question 14 The applicant to verify whether the spring hangers on the SRV piping can accommodate the additional weight of water during the alternate shutdown cooling mode.
(DSER 3.9-35.)
Response
The spring hangers on the SRV piping can accommodate the additional weight of water during the alternate shutdown cooling mode.
(DSER 3.9-35.)
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Question 15 The applicant must clarify the criteria used to evaluate stresses in the break exclusion region of the feedwater piping and other non-seismic Category I.
Response
All break exclusion region piping is supported as seismic Category I.
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Question 16 The applicant to provide test results that show functional capability of piping systems is assured at ASME service Level C stress limits.
Response
The test results referred to are those referenced in the following:
1.
ASME Paper 78-PVP-83
" Evaluation of the Functional Capability of ASME Section III Class 1, 2, and 3 Piping Components."
2.
ASME 74-NE-1
" Plastic Deformation of Piping Due to Pipe - Whip Loading."
3.
" Evaluation of the Plastic Characteristics of Piping Products in Relation to ASME Code Criteria."
This work reviewed a significant amount of test data and found that none could be found "... in which large enough displacements were applied to produce significant reductions in flow area; e.g., 50% reduction of flow arca" (pg. 10).
The report goes on to say that "A local reduction in area in the straight pipe of 50% does not mean that the functional capability of the piping system has been reduced by 50%.
Such a local flow restriction might give only a 1% loss in flow capacity, for a given pressure drop, or 2% higher pressure drop for a given flow rate."
(Pg. 10.)
Question 17 The applicant must provide to the staff for review and approval its comprehensive reactor internals vibration test program for a prototype reactor (BWR/6 - 238 inch)
Response
In a letter from Mr. Glenn G. Sherwood to Mr. Edson G. Case dated April 24, 1978, General Electric describes the reactor internals vibration assurance program for BWR/6 plants. This letter applies to the Perry Nuclear Power Plant as the prototype for the 238-inch vessel size.
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Question 18 IE 79-02 to be resolved.
Response
IE Bulletin 79-02 was provided to NRC Region III.
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