ML20028G079
| ML20028G079 | |
| Person / Time | |
|---|---|
| Site: | 05000447 |
| Issue date: | 02/02/1983 |
| From: | Sherwood G GENERAL ELECTRIC CO. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| JNF-006-83, JNF-6-83, MEN-18-83, MFN-018-83, MFN-18-83, NUDOCS 8302070377 | |
| Download: ML20028G079 (88) | |
Text
{{#Wiki_filter:~ GENER AL h ELECTRIC NUCLEAR POWER SYSTEMS DIVISION GENERAL ELECTRIC COMPANY.175 CURTNER AVE., SAN JOSE. CALIFORNIA 95125 AFN 018-33 MC 682 (408) 925-5040 JNF 006-83 February 2,1983 U.S. Nuclear Regulatory Commissio.n Office of Nuclear Reactor Regulation Washington, DC 20555 Attention: Mr. D.G. Eisenhut, Director Division of Licensing Gentlemen:
SUBJECT:
IN THE MATTER OF 238 NUCLEAR ISLAND GENERAL ELECTRIC STANDARD SAFETY ANALYSIS REPORT (GESSAR II) DOCKET NO. STN 50-447 REVISED DRAFT RESP 0tSES l Attached please find revised final draft responses to selected questions of the Commission's August 25, 1982, October 5, 1982 and November 15, 1982 information requests and the Mechanical Engineering Branch draft SER meetings. Only modifications (new or revised) to the responses of the referenced letters are provided. Responses are provided in the attachments as indicated below: Attachmen t Number Branch 1 Structural Engineering 2 Power Systems 3 Effluent Treatment Systems 4 Containment Systems 5 Reactor Systems 6 Core Performance 7 Mechanical Engineering Sincerely. Glenn G. Sherwsod, Manager Nuclear Safety & Licensing Operation Attachments cc: F.J. Miraglia (w/o attachments) C.0. Thomas (w/o attachments) D.C. Scaletti L.S. Gifford (w/o attachments) ()OO 8302070377 830202 PDR ADOCK 05000
ATTACHMENT NO. 1 DRAFT RESPONSES TO STRUCTURAL ENGINEERING 3 RANCH QUESTIONS . l l l
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]20.0 STRUCTURAL DGINE RING 8AANCH 220.01 It is not clear fn Section 3.3.2.2 of your F5AR how you combine the *
(3.3.2) effects of the wind, the differential pressure and missiles all associated
. with a tornado. Clearly state the tornado loading cominations which you use in the design of all sefsmic Category I structures. A method of coetning these effects which we find acceptable is given in Section 3.3 2 of the Standard Review Plan (SAP).
220.01 e h The tornado loading combination for which all seismic category j I structures are designed is D = D + L + To + Ro + Nt + H (static earth
, pressure, as applicable) noted in Sections 3.8.4.3.1.2 (2) , 3.8.4. 3.2.2 (2) and 3.8.4.3.2.3 (2),
E.Ls k 5 o6msinessummer h 1:0 3.8.4.3.1.1 for the definition of the load W t, and 3.8.4.5.1.2 .,9L'* l for design criteria g ' ^ g
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- A study has been made on the shield building to ::v%tmc. .h ;r st resses i .
_.;- ' due .to tornado missile load co .bined with other applicable loads. t
, ,;g. ..a three-dimensional finito element riodei was prepared to represent
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.the building. A 12-inch dia;eter pipe with.a length of 15 feet
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The results show that the combined stresses are well within 90 pcreent of yield stresses as specified. . y . _ ;. ~
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.t 238 NUCLEAR ISLAND Rav. 0 3.3.4.3.1.1
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/ Loads and Notations (Continued)
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Y, = missile impact equivalent static load on a structure generated by or during the postulated break, like
' pipe whipping, and including a calculated dynamic ,
factor to account for the dynamic nature of the load. W = wind force (subsection 3.3.1.) . Wg = tornado load (Subsection 3.3.2) (tornado-generated s' ' e cribed in su - section 3.5 y1and bet'rter Ot$gn ymcwdureS m S .O 9 P, = internal negative pressure of 3.0 psig due to tornado; accident pressure = 7 psig at main - steam tunnel piping embedment ' N B ,= uplift forces created by t' ' p*f s9e 8 water table " 15 - F = safe shutd' * ! eqs t Wt @0 g.es
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A- ., wall and dome. l Both summer and winter ' vperating conditions are considered. In all { . cases the conditions are considered of long j I enough duration to result in a straight line temperature gradient. The temperatures are as follows: l (1) Summer operation: .
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(a) air temperature inside building - 120.*F
} (b) exterior temperature above El 50 ft - Il5*F (c) exterior temperature below El 50 ft - 90*F l
3.8-103 '
238 N[C[Eh"d kkIAND y ;eg kk. 0 3.8.4.5.1.2 Materials Criteria (Continued) where f; is the specified compressive strength of concrete and f is the specified yield strength of reinforcement. Excessive deformation of the Shield Building could affect safety-related components. Hammering against other structures or seizure of lines are recognized hazards. Maximum relative horizontal displacement between the Shield Build-ing and the steel containment at the steam tunnel is on the order of 1 inch. External joints at the other locations allow the necessary displacements. 3.8.4.5.2 Auxiliary, Fuel, Control, Radwaste Substructure, and < Diesel Generator Buildings Structural acceptance criteria are defined in the AISC-1969 Speci-fication and ACI 318-71 Code. In no case does the allowable stress exceed 0.9 F where F is the minimum specified yield stress. The design criteria preclude excessive deformation of the building. The clearances between the adjacent buili in s are suf icient to eveng imp,act during seismic eve t. } t, % Ado 0 U k seth hkOS % t 00Lrmb Q bl. Y Y hC(5% b u
- 3.8.4.5.3 Other Seismic Category I Structu s Structural acceptance criteria for Seismic Category I structures outside the Nuclear Island will be provided by the Applicant.
3.8.4.6 Materials, Quality Control, and Special Construction Techniques 3.8.4.6.1 shield Building See Subsection 3.8.3.6.1 and add the following special construction techniques.
- 3.8-128
W 220.18 Describe the procedures used in the seismic analysis of the polar (3.7.2) crane. Discuss how you account fcr the effects of cable jerking. r . ,
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w., p.a.q The overall seismic analysis for the Reactor Building was performed
'using an axisymmetric model with the polar crane mass smeared around the containment. See Figure 3.7-24 and also our response to Question 220.10.
The seismic analysis of the polar crane was done using a multi-degree of freedom beam-element lumped-mass computer model. Acceleration response spectra were used as input. c Cable 3erking was investigated using a non-linear, tension-only member for the hoist cable in the computer model of the polar crane. It was found that the jerking stress was only of the order of 1,000 psi. This is a negligible transient noting that the hoisting system is designed to support a static load three l times the rated load. i 1 l 1
l ATTACHMENT NO. 2 O DRAFT RESPONSES TO POWER SYSTEMS BRANCH QUESTIONS f
l , l t . l l 1 430.03 Provide the minimum starting voltage of the Class 1E, Division 1
~ ~ . (C.3.1) and 2 motors. Indicate the minimum difference between the motor torque and pump torque of the Class IE, Division 1 and 2 motors.
during acceleration. Explain the sentence in section 8.3.1.1.5.3,
,_ , _ _ , _ , , part (2) of your FSAR in which you state: "In some cases, motor sizing torque and load requirements are accomodated to limitations imposed i 8
by the circumstances of the system or specific functional requirements." l RA 5 D eW J C. . . Minimum starting voltaoe for all motor operated valve (M0V) notors is 80%, for all other safety related motors is 75%, recovering to
' 90% within 2 seconds in each case. '
kk As stated ingsection 8.3.1.1.5.3(1), motors are sized in a:cordance 1 with NEMA standards including starting, pull-in and driving torque
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rav, 0 8.3.1.1.8.1.2 Ratings and Capability (Continued) The size of each of the diesel-generators serving Divisions 1 and 2 satisfies the requirements of NRC Regulatory Guide 1.9 (March 1971) and conforms to the following criteria: (1) Each diesel generator is capable of starting, accelerat-ing and supplying its loads in the sequence shown in Tables 8.3-1 and 8.3-2 without exceeding a 5% frequency drop. The generators are capable of recovery to 9J% of normal frequency in 2 see or less. (2)- Each diesel generator is capable of starting, accelerat-ing and supplying ,its loads in their proper sequence withoutexceedingahk%voltagedropatitsterminals. The generators are capable of recovery to 90% of normal voltage in less than 2 sec. (3) Each diesel generator is capable of starting, accelerat-ing and running its largest motor at any time after the automatic loading sequence is completed, assuming that the motor had failed to start initially. (4) Each diesel generator is capable of reaching full speed and voltage within 10 sec after receiving a signal to start, and can be fully loaded within 30 see following the start signal. (5) The speed of each diesel generator does not exceed 75% of the difference between nominal speed and the over-speed trip setpoint, or 115% of nominal speed, whichever is lower, during recovery from transients caused by disconnection of the largest load. l i 8.3-21
t l 430.06 (8.3.1) In Section 8.3.1.1.8.1.1 of your FSAR, you state that separate unit
- station service transformers and separate reserva station service t
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transformers are used for the nomal and alternate preferred power supplies for each division. Indicate whether this arrangement is specified by the interface requirements. State rhether there are other arrangements permissible under the interface specift;ations. Indicate why there ( , is only one feeder from the preferred power sources provided for Division i 3 while two are provided for Divisions 1 and 2. I IC 40[ [F/, r/rper" Or/arp niir, .ra Au s..r.i.i. s. t. i <> ~f(fr7#/7/ r/a/M /h 4dIEE916
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} . CESSAR II 22A7007 j 238 NUCLEAR ISLAND Rev. O i -
. S.3.1.1.8 Standby AC Power System (Continued) 1 Rach standby power system division, including the diesel-generator, l its' auxiliary systems and the distribution of power to various class IE loads through the 6.9-kV and 480V systems, is segregated l
and separated from other system divisions. No automatic intercon-nection is provided between the Class 1E divisions. Each diesel-generator set is operated' independently of the other sets and is l connected to the utility power system by manual control only during i testing or for energized bus transfer and then only one division at a time. 8.3.1.1.8.1 Redundant (Division 1 and Division 2) Standby AC Power Supplies 8.3.1.1.8.1.1 General The diesel generators comprising the Divisions 1 and 2 standby AC power supplies are designed to quickly restore power to their respective Class 1E distribution system divisions as required to achieve safe shutdown of the plant and/or to mitigate the conse-quences of a LOCA in the event of a coincident LOPP. Figure 8.3-2 shows the interconnections between the preferred power supplies and the Divisions 1 and 2 diesel-generator standby power supplies. E q: - t e ; .i ; en. d mu in EEC Tr Lirvi,~r: _ - - M pe r = + = = - - 5+ " '- r:-i r: n.i. f rr c r; =c: 2 : d fa b 41, . . . . n9 A
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A detailed discussion of the Division 3 diesel-generator system (HPCS) standby AC power supply is presented in Subsection 8.3.1.1.9.1. I 1 8.3.1.1.8.1.2 Ratings and Capability The diesel generators for Divisions 1 and 2 each have a continuous nameplate rating of 7,000 kW on an 8,760-br basis (with 10% over-load permissible for 2 hr out of every 24). This exceeds the loads required at one time, as derived from Tables 8.3-1 and 8.3-2. C.3-20
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! 430.07 j Provide the following infomation regarding the Divisions 1 and 2 (8.3.1) diesel-generator qualification testing discussed in Section 8.3.1.1.8.5 t { of your FSAR: ! 5 a. i You state been run onin similar Sectionunits. 8.3.1.1.8.5 that the 300 start tests have If the tests were not performed on identical ' (' units, the Divisions 1 and 2 diesel-generators must be requalified
- - in accordance with the requirements of Sections 5.4.2, 5.4.3 and
, 5.4.4 of IEEE Std. 387-1977. i b. The load capability test was conducted in reverse order from our position stated in Item C.14 of Regulatory Guide 1.9, Resision
- 2. Provide justification for this difference.
- c. Provide the test results for our review.
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430.08 In Section 8.3.1.1.6.4 of your FSAR, you state that, the ciesel-generator (8.3.1) overcurrent relay protection has a voltage restraint so that disturbances in the plant auxiliary power system which result in excessive voltage drops, will not damage the diesel-generator. Indicate how far into the plant distribution system from the diesel-generator the relays will sense a disturbance. State whether these relays are sensitive to voltage transients created by normal power system evolutions such as motor starting. b 5 P a w J-4t. f The 51V relays (very inverse time overcurrent relays with voltage restraint), will sense a disturbance at the 6.9KV systemlevelonig h khese relays are not sensitive to voltage transients, created by normal power system evolutions, such as motor star . 4 6his time ovsrcurrent protection is typical for diesel generators division 1, 2 & 3. s \,3._A. - acceM U M mttit 91 "---~d fmus 19mocommj9.3.1.1.6.4 : : ; J. t :!.. ^. % ;). hasbeenrevisedg:.dthE: '
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l 430.06 C',pyy$jpg q q,qgqp[ l l IIIIIIIII Division 3 (HPCS) bus is normally fed by the preferred feeder from a bus in the Turbine Island. It is a requirement that the Turbine Island bus receive power from two independent offsite power sources and that the transfer between the power sources be automatic. The transient due to the Turbine Island bus transfer will cause HPCS bus to be transferred to the diesel generator. Transfer back to the Turbine Island bus may be accomplished manually by the operator. This arrangement insures the availability of *he offsite power to the Turbine Island bus and in turn to the HPCS bus within a few minutes following a loss of the normal offsite power supply. l JEM: cal /K01247 l 1/24/83
4 DUESTION 430.12 l (8.3.1) e J
, The separation you describe in Sections 8. 3.1.4.2.3.1 and 8.3.1.4.2.3.2' of your FSAR for the scram solenoid circuits and the main steam line
. (MSL) isolation valve circuits must be justified by analysis, based on tests, to show that there is no detrimental effect en Class IE circuits, trith which these circuits are run. Additionally, demonstrate that the function of the scram solenoid circuits and MSL isolation circuits will i not be impaired by this arrangement. Explain how isolation is maintained between the Class IE power supply feeding the "A" solenoids and the non-Class IE power supply feeding the "S" solenoids since these circuits are run in a common condult. -
fi,'. Explain the use of the D1 through D4 inputs shown in Figure 8.392#.of your FSAR, coming via isolators into the load drivers of the "S" scram solenoid circuits.
RESPONSE
The scram solenoid and MSIV circuits gg,e run in separate conduits. Cables from other circuits are not run4these conduits. Within the PGCC the conduit is flexible and since the circuits are non-divisional the flexible conduit is routed within non-divisional PGCC ducts. There is no mixing of divisions. Optical isolaters have been provided for electrical isolation within the panel between IE and non-lE interfaces of the logic circuit. Tne power supply feeding the "B" solenoids is of the same type as the one feeding the "A" solenoids. Solenoid "A" is fed from ron-lE bus A via an inverter and an EPA. Power is maintained within IE parameters and the equipment is used for the power supply systen is of high quality, + 1/2r voltage regulation. Solenoid "B" is fed from non-1E bus "B" power supply similar to bus "A". ^ It is acceptable to run "A" and "B" solenoid power circuits together since the isolation is provided in the logic cabinets. Separate non class non-1E power supplies are provided to enhance plant availability. Figure 8.3-25 will be corrected on the basis of the above discussion. See attached marked copy. For more discussion of the N5P5 Power Distribution refer to Question 430.14 response. 1
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P ! GESSAR II 22A7007
- 238 NUCLEAR ISLAND Rev. 0 4
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QUESTION 430.14 state in section 8.3.1.1 of your FsAR, whether the nuclear systen pro-tection system (NSPS) non-Class 1E power supplies which feed the "B" scram solenoids have a separate and redundant Class 1E protective package ' installed between the power supply and bus consisting of overvoltage,
,- undervoltage and underfrequency protection. If not, this package should be installed to protect the solenoids against a condition which could fail thee in the unsafe direction. Discuss the susceptibility of the load drivers to power supply anomalies such as over/undervoltage, over/
underfrequency, voltage transients, voltage spikes, EMI and harmonics. The protective package must provide protection against any conditions which would fall the load drivers in the unsafe (i.e., shorted or closed) direction.
RESPONSE
Current NSPS design by GE consists of four independent Uninterruptable Power Systems (UPS). Each includes a battery charger, 125 VDC battery, an inverter, a static bypass switch, an Electrical Protection Assembly (EPA), a 480-120V regulating transformer, a manual bypass (inverter bypass) switch and essential metering devices for the UPS. The EPA consists of over-under voltage and under frequency monitoring circuitry and an output disconnecting device (circuit breaker). The EPA provides over-under voltage and under frequency protection to the RFS loads and related power supply associated with the scram pilot valve solenoids and the MSIV pilot valve solenoids by disconnecting the bus from the input power whenever the UPS output voltage and frequency deviates beyond the preset limits. These limits are to be established by field calculations. The EPA is designed to meet the IEEE-323 and IEEE-344 l l requirements. l A second supplv is through the 480-12CV regulating transformer and the isolation transformer. The regulating transformer provides voltage regulation of ib%. It is furnished with a harmonic distortion filter to reduce the maximum output distortion to 5% RMS. The regulating transformer is virtually immune to lightning or transient surges on the line and has a low audible noise level of 65 dB at 5 feet. The transformer is qualified for IEEE-344 and IEEE-323 requirements. MP:csc/110146-6 11/4/82
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l The power supply through the bypass circuit, including regulating trans- ; former and the isol,ation transformer is also monitored by the EPA as described above. The new load driver cards are designed to meet the 4 requirements of IEEE-472 and are capable of operation within voltage variations of 24-200V AC or DC. They ofso opede oh squarc wave fukes and are therefore
- insensi+i ve in har monic vol+ ages frorn thest power sug ly.
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430.31 Provide the following additi::nal infomatten regar' ding the protection (1.8) derdi;; "; pre 6Cer, cf contafrument electrical penetraticns:
- a. You indicate in Part I.2.13 of Section 1.8 of your FSAR that an analysis is required for circuits nomally protect 6d by small fuses
~ or breakers such as control circuits, alarms and solenoids. Provide this analysis. '
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N.0*3f b. In this same portion of the FSAR, you also indicate that where
- very low currents are invnived such as in instrumentation circuits, thermocuuples and annunciators, no action is required and that confomance with the provisions of Regulatory Guide 1.63 is - - - - - -- accomplished by inspection. Explain what is meant by the phrases "no action required" and "conformance by inspection." It is our position that if the fault current available from these circuits is greater than the continuous current rating of the penetrators.
the penetrations must be protected by at least two fault current interrupting devices. I Rw - M M _ . . . .. L m r> g n z. z. ia . i . _em p e. e a e WW e * * *
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4 3o.31 238 NUCLEAR ISIAND. R3v. O l 2.2.13 gRulatory Guide 1.63, Revision 1, Dated May 1977
.: (Continued)
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' of the redundant protective elements so that no event causing a need for the protection can disable the
,, protective function.
i 1.2.13.1 Analysis of circuits penetrating prima containment. A. 6.9 kV cricuits by two for recirculation circuit brenkers pump in series. on motors the are protected , C,9/< v Wopg circ un The recirculation pump motors are also fed from the low frequency motor generator sets. This feed is directly protected by one 6.9KV rated breaker. If this single breaker should fail to trip on a fault within containment, the penetration is capable of with-standing the maximum theoretical generator fault current for 250 seconds without damage. The fault current which could be as high as 13.5 times rated current would be terminated in less than 250 seconds by one or more of the following:
- 1. Overtemperature failure of the generator output winding.
- 2. Overcurrent in the field circuit.
- 3. Tripping of the mg set drive motor breaker on overload.
B. Power circuits for motor control center loads are protected by a circuit breaker nnd a fuse per phase in series. The application of penetration wire protecting devices is shown on the MCC. single line diagrams. C - MCC control circuits have a single fuse per control circuit, , which we consider adequate to maintain the electrical penetration integrity. I Our conclusion is based on the following study. 4 We usedaNEMA size 4 starter in this study as the limiting case, l since larger size starters have auxiliary contactors and smaller l Control Power Transformers (CPTs). For a NEMA size 4 starter, ) l the standard CPT size is 250 VA which means that the rated l 1 secondary current is 2.08 Amps (250 4 120). The field cable used between the CPT secondary and the containment penetration is size 14 AWG, which is rated for 22.75 Amp continuous at 40* C (104' F) ambient. l l l
Tho maximum chort circuit curgent thnt tho CPT can lot through, cocuming 54 impedanco, to be.5onccrvativo, io - 250 VA -..
.0Sx120 = 41.6 Amps By ignoring the field cable impedance, (conservative), the l
maximum short circuit current at the containment penetration is 41.6 Amp. 2 The 1 t curves of the Westinghouse penetration indicates that the penetration seal can carry 60 Amps for 1000 sec (1.6.7 minutes) and mainta)n its mechanigl and electrical integrity. Tid pe-nstm//en
?l4-W.-;O pistail ui M can g carry 110 Amps for 10 seconds without insulation damage.
it M Corsc /uded fhot' From the above, the CPT will fail much faster than the' field cable,. the pen:t_ 4 ti'= pigt:iL^ or the penetration seal, should the CPT secondary fuse fail to open the circuit under an overload or a short circuit condition. The CPT may fail and act either as an open circuit or a closed circuit. If it acts as an open circuit, the penetration will be isolated. On the other hand, if it acts as a closed circuit, the MCC feeder breaker /or fuse combination will trip and open the circuit, isolating the ces.cnd /c ven t of };u n r cpenetration. Tha /amv,v,,nl r/c vi-re<o r f m tet twn Based on the above, the present design is considered to be adequate to maintain the electrical penetration integrity.
- 125V d-c instrument and control circuits will be protected by a 2-pole circuit breaher and a fuse in series where needed.
- 120V a-c instrument circuits and space heater circuits will have one single pole breaker and one fuse in series.
D. Specific circuits, having a limited power source, that cannc produce any short circuit current, damaging to the conducto2 insula, tion, do not require a protective device. Included in these special circuits are:
- Thermocouple circuits
* ' Shielded cables for low level signals ( 4 to 20inA - LPRM, IRM, SRM, RPIS instru mentation circuits)
' Annunciator circuits I
- , 2.2.13 Requictory Guida 1.63, Mr.vinien 1, Date
- d M7y 1977 $
4 t f (Ccntinund) l , i i- - Sununary Table of Conformance with Regulatory Guide 1.63 y for Circuits Penetrating Primary Containment t Very Low Currents Use of Two Involved Interrupting Devices in Analysis
- Mc '.: tion L wired -
Series C07.fornir.00 by
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c Cl'YCUllce *Yeql$ Power Circuits on ' SClf motor control X go/(c//Wf . { centers . l Control circuits, alarm, solenoids, etc. - circuits, .; ( normally pro-tected by small fuses or breakers X 1 Instr'.unentation circuits, thermo- . couples, annunrA6 tor - all low-current-level applications X (
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c. 4 Provide the fault current clearing-time curves of the primary and secondary current interrupting degices for the penetrations plotted against the thermal capability (I t) curve of the penetration.
- Our concern in this matter is the maintenance of mechanical integrity.
Provide a simpif fied one-line diagram showing the location of the (" protective devices in the penetration circuit and indicate the naximum available fault current of the circuit. If the overcurrent protection is not fault current actuated. identify the power source to the trip circuits. It is our position that the power source for the primary protection device should be from a division different from that supplying the secondary protection device. I eme.ums-
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' JU3 M 4 CUSTOdER PAGES 0 PAGE l APPARATUS Jan s 312.- P DATE BY ITEM S QuetTION 430.36 Identify all e'ectt ical equipment, both safety and non-safety, that may become submerged as a result of a LOCA. F: all such equipment that is not qualified for service in this envircr.:::ent, provide an analysis to determine the following: G' a. The safety significance of the failure of this electrical equipment (e.g., spurious Ac.turtgon or loss :f actuation i function) as a result of'
- o b.
uerf9 . The effects on Class 1E electrical pewsr s: -ces serving this equipment as a result of such submerga..:e. '
- c. Any proposed design changes resulting froc this analysis.
FES PONsE 4 30.36 Au, op THE @ El. ECTR.ICAL EdutPr4ENT, BOTH SFFETy AHb HCN-SAFETb TipTt*1Ay EGC6W1E SU6 MERGED hs A REsvLT OF A Lc:A ARS IDEt<TIFlib IN
*rpgt,E (A)g . AN Analysis FOS ALL sueu EaulPt:EHT THAr is NOT GVALIFil F6R SERWCE IM THis ENvtRoNMENT ss ALso PRevitik IN TAst,E (A)&
THE ANALy$ts biTGRMrN&s THE FoLLChilN6 a:. TnE SnFETy
$1&ntFlesucs of THE FAILvPS :5 TMis ELECTfkAL TCu (e.g., SPURT 00s ACTUATICW Q Loss OF ACTul,71CH ConcTicW) A: A REsuL on svaMeusNes .
C
- b. .
THE EFFECTS GM class LE ELECTfl CAL P: win sovReEs SERVIN EQutPMENT At h RESULT OF Cutit 50 AMER Cents .
- c. Any Psoposep DEstart CHANGES REtutT:n& facm Tuis ekstysis.
Seconc/ suhmith)
\. Re vi s e pages SW C, E.3 4 44 fem Ta A /e A a re cd fac hec /.
1 l l
I m a Il [ IEInER SS-4
.^f 08
- EUIN DATE: 12/10/81
- REPORT lli ALL CABLES SORTED BY BUILDING.El.EVAT
- TODAY'S DATE IS 10/27/82 1 tilV HOW Calli E NtiP51ER f 4/AtfD '
TO/FROM DESCRIPTION TO/FR"JM EQUIP fAMII , a tti 07 42 uaJA-2 NetVI B33 MTR COND BOM B
-01 07 42 A B33 C00lBCB -
IS-NOVI B33 TitENN0 COUPLE J BOX B B33 CODIBJB .
-On 07 42 Al IIS-NOVI #
-01
-01 07 07 42 A -NOVI B3' ~ ^ '
- u - ~ ' ' ^ ' ' * * * /Y'O MO 42 -833A-2 -NOVI *C"'^^^"""R ' ' -
TOTAL QUANTITY THIS ROW IS I TOTAL QUANTITY THIS ELEVATION 781
-el os se --
- -.-02v2 SPCU SEC CONT ISO VV G38 FF045
--s e ce : yx - _ c::: :.:= ::: ::::- : -- ::: rve- [d$ yg,3 TOTAL OUANTITY THIS ItOW 3 TOTAL QUANilTY THIS ELEVATION 2
-01 09 17 Al -EbessW99F-D2V2 SHUTDOWN MAN SUCT VLV E12 Foto .
[f$ -. TOTAL OUANTITY THIS ROW I ~ a TOTAL QUANTITY THIS ELEVATION I
-02 00 34 55 - D2V2 CRW 1/8 ISO VALVE
-02 00 34 A PSS FF021 25-D2V2 .mau. _" _: -- ~
-02 00 34 A -D2V2 ^
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- - 7 0::
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-02 00 34
-P55 .1 2V3 ^'
i .m. .~.m _ " 000 TOTAL QUANTITY THIS ROW 4 TOTAL OUANTITY THIS ELEVATION 4
-02 03 36 -HOVI SOUNO PWR PilONE JACK RBI J7 Il go go TOTAL QUANTITY THIS ROW I
-02 03 41 -RSI -NDVI 02 03 41 A SOUNO PWR PHONE PtlLL BOX RSI P30EA 223-NOVI -
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-02 03 Al Al 2G-NOVI :- - ' ag gg
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-02 03 di H58 - NDVI SOtlND PWil PilONE JACK NOI JI7 13 l
'9 TOTAL QUANTITY THIS ROW 4 *
-02 03 A7 m -Nov1 SOUNO PWR PHONE JACK RSI Je 8 sq
' N l TOTAL QUANTITV THIS ROW I
-02 03 55 h m ~ NDVI SOUNO PWR PHONE JACK RSI J8 7 It i
TOTAL QUANTITY THIS ROW I . l TOTAL QUANTITY THIS ELEVATION 7* i -02 05 is - -oiva vtv C4: r00e C41 F00s y'C5
j -
.
- r 't t
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-tM HietER 50-8 CA*'LE Ph..s%AN
- 68 tW BEGIN DATE: 12/10/89 * ' *
*I REPORT lli All. CABLES SORTED BY SUILDING.ELEVATit
- TODAY'S DATE IS: 10/27/02 i
~
1 LOG ELEV ROW CABLE NUMDER
...................... TO/FROM DESCRIPTION
........................ T0/FROM EOUIP IAE8f*I' I'#dM* 8 i
y, . TOTAL QUANTITY THIS ROW 3 '.
- D +04 08 SS """
^ " ,02V2 CONDENSATE SUPPLY CONTAI P48 FFl83 l
TOTAL QUANTITY THIS ROW I i O +04 08 85 -833A- -NOV4 NONOlV PENET P 201 O +04 08 65 83 ' ^ ^ ^ ^ ^ ^ ^ - - - T ^ ^- - - ^- " INBO RSI TT2Ol I D +04 08 65 60 " ' _ 7 0 0 M "r-- i- ~ ' ' ^ ' , D ' " ' ' ' '
+04 08 SS -60 ::2. - ^ ^^
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TOTAL QUANTITY.THIS ROW G TOTAL QUANTI TY THIS ELEVATION IR . D *06 00 32 ~ ' """' "- . ALVI T-C.DRYWELL , , , ,, ,E3% M0178 d*
. 6' ,
TOTAL QUANTITY THIS ROW I .
- t O +06 00 45 enhumuAdemmmesSIP-DIV2 DW O/5 150 VALVE D *OG 00 45 ^ ~ ^ ^
^^'2^__ P42 FFl03 e
~ ^ 2. 2 ._ - a "'". - ;"^ e TOTAL QUANTITY THIS ROW 2 TOTAL GUANilTV THIS ELEVATION 3 3 +0G 01 42 P50 02VI SUPP POOL WATER TEMP 3 +06 01 42 H D2VI P50 NNOOSS j 3 *OG Ol 42 SUPP POOL WATER LEVEL P50 HNOO38 D2VI /gg SUPP POOL WATER LEVEL /vgg P50 HNOO48
, , e i TOTAL QUANTITY THIS ROW 3
- s 3 +06 01 43 N .D3VI .
- SUPPRESSION POOL E22 N0550 A '
Yg - j TOTAL. Otf4NTI TY THIS ROW TOTAL QUANTITY THIS ELEVATIOH l 4 lM i ) *OG 04 , f
) +0G 04 35 35 N O2V2 DRW I/8 ISO VALVE
- _. . PBSFF02S") # . .
1 +0G 04 35 . - - - ^^ -_
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- I TOTAL OUANTITY THIS ROW 3 i TOTAL QUANTITY THIS ELEVATION 2
; ) +0G 07 ' . 58 P55 - 02V2 $
t +0G 07 58 " ' " ' ' ' CRW 1/8 ISO VALVE
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- tw NI21nER 50-8
- tw CEOlH DATE: 12/10/81 C3 CACLE PI JtAM
REPORT 111 ALL CA t.ES SORTZD OY Cult.clNo.ELEVAfl0N . TODAY'S DATE 808 10/27/82 *
- I -
9 TOTAL QUANTITV THIS ROW 3 . TOTAL QUANTITY THIS ELEVATION 8 - G el3 08 34 N D2V2 REAC WTR SAMPLE VLV S33 F018 A , TOTAL QUANTITY THIS ROW
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N .5 I . G +13 OS 43 h - alV2 s *I3 OS 43 M IV2 REAC WTR SAMPLE VLV
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1 TOTAL QUANTITY THIS ROW 3 * ~ TOTAL QUAMitTV THIS ELEVATION 8
- 18 +14 00 28 N MOV3 R8 MAINT HolST MONORAll T33 EE006 .
g t TOTAL QUANTITY THIS ROW I , # i ! i 2 *I4 00 ~ SUVI
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- B *14 00 di RS' 223 NOV9 l SOUNO PWR PilONE JACK RSI Je 12 j 'O *ld 00 41 -225 HDVI SOHND PWR PilONE JACK R58 J22 12
] S *ld 00 di i -225 NOVI ' ' ' ~- - ^'" 2. ] :D +14 00 di -225 NOVI '^ ^ ^ - - - ~ ~~ m i :D *I4 00 45 -225 NOVl _ . j in +14 00 S0tIND PWR PilONE Put.L BOM R5i P30AB , 4l H 22b NDVI 1 :B +I4 00 41 -8151 HDVI . -.. - , yun 1 2 +I4 00 41 -RSI NOVl SOUND PWR Pil0NE JACK R58 J22 14 s +14 00 di 1 -R59 . NDVI __
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- A 1 :S +14 00 46 ja ""'
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f TOTA:. QUANTITY THIS ROW I .
- tB +14 OD 50 ** --^
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- yt1 SS TOTAL QUANTITY THIS ROW 2 i O +14 00 53 4'O'8945MEE'S 02V2 WPS CONT ROD DRIVE SEAL P60 FFOSS 1
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CucST/cN 430.31 Provide a listing of all motor-operated valves in your proposed nuclear island which require power lock-out in order to satisfy our single failure criterion. Indicate the features of yoar design which permit you to satisfy this reautrement. RE.rponse 4.30.3i . Acyom A die yusa% i.s ise/a /c/ in 3,4a.rt , 9.s. s.3, .u raty Eva h sti.u, a.< / tr.1.S> Ina freme<%.fi*~ depimf.r, 4e h 1Xere leiay a/y e n m.for poenfed va/ve which reyatres powev /ekmt fa ss fi.ryf W e .rie Silurs criferim. For more / fa]ls .s = < *
/<.yeawa 42/. 0z % .%j/e Q) ow orP Ics2 is(Ara).
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. . ., V CESSAR II 22A700 238 NUCLEAR ISLAND R;v. $1P ~
( 9.5.9.1.1 Safety Design Basos (Continued)
) .
Code Section III, Class 2, Quality Group B and Quality Assurance D requirements. uy eenfaime.*Y pensfre$enA,inlsYieu 4lvu, an*lf{f! *y up to tiesa v.hA (2) (%p4tg. Q@goetTatc!ht?tieb'sdMaantztir}nten$?ht/ designed to Seismic Category I, ASME Code Section III, Class 3,
, . Quality Group C and Quality Assurance B requirements.
(3) The deep-bed domineralizer is designed and fabricated in accordance with ASME Boiler and pressure vessel Code, Section VIII, Division 1. (4) The two horizontal, centrifugal SPCU pumps are designed and fabricated in accordance with API 610. (5) The remainder of the system is designed to the ANSI D31.1 L Power Piping Code and Quality Group D requirements. 9.5.9.1.2 Power Generation Design Bases (1) During normal plant operation, the SPCU System is designed to recirculate approximately 1,100 gpm of suppression pool water. (2) In circumstances of high suppression pool activity or conductivity, such as following a blowdown transient,
. the system is capable of providing cleanup of the suppression pool water at the rate of 2,200 gpm.
(3) The system is designed to maintain the suppression pool rg water quality at or better than the following conditions: %v - Conductivity 1 10pmho/cm at 25"C [ , Chlorides 1 0.5 ppm 9.5-39
(14MAnJU 238 NUCLEAR ISLAND R:;v. i 9.5.9.2 Systems Description (Continued) - Dj Jl
- In the event of a LOCA, the SPCU System function is automatically terminated to accomolish L..a i.,ees) containment isolation. Power for the SPCU System pumpFis supplied only from the preferred power buses.
Containment isolation valves are provided with Ciass 1E preferred and standby power. . The SPCU Cystem, consisting of piping, valves and instrumentation,
~ is shown in Figure 9.5-18 (K-172). The system has no unique major components.
9.5.9.3 Safety Evaluation , The system has no safety-related function as previously defined. Failure of the system does not compromise any safety-related system or component and does not prevent safe reactor shutdown.
~
However, the system does incorporate some features that assure reliable operations over the full range of normal plant operations. These features consist primarily of instrumentation that monitors and/or controls SPCU operation and performance. Portions of the SPCU System that penetrate the containment are provided with isolation valves which are automatically closed by
, an isolation signal.
The containment isolation signal logic receives rea: tor low-water-level signals and drywell high-pressure signals. These inputs
, isolate the SPCU System to prevent containment bypass leakage. l Aimed)
Emergency power is supplied by Class IE buses to isolation valves andleakdefetioninstrumentationfortheDBAandforLOPPevents. A p .r L , e f fle .5/cu. Syatw ;Lfye,,afnfes e>,lr We .s*c=" Arf c~hI"*el l' \ pr. 14 / or1% eu is lafr. n/ve./F038, nireA is eerm e Asel=lf! if.s p**'*r afia<=*Y %
=s fl* ,e w a .f /esi .e .:n.r1 a' .1 le S; fare 11: If << se on ""lesir=lf< . J e- y,eae f asfi.,,Qa 'yloj ear r is. M u 1/e. syr e~~ nfmeJ h A4.,- C4:.r .1c~ due:
se
- q. uaAry a,, <. a u d1,.,,,e.:f '/yp:a&. 74is v.
ss /e,Ka y e . Awer is syf e ly i a s., ,.. n. i q .g , i
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,,, bLdshu 11 J4A/UU 238 NUCLEAR 'ISIAND Rov. QJP
. e I
-- \
9.5.9.T. Instrumentation Requirements (Continued) !
. turn it off using a hand selector switch located in the control room. b'5Qbspep]l n5M4%36fd($4tgrf.1;s$] *
]
The' containment isolation valves are supplied with position indica-tion in the control room and remote-manual as well as automatic oper g (Sne_ Subsection 7. 3.1 for_ details. ) - gsu_ation. 9.5.10 e st. a . t. a .s.,.s a s. > Nuclear Island - DOP Interface z 9.5.1'0.1 Nuclear Island Fire Protection System - BOP Interface The Applicant shall provide the water and CO2 supplies for the Nuclear Island Fire Suppression System. . 9.5.10.1.1 Design Criteria Fire water supply for the Nuclear Island shall be provided by the BOP Fire Water System, Essential Service Water System and Conden-sate Transfer System. The Essential Service Water System provides albackup Seismic Category I source of water for hose reels for essential equipment. Condensate is the preferred source of water for the Wet Standpipe System inside the containment. The ESW and condensate connections and actuation and isolation provisions are within the scope of the Nuclear Island design. f CO2 shall be supplied to the Nuclear Island at a sufficient rate
~
and duration for the Diesel Generator Building CO 2 Fire Suppression System. ( The classification of the BOP Fire Water and CO 2 Supply System for the nuclear Island may be Quality Group D and nonseismic Category I. 3 i
. .; . %1 J
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- C F CRAUN O CO j "- JCB NOTES CUSTOMER PAES PAGE 8' APPARATUS JOB I DATE SY s ITEM 4.5.1. $ .[as'fYubeekYaki.k Atfuireareesif shfuuel m - n . , n ,... a. . i. s . , a - ,,. a. i
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- 81. Fmilnsa h. J.f4 de 'Sif f. (n.<:lteu " sexos auf rWe %.desira/4 f..d }
seuse . f c.s y .e e. fs i.; s/edrisJ syrfews ia./u/iy osluss ../ offer f/u;af .17sfes a..y.oreda sem c.esidereJ in de:tyainy .,.4.s- siy f e ' .{ failav<> cre~ ,da.<j/ #s v.a/ve. er af/er ff' aid sp7tw e yeare f y a t to ea //e/y.- 7'a flueff i., ayiren safe /,, y ei.,freaJ sayae,<o.
- 82. For es/ve McJe if was /sterariae/ fja' Gifure of an efedrialpsfer
<.y.peaf aa >< aa u.se uo/esired s.,,.Jaori / us.fr.or .-f 7'Je w/>e aa/
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fe dize.,eeedp . er 1i rJe ym./vw.*a eleefria. m./$r opeafar. 7:is Artes is e /~l-mu,le/Jey.foede/ A..,4ymfd Aa.-p sifra'; me.fiuel-eo,afaets sudlcb w$ic$ i.r e**ultecf*lbtfuttM YL Yuffth t'ever2h'*j4 ' Yor .5karYtr a / ifs ayew'.s.r met.r.ris suaifali 1.s a.rmJly off sahr<A *%=*ufs
$t 'f80 VAC fSf ewerf f/t re/ve yrwfors oe.Y.r, .n or al .'af.soinkyrf 1 0 1$s /20 VAC ff fr.wer I6e 7$e p.nbrt ttvers?.Jr n,*T$r .skntferJc.or7WJ.,
Ys f nsef /As$uiss{ fCEAcalio64s de7llin./ult f/I.s altctronll-yeeJed r /ve, d fde yepire/pesir%e .F ffe' n /ve,, to w4re4 (' pfe. murkes.<e<t'-9< tr>*rovJ .F e/eefetep wer ir yphedsa ader fa s ferfy#e .sla;/r Gl{ure eriferio <. *ife u.filify 77d* s e *fu;// Le refaire/ f,, a.s yy. t 83 D'is toorti.., do <> u.t .ppty aroaae 141, atae.nte /p epe,,faa/ yates is or.7'e/sssi9e/ a.s a s, "a. dire. vs.lv e.. 8Y. .fi6ece Wt siylC (E((anre'cyifeyr~ene Isstf*g((af i pejeneyn / .f'glseftitaL! n p.wey fm fJe n/vs .4rcride/ u 'B2.f </*ve. , //e v.n h e A a.s h u a ,
, sets e f p asifr i,,h e.a % .s i.< fde x<< ia e.u /r< / roa ~ a./ //e jeaib*e.risea(icAllIwos keonr.rsllye3 )geeA:.f t'/t'. gjo sg -Sr(fuVt c yof yicJr.
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p V.g;, '. . Jos HoTE5 cusio<tR PAGr$ PAGE 8 APP'ARATUS JOB DATE BY ITEM f^ 9..r.f.5 Ins 6 n.ef fim depairemb (eufinud)nu :, ware 4 i.s ocpawandy frJ.e a..ae.faJ LJf,y J..kat. zu .stiefa da sa au,<unc.;af. yare,.in44..yss. M ..,,... / / to a e.- ,t. sa w <. ein.;i. H de sein y<raf.r ta su.ve/ fv'. #< 6.If aloself siflu, l ar 12fs cw .rwtfcA 1s nr.ved f. tt's., oA/pos/fi.~ Ma a .,. einforpr./aee.s er ..diL/e a.u/ vi.,ud i, A.tr.u. \ gne sa a li A di.~ <.es stan .awrfd . rVe es tra a.. <./</ f. tudi an,<y /,y46 whteA ara Ac powered, .t,a., / yesa /. 4 eked:zf M., omke 13 u<aal Gws fic.6//y ./.seJ p aciy% ths Gffsp.,/aes r,su / a - ti... . n .- . ns arrenite a N seyreytd hy e y riu. sfy% ed uarny r.u16.r ar rye airw th. 9 ec. n< phea se *</<Jrics0 7 y->de/n/ve.r laa/a/es A.7% n/uas ywnW direJ/y 47as, Vaef>r%I /avie e (a.y.,a ,, der ye><A-/ v./ve av
- t a. adeu.rd-eparded sJv<-) of 7f..se _ va7ves p en/e f;-.!n d, h .u a/eefruIdevie.e e A u dee. Aaee air sup,s,fy i.s e .</>-s//ed(y(d.4., .u decnresair-o,l .s.Ax.id udee.). .
T , ( . f
.;... l
l
. 1 ATTACHMENT NO. 3 6
DRAFT RESPONSES TO EFFLUENT TREATMENT SYSTEMS BRANCH QUESTIONS 1
/
I
- 460.10 Add sections for effluent radiation monitors and engineered safety (3.2) feature (ESF) filters in Table 3.2-1 of your FSAR. Also add to this table, under appropriate sections, the recombiners in the off-gas system and the process radiation monitors themselves.
l (i 3 R epows< . . . _ I I EW/aent ndi.7*.'.o. yaguy i.r ~4 e4/in Gr.y X,hnen A4 f;*~
- A.% .yr sba, .fp r.ua 3.2-/, e s F fay te r.r a r e. In G ~p XXXI,
- . st. dly Gas Tre<%ed Jpfam,.4 7 4/e 3.z-i. The rea liaer.r te de. efy'~ys . ryas .re sa./.ded an% rW. prusan resssh i G~y XM, ofY y a 3,tw, 7 .f Taki 3.2~/. IX*- pnsass r*hJk mdiiv.r twss/ru .re raf./e/wiS ru eActria.J me/~fu in Gr*y k A.s<as 44. free A%il'v .psfen., of Talk JC2-/. RecoA n e,v-s
(~ u tw ch.dr d e r sb 9 uM Gvowp XXX . B w, l ohwO Es pru n N n, 4-L<. % ; d To. L L. s . 2. - ) . 4
-N^ mhe 4 =e-m--e +.1 e e
(
. _ . , , - . m ,
w g-- p s, , l Tablo 3.2-1 . I l l
, EQUIPMENT CLASSIFICATION (Continued)
Quality Group Quality ; Classi- Assurance .seissio - Principal Component, sareeg Class Iccation, fication, Requiressent, Category, Cossments I-X i l P h se RadiationuJu yenoua s0 == sktr5eenwmW& _
- 1. Electrical modules w : '14 san f 2 Aj 3M N/A B 1 [
--.r.a-% j
.yentilstion monitqse- Tenc//.ef Gn f. le.1 m en 6?s s)
- 2. Cable e-ine4W:dteamlinhnt 2 A,C ,X N/A B 1 y j mentainment=rejttITZMih r= ff se
- isantters /Lu rein .g I
- c
- O l' y, '
nn M C# l XI RHR System y$ g; w WW t:
- 1. i l Heat exchangers - primary side 2 A B B I ww se H g.
- 2. Heat exchangers - secondary 3 A C
- 8 I I side .
o j;
- 3. Piping within outermost 1,2 C A/B B : (g) isolation valves [
! 4. Piping beyond outermost 2 A B B I (g) . isolation valves -
- f i 5. Pumps 2 A B B 1 i i
- 6. Pusep motors 2 A N/A B I l 1
i w I 1 7. Valves - isolation, LPCI line 1 D,A A/B B (g) Ww l I e 3r 4 < -J
- 8. Valves - isolation, other .o 2 D,A B -
B I (g) g j - i - 4 ' !: l . J
! l 1 i-4 i
, o e . ~ , -
f-f- ? '. rm ' I 6,
'...i Table 3.2-1 I
.j EQUIPMENT CLASSIFICATION (Continued) 3 Quality Group Quality Safety {
Principal Component
- Classi- Assurance seismic d Classb Incation# fication Requirement
- Category Comuments XXIX -'t (Continued) .
- 12. Air ejector equipnent Other T N/A N/A * '
N/A g ,
- 13. Turbine gland sealing system. Other T D N/A N/A components ", ,
M i w
- XXX Offgas System .
W g ; u 1. Tanks Other T D s cc i 8 N/A N/A (p) Oy l U 2. Heat exchangers Other T D
. En i N/A N/A (p) g$ 3
- 3. Piping T Other D N/A N/A (p, q, s) m 4 Pumps T b I Other D N/A N/A (p, s) *g ,
- 5. valves - fleis control Other T D N/A N/A (p, q, s) *
- 6. valves - other Other T D N/A N/A (p, q, s) -
- 7. Mechanical modules with safety Other T,A D N/A N/A (p, q, s) '
function
- 8. Pressure vessels Other T.A D
, N/A , N/A (p) 95N/ f 9. RecoQuers O&er T D N/A ti/A (P) [h
~
W
%e
Table 3.2-1 EQU1PMENT CLASSIFICATION (Continued) I Quality Group Quality Safet Classi- Assurance Seismic Principal Component" Class Location fication Requirement
- Category N nts XLIX (Continued)
- 2. Cask crane 3 R N/A B (X)
L Emergency Lighting Other A,C,X N/A N/A N/A w w so LI Heating, Ventilating, and Air Conditionino Systems Q W 5T\a e A 6 O.to
' O Y
(Applicant to Supply) 2 - fh n y ,, LII ECCS Equipment Area Cooling $N { l. Fans, flow meters, ducting 3 A N/A B I valves, heat exchanges, chilling units with safety function .
- 2. m tors 3 A N/A B I
- 3. Instrtmentation and controls 3 A N/A B I j with safety function i
- 4. Electrical modules 3 A N/A B I u
j 5. Cable 3 A N/A B I gy
- .o S
j , i l 1
Table 3.2-1 - EQUIPMENT CLASSIFICATION (Continued) Quality i g Group Quality Safety Q ;' Classi- Assurance seismic - '. Principal Component
- Classb Location c ficationd Requirement Category Comuments =
O .'
- i
, . L 1. Heating, Ventilating, and Air l Conditioning Systems 7 0
$ =
- 1. Control Building
, p
- o. n 14e r., 3 x N/A 8 ;
I gg b . vo lves d., d'ek
- c. c.a Lt4 w,4L s.F.h Acb.-
3- x x N/A w/A B a r. I gj - 3 y g d b a horel w-ekev 3 % N /A B I , p 4,
- 2. Diesel Generator Building 3 S N/A B I cl. H .
- 3. Reactor Building shield annulus 3 C,E N/A B I th O -
2 -
- 4. Drywell Other D . N/A N/A N/A Q g$ '
- 5. Containment Other C N/A N/A N/A *
- 6. Radwaste Other W N/A N/A N/A .
- 7. Auxiliary Building Other A N/A N/A N/A
- i I '
l
/
s
. . . . . . . . .. o. . . . .
( 460.11 Provide additional infomation on the following items for the ESF (6.5.1) filters of the standby gas treatment system (SGTS) and the control , building:
- a. State whether instrumentation for measuring flow rates through
(' the ESF filter systems will be provided in accordance with Regulatory Guide 1.52. Revision 2 (March 1978). Resp,wsc. Overall filter train pressure drop is monitored byAYndicator and recorder in the control room. Individual components inside the filter train ( demister, Ire-filter, HEPA filters, and charcoal adsorber)pressuredropsaremonitoredby*dndicatorlocatedin local panel. These pressure drops are monitored for each filter rpin. cce,d c.o wdk Ra3d Ade GwiA.a.t.52;4 m[w4M mi flow indicator and recoherg n i the control room for each ( - filter train as well as alamsg will l,c a M d wsWiMiEhW61 arms for the pressure dropswdI al.so L swedw - m. uo . n W c, J a e ) . (
- es .g
-m,
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, s , , , I 460.12 Provida icf;reation ca sourca terms for the following items: ) - (11.1) ] 4 Provide the appropriate data for the items listed in Chapter 4 4 s j of NUREG-0016. Revision 1 (January 1979). For those items for which j% i information has already been provided elsewhere. cross-references j ' to the applicable sections are acceptable. 3
- . Rese.nse c ..,2 a.
1
~~
- 1. Re9aested inf.,mation t.* eenc(enside/#,e4mte,.
systewi ir rer,*ted is reep. hse to gaaet,6= ( 4 w is c. l 2. bera.ge" d ily 41..a ra.tes 4 5 1. e,nd etivity, ! he~qli c.e Adnetivt ty ahd ckamical wa,$t,e is ' ~ et patted. (n respense to gueg1'f,n 4rgo.f3 g, nog. Af CAe speejFe ace. sic. inymit C4h ecCWA at va,r. s times a.nd 4.> slie,1 dava.tiens. ne deity '
~
St ws rep. ted repend the expected itea>11 iggin,t3 a.verd.9dd en a. da*tly basts.
- 3. Pc A S. , h a.) . $ c as a Sola kte/rw:J4
'I 4,grion pr. nets a.nl a.c.fivatiba ydedudt of ggpiJ uraste Ave vegevied ih T'Att.E I K.Z-II
- 4. Desqh base cN'S fe" rAdwasic proceS38b9 l
11.2-2, e,qiripme ni is reported th TA Bt.E A ,n d ol.is e sA ll e d I N St'b I* h II 2 2* 4 - S. EstimaiecI rele48ef *f Is'qdfel t, . The envivenme nt is discusse8 in sec.tien 11.z . 3 6: G. Devw m evali ev- cowdews ade glo wd shal s4eo.va f { ow
. . ts $ 5,40 0 b/br ( s e etsow 2. 2. 6.1 of Nunsc-oon 6,
_ _ . _ 44us sealia$ sha per.ts . cowssd es cl _e.k a a w ) .N ho\ dup ks w4. is ho k e-5 5 .4hm y, 3 0 seemd.
- 7. GE SSA R IE desq n do
and high conductivity subsyste.ss, esplat.1 how yoJ can set ectt,ely
! prevent discharge of excess water from the low conductivity suis /ste, during the time when excess water from the high conductivity st.3systea ts discharged to the envfror. ent. If you cannot prevent Jtscharge of low conductivity wastes to the enviro.nent at all tfas, tcea include the appropriate fraction of vaste dischse;d frc., this subsystem to tne enviror.nent. Respenst: At,o.t3 9 A
- 1. "D tSCH ARGE P I PIN (. IS ARRANGED So THAT Excest W ATE R. PRO M Tite L ou.7 coND41c7/s//rp Sv8Jffff4 CANNor 86 blSC14462D 7e *TNE ANU/MMfN)~ fM/S KgcpC5 WArfR YMNM FUNefleMs 53 A SUR46
- 7~n WK .
- 2. U N OS UM L OPERRTING ceder 7/dNs M/AN7eecWe l teNgtr ptseNon4e on 1Mctst wAret FAsM t//f M/dy ceNenT/v/ff SWBCffffM /3 NEEss44f (6EE SECT /8A/ //.2 3). PRielt 70 DiscHAR6E 6F THE HisH ceMDucTIutTY EXCESS w M TER TAUK, THE YnMK l$ 5 ES L MTED, SM M PbE> AMD AWMLVEKD. Snstb D Ps N TH i s A N A t t's t s, O t 3CH A *4 E 83 PERMartfD AT A SPfctrifD 2141411 Reff M Otwr/oHMeff.
(no cy,,m m3p- % 960 t3 3)
46 0.13 (CowkwM)
,I h. Describe the provisions for preventing uncontrolled releases of 4 -
radioactive materials due to spillage in buildings or from outdoor l - ~~~~ tanks if the latter is within your scope. If these provisions will i be described in your response to Question 460.09, a cross-reference to the relevant portion of Section 11.2 is acceptable. (.... --. g.. . . Respaws% S e cM ows 12.3.1 1 aw l 9.3.3 ch s c wJ.r a-J k 04-3\ ~ Coh4A a-P + h \) wAck inb A \ oct.c. t v4L.s+
$)I 3.u ALa - y ALw s c:. o p. .of q 34k Gsss a".E cAs-os a is sw cow aa.n c.m.
< A ,_
wdh Fm y\ nM.g w G m A.s.
- i. in s , c.i .
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9
..eop. =
+ . +Nen-ea
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1
. 4 6 o.13 ( Cdswwed) 1.
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Provide the concentrations of radionuclides in the excess water storage tank. Verify and correct, as appropriate the amount of radioactivity. in curies, for 1-131 and the total curies,in the concentrated waste _ _ _ _ tank given in Table 12.2-13 of your FSAR. Re3 po yss e. Ik Coh ca w mbo Ltb im & e x ass w br n& of V-m Ck t Ghwc s40m ss L agacW phk.s w eau sourm s n4e ~ +s - com sx A os
- L l e. .
of tw J.s bu h h t+6c L w tow Cok e S t \n Xfhf'SwNS sk W. (wo ekxqa tw wspawr ho 960.13j) 4 e e . ,
-wo-
- h em-.
l l . l - - - . .
..--,a
l .- ; l t EMCESS WATER TANK 5 $ 8/f I
!:ource volume = 100000 gal. l Total curics = 0.f6 "
t f soluble Fission Insoluble Fission Activation ' Italoriens Products _ Products __ Products . Isotope ,,Curles Isotope Curies Isotope _Curles Isotope _Curles nu-H3 3.8E-Of Sit-89 f .5 E-0 2 7R-95 / .? R-O f NA-24 3 .4 E-0 V nn-84 I .8 H-0 f SR-90 f .3 E-0 3 ZR-97 OR-05 O. SR-91 6 0 E-06 P-32 # .sE-0 f 1-111 3.o E-0 3 NB-95 4.O E-O f CR-51 9.O M-0 2 SR-92 2.fE-0 3 BII-103 E.6E-03 I-132 2.0 E-O L T .o E-O f MN-54 2 .fE- 0V Y-90 / .3 E-0 3 RU-106 t-133 4.1E-02. Y-91M /.VE-0 8' MN-56 #.eE-02 5.5E-0 3 Rit-103M j .0 E-0 s CO-$8 I-134 7.oE-0 V MO-99 2.5E-0 Y RH-106 E .9 E-0 Z I-135 /.6 E-0 2, .TC-99M I .VE-0 f CO-60 3.L E-0 3 f.SE-0 J LA-140 J.6E-02. FE-59 f.6 E-o f TC-101 3.g E-0 6 CE-141 / . fE- 0.4~ TOTAL l.7E-0J TE-129M WI-65 6.et-to
/ .fE-0 3 CE-143 */ . f E - 0 f Z N-(' 5 TE-132 /.9 E- 0 2 CE-144 /.2E-0f CS-134 9.f E-0 V f .?F-O f IN-69M f.)E-04 PR-143 f.YE-0Y AG-110M CS-136 3.6 E-0 f ND-147 J .8 E-0 Y 4( .f E- O S* W-187 CS-137 /.YE-07 ? .fE-0 f CS-138 f.ZE-DV TOTAL 3.7 E- 0 2. TOTAL DA-137M /.3 E-0 3 3.8 E-02. *
- BA-139 /.2 E-0 3 l BA-140 3.1 E-0 2 .
HA-141 / .fE-0 5
- BA-142 / .oE-0 6 I NP-239 2.1E-0/, I TOTAL J .Z E-0 /.
I
9 i
~
460.14 Provide additional information on the following items applicable to ' (11.3) the gaseous waste management systems:
, s. Since your system description, tables and figures in Chapter 9 of your FSAR do not clearly indicate whether *here are provtsions for both HEPA and charcoal adsorbers for the reactor building pressure control mode and purge exhaust, provide the appropriate information relating to filter units for the reactor building.
-- % pow a .
Th e. G E S SA R. E dst..si p c:1o r s,6 e7 sw c.lw A.c. O c.
.hs h r .tAnskJ ( NG 't A [i k ?Er.S o ch ay c.o o !
R O J D r- r) I" Cm Ash cLM tx w.sY-r\ M A V n <:JL OF G k r kh I fav
.\J p y oV\. b . t h H t A. m c kd o g e.,r do tw edwk o ys e. {\t C %
g es c J _o w n &c -6 h 9X t d s- Iv,mo,D
.C.4 w OA.gvs 4 A 1 wJ[ t.5 W 6 7 h
.wbd
. Dw6 p]fN pA op% o on STGS h.c( f
~
s krrr % r e w. r L#ts,bW '
..
- EX a ws k d (d%J FY'p w MA c.o w m m a i r t c u %s m .
. J The Nuclear Island HVAC design sot provide space, etc. for installing
'- exhaust filter units, other than the primary containment purge and exhaust system. All ventilation exhaust have process radiation monitors in the exhaust stream that will detect the release of radioactivity.
In event that high level of radioactivity is detected, the ventilation exhaust will automatically be shut off and the Standby Gas Treatment System will automatically actuated to ventilate that area. The above applies to the secondary containment buildings; that is, Shield Building Annulus, ECCS/RWCU Pump Rooms of the Auxiliary h .. Building and the Fuel Building and the primary containment. (No ch a :r E dyewn to 4 60.14 b)
- . ....,......~1.... . ..... . :.a .> .:... .
4 6 0 .14 ( c o wds wa.s ( *
- c. Add flow rate measuring devices for the monitors and samplers for all the airborne effluent release pathways.
. 9 = 3 P d V\ S-e ..... . .. . . . . .
( w d.t. h.. . ,, ,
'We.. fo \)od. n}...e...f6....u. e..w... .t. ....pa.t...h... q5. . a...< c.. ou...v.. .
._._ ho wa...tc. .i..n........t...h..e. W duct :
m...e.n...s..w. ..i..n...}.. ...A..c. .. . S.am.o...i..ce..s... P....... . . ... 5....$s.t...e...m......o....r..
. . . . . . . . . - . ~ . .
_(t) ._o. . ey;. {.r..e..t.-<ea.tm-=ewt (m .F..i. ....7. 9.. .. t on.). . _ @ o ff. . . t,u.t..t....v..e..a. .t..mewt-- - ...(...F.
. g. .. 7. . v .. 10 0... ....
s...e...e. . .. .. _M....o.ffgo,s. vowt fire _ (sec. _.E...i.g.. . . 7 6 . 50.k..).... . .. -..
@. contoinment. 9ed.n...o.. t. .i.o..n.. .. Q:.bg .......m._.o..v..u. .
t<..o.....i.n...q... (. .F...ig. . 14 - i.o A. ) . R 9.... el >....G.TS - t) _@.. .. .s..h....i.e. . . . . i.d. .... bs.. H V A..c. o...m......wu.. . - . . _N....3.. ...- - . . .
. Bop.....vwe..a. .su< in. ..deo.i.t..e_s...
- g. T_he.. l...o. ..n...o. w. . . . .i.ng ... . . . W.
.. .... . .. 44.. M.. o..t.....h....a dcAi.takd. d A te $.c.- frqges.s. RadAtim.'fY1oMolic9.. b. .d. k....a...e.c., .. ..
._ ..Sm. "Mimuw - fo.ked .dIOWS ..QSwe.bied ,,,gh..h. cbct ,,$houg, g3,
%eir r e 6 g.t..e..t. 2 9twhiint. ion...&w.i.9.m Odadio.yJ f.b.. .w......vv.asu..
.. .d...t..o...i._ces...-fo...r these._.d. u.. c,is... .w. c. _ u..Mhb%.... .
.. . . . . - ...s...c. o..p. .e. ...
f.w....c..a. . .,..d....s..... . .
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(51 P l ad ..h.w.. r ..... , l . ..... .w Twe. Sta.@...y. 6%. ..T..r.e. a$..m.. test.... %...... fbw:aste B.uhdq~ ... l 1 h.v...A...C.5. t.. s..t.em. v.s.o.t. ed. .-.. . .e..c. . ..D. ..W... ..G.......o....v..g..
..... . ar..e.. . 7 4 ... O ct.
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ynenswd n4...d._e..o...d....e..s... .'i..n b4 .....tk.. ..hg.. .. 5.\..a.M..h. .e..d_,,... Me.( .
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! 4C0.W ( C.o w k . w o
- d. Since the off-gas system is located in the turbine butiding which is not within the scope of your design, state whether the design of the off-gas system lies within your scope. If not, state whether
_ ' the off-gas system you have described is an interface requirement for the balance of plant. R. e S p a s s 9. 9 --
. . D e' O AJ p33 iS tu mk s c. p of 4 % Gu s S Aft ".Ir dst s\3h.
- . er State whether the source terms you have used to evaluate off-site doses due to a postulated failure of the off-gas system are consistent with Branch Technical Position ETSP 11-5 (July 1981).
. - - . . . % p ow 34 _ _ _ .__- . _ _ .
3 _ . Tb__GssSAit_E s oa v c_a svh r . a. 2 c_owsu-bd pcth 3TP E T.SP l%9 ar CA N o .S .E \ o t NA}t LA v 4 07 %
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NaS %7 cr c.k\ v~-R. Qu o C 4 n Adu .e n k d d cx n d
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- f. State whether the seismic criteria for the proposed off-gas system will conform to Section C.5 of Regulatory Guide 1.143. In responding to this question, a cross-reference to another section of your FSAR
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( ! 460.15 Provide additional information on the following items applicable to the (11.4 ) solid radwaste system:
- a. Provide the isotopic breakdown of the total curie content of " wet" solid wastes that are expected to be shipped annually to a Itcensed
(- burial site, accounting for the minimum decay available during storage prior to shipment. The total should include contributions from: (1) evaporator bottoms associated with high conductivity and detergent wastes; (2) spent resins associated with reactor water cleanup, - radwaste, regenerant condensate deep bed, fuel pool and suppression pool cleanup demineralizers; and (3) filter sludges. Provide an estimate of the number of containers which will be shipped annually. R e.s e on s q
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polishing system can generate a significantly higher volume of solidified
" wet" solid wastes (i.e. about 41,000 cubic feet for a 3400 Wt plant) than that presented in Table 11.4-2 of your FSAR. Accordingly,
) C- verify that your inputs to Table 11.4-2 of your FSAR are correct. [ Res pon se ExyesCed so/ij ndwasf.e unla es Se> .selidtfael wnste s s 2z ene st%r. Oa. resw 11.1-2) ) ni.e is a ayee,ne.1wist aperatoky aye,,w s p b 4.eep bed. glahts. .sc > age space few Ike so)(dif,W o=4h is eli assd in q uests;,, Mir e ,IS'sk g . (No ekay w r ep mv. ho 4 60 \ 5 c)
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- d. Describe your provisions for complying with Branch Technical Position l ET58 11-3, Revision 2 (July 1981). Your description should include:
l (1) the curbs and drainage provisions for containing radioactive j spills; (2) a reference to the process coritrol program as an interface requirement; (3) heat tracing for evaporator concentrate piping i
.( -
and tanks that are likely to solidify at ambient temperatures; (4) flushing connections, wherever appropriate; (5) the direct venting of equiptnent which uses compressed gases for the transport of resins or filters sludges; (6) the appropriate waste storage capacities for i tanks accumulating spent resins from the reactor water cleanup system and other sources and filters sludges in accordance with our i position in the branch technical position cited above; and (7) the volume of the available waste storage area for both the high and low-level wastes. R.a.s p on st b (i) sectik 12 3.i.: ed. Sect. iin 13.3 discus pr awif/eas 4.e c.onta.inkg yaeL\eadive api 11:. a f 4 wird p4.,d/s W ,- -ed'g M/ M (a)O.p 1' n co l
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' Provide additional information on the following items appifcable to
.y- Item !!!.0.1.1 of NUREG-0737:
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Add the containment and primary coolant sampling and containment spray - rectrrulation systems to those systems requiring periodic leak tests. {- ~ L._.
-T-f h.. State whether high pressure injection recirculation is part of the
_P leak test programs. ,_ 7" T c. Gescribe the leak reduction measures which will be incorporated into _, ; a ' _your design. .
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ATTACHMENT NO. 4 l l DRAFT RESPONSES TO CONTAINMENT SYSTEMS BRANCH QUESTIONS
480.33 ( G. z.+)
'The reactor core isolation cooling (RCIC) head spray line (penetration 48c) contains a simple check valve outside containment which does not meet the '
requirements of GDC 55. Provide a commitment to meet the requirements of GDC 55 or justify the deviation. O - D g1.p0ys on FlWart S.A.gs e RE @-MSE l i EliABC-O (4)s'TH Tae RRR. wuToovm cxuMG UME 6de.rwsu-ATHE Rcrc. Hao SPDAf UNe } Coeffpa 45 AN A(R TESTANE CHECic VP8 VE MD A S gg "ohv <-w seo ** omameo w&n..fon ee 14 FIGQunw qws MSLDE CCMTAN rneNyt . THEE. L.aNEE 15 CLA% T VP ro Th e. rwoTod OPERATED V4 uE . THc Rcrc. 60 SPORY uuE Hr % GeEA) Ree.coreo ico me FEED W A TG R. Ud - t NCOPJTAIP4FM6A27 4 : dt--s. -e stum
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4, ATTACHMENT NO. 5 DRAFT RESPONSES TO REACTOR SYSTEMS BRANCH QUESTIONS l 1 l l l l l
? QUESTION 440.20 We state in the SRP (e.g., in Section 15.1) that for anticipated
- (15) transients, the most limiting plant systems single failure shall I be identified and assuned in the analysis. Accordingly, describe I the worst single failure for each events analyzed in Chapter 15 of your FSAR. Provide analyses including these postulated failures for the five most limiting events identified in your FSAR.
RESPONSE
440.20 Question 440.20 is directly addressed, by the following two paragraphs. (15) This original question, however, has been changed verbally by the NRC Staff at a joint meeting held as an infomal review of Reactor Systems Branch Questions and Responses. This verbal question is addressed subsequent to the two following paragraphs. Chapter 15 of GESSAR contains evaluations of postulated single failures associated with anticipated transients. The system-level, qualitative-type Failure Modes and E'ffects Analysis (FMEA) of essential protective sequences in Appendix 15A (NSOA) shows compliance with the Single Active Component Failure (SACF) or the Single Operator Error (SOE) criteria. Postulating single failures in conjunction with anticipated transients of moderate frequency puts these events in the category of " infrequent events / accidents" Rev.1 (July 1981) of the SRP implies an additional requirement to perfonn radiological dose calculations for such events / accidents. These infrequent events will result in radiological consequences well within the limits of 10CFR100. The verbal question from the NRC Staff pertains to taking two transients, Turbine Trip and Generator Load Rejection, combining them with a single component failure called turbine bypass failure, then examining such events with an additional single active component failure (SACF). This was challenged by GE since transients plus multiple failures are not within the plant design basis. The GE position on classification of these events is stated in the answer to GESSAR Question 440.22. Regardless of the event classification, however, the Turbine Trip and Generator Load Rejection events are mitigated by the Reactor Protection System functions. The Reactor Protection System is a safety system designed to satisfy the single failure criterion. In order to clarify this subject, the following supplementary paragraphs will be added to Section 15.2 of GESSAR. This infomation is similar to Chapter 15 in other BWR plant FSARs. SUPPLEMENTARY INFORMATION: 15.2.2.2.3 The Effect of Single Failures and Operator Errors
9 4 0. 'Z.0 L Ce7J Mitigaticn of pressure increase from Gen ratcr Load Qu2stien, tha basic nature of this transient, is accomplished by the reactor protection system functions. Turbine control valve trip scram and @T are designed to satisfy the single failure criterion. An evaluation of the most limiting single failure (i.e., failure of i the bypass system) was considered in this event. Details of single failure analysis can be found in Appendix 15A. 15.2.3.2.3 The Effect of Single Failures and Operator Errors 15.2.3.2.3.1 Turbine Trips at Power Levels Greater Than 40 Percent Mitigation of pressure increase, the basic nature of this transient, I is accomplished by the Reactor Protection System functions. Main stop valve closure scram trip and RPT are designed to satisfy single failure criterion. l 15.2.3.2.3.2 Turbine Trips at Power Levels Less Than 40 Percent NBR Same as subsection 15.2.3.2.3.1 except RPT and stop valve closure scram trip is nomally inoperative. Since protection is still provided by high neutron flux, high pressure, etc., these will also continue to function and scram the reactor should a single failure occur.
- 440.22 In Section 15.0.4.5 and in Table 15.0-2 cf y
- ur FSAR, y:u (15.0.4) classify as " infrequent", the events identified as Load 1
Rejection without bypass and Turbine Trip without bypass. Until approval is granted to reduce their classification, it is our position that these events be classified as l " moderate" frequency events. RESPONSE: The classification of the transients such as Turbine Trip and Load Rejection combined "with failure of the bypass valves to open" is considered to be " infrequent" by the General Electric Company. The classification subject is discussed in reference LETTER MFN-331-78 ED FULLER TO DENWOOD F. ROSS, ASSISTANT DIRECTOR FOR REACTOR SAFETY, TRANSIENT RECLASSIFICATION-RESPONSE TO TECHNICAL ISSUES, previously submitted to the NRC on 8/10/78. G.E. will, however, classify both Turbine Trip with Bypass Failure and Load Rejection with Bypass Failure against criteria for moderate frequency events on an interim basis even though they are considered infrequent events. Any further discussion will be with respect to plant unique applications. 9
e e ATTACHMENT NO. 6 ., l i i Y e D e 4 DRAFT RESPONSES TO CORE PERFORMANCE BRANCH QUESTIONS
QUESTION 490.01 GESTAR-II (NEDE-24011), which contains the fuel system design safety analysis for GESSAR II, does not contain clearly identifiable design bases for most of the fuel damage, fuel failure and coolability phenomena listed in Item II.A of Section 4.2 of the Standard Review Plan (SRP). Thus, except for cladding overheating (Item II.A.2(c)) and fuel pellet overheating (Item II.A.2(d)), we have not been able to identify design basis statements in the text of GESTAR-II or in the referenced documents, even with the aid of Appendix A in Amendment 5 to NEDE-24011. While it is possible in certain cases to infer the design bases, it is preferable to have them clearly stated. Therefore, for each of the fuel system phenomena discussed in Section 4.2 of the SRP, except the two cited above, provide a concise design basis statement which indicates the design objective related to that issue. In responding to this question, provide a cross-reference to Question 490.05.
RESPONSE
The design basis for each of the phenemena listed in Item II.A of SRP 4.2 was discussed in the presentation to the Core Performance Branch of the NRC on January 25, 1983. WAZ: cal /K012020 1/20/83
~
QUESTION 490.02 Unless otherwise stated in Section 4.2 of the SRP, you should provide a design limit for each design basis. This design limit should be a numerical value of some parameter which provides assurance that the design basis (i.e., the objective or need) will be met. For all but the following phenomena, adequate design limits have been supplied or adequate explanations have been provided for the lack of design limits: (1) Fretting Wear (Item II.A.1(c)); (2) External Corrosion and Crud Buildup (Item II.A.1(d)); (3) Fuel and Burnable Poison Rod Pressures (Item II.A.1(f)); (4) Fuel Assembly Liftoff (Item II.A.1(g)). Accordingly, provide design limits for the above listed phenomena. Alternatively, discuss why no limits are required. Design limits for cladding rupture (Item II.A.2(g)), mechanical fracturing (Item II.A.2(h)), ballooning (Item II.A.3(c)) and fuel assembly structural damage (Item II.A.3(e)) are being addressed as part of separate generic reviews and need not be discussed in your FSAR now. When our generic review of these matters is completed, you should incorporate the appropriate resolutions in your FSAR.
RESPONSE
Each of the four phenomena listed above is discussed separately:
- 1) Fretting Wear - No design limit ar fretting wear is required since testing and experience indicate that fretting wear has been eliminated as an active wear mechanism.
- 2) Corrosion - No separate design limit is required for corrosion and crud buildup because it is considered in the design analyses.
Corrosion and crud buildup impact the calculated cladding temperature and material strength, and the ability of the clad to meet the stress limits prescribed in NEDE-24011-P-A-5. Thus, the amount of corrosion and crud buildup is limited by the stress limits on the clad.
- 3) Internal Rod Pressure - Limits for internal rod pressures are discussed in the response to Question 490.06.
- 4) Fuel Lift Off - The GE philosophy with respect to fuel lift is that under worst case hydrodynamic loading conditions vertical lift-off forces must not unseat the lower tie plate from the fuel support piece such that the resulting loss of lateral fuel bundle positioning could interfere with control blade insertion.
WAZ:hmm/D01207 1/21/83
QUESTION 490.03 The fuel assembly description and drawings contained in GESTAR-II are much less comprehensive than called for by Item II.B of Section 4.2 of the SRP. This particular item in the SRP contains a list of the informa- ! ! tion commensurate with an acceptable fuel system description. Accordingly, l provide the information identified in Section 4.2 of the SRP.
RESPONSE
The GE BWR fuel assembly design is described in Section 2.1 of GESTAR II. Design specification limits are provided in Table 2-1 of that report. The other information listed in Item II.B of SRP 4.2 is not provided because it is either not used in analyses or because enough detail is provided for the specific item to be derived. End plug dimensions are a function of U-235 enrichment, gadolinia concentration, and dependent on whether the end plug is in the upper or lower end of the fuel rod. The many sizes of end plugs possible make the amount of information available voluminous. Cladding inside roughness and pellet roughness data are available only as tolerances and so are not provided. Tolerances are considered in the safety evaluation, which is performed and measured against approved safety criteria. The safety evaluation is performed assuming the most limiting combination of tolerances for all critical dimensions. This is discussed in Section 2.4.1 of GESTAR II. In summary, GE believes that enough information is provided in sufficient detail to provide a reasonably accurate representation of the fuel design, thus satisfying the intent of the SRP. 4 JSC:rm:csc/112211*-1 2/1/83
QUESTION 490.04 In the recently submitted Appendix A to NEDE-24011-P-A-5, you state that a " the channel deflection analysis is provided in Section 5.3.2 of NEDE-21354-P. However, no such section exists in that topical report. Correct this reference. Furthermore, since the referenced channel box deflection report is relatively old (1976) and more data are available now regarding the magnitudes and rates of channel box deflection as a function of
. service, indicate whether: (1) the data verify the predictions of the .' deflection model in NEDE-21354-P; (2) your model adequately addresses channel bowing as well as bulging; and (3) you still recommend the periodic settling friction tests and measurements described in NEDE-21354-P, and if so, on what schedule. If you now recommend some other approach, or if NEDE-21354-P procedures have been revised, describe the changes and discuss their rationale.
RESPONSE
The reference in NEDE-24011-P-A+5 should be to Section 3.2 of NEDE-21354-P. Correction of this will be made in the next amendment to NEDE-24011-P-A-5. The other parts of Question 490.04 are addressed below: (1) Additional channel bulge data to higher exposures have been obtained since the publication of NEDE-21354-P. These data confirm the adequacy of the model used by General. Electric to predict channel bulge. (2) The model in NEDE-21354-P addresses only channel bulge. GE has recommended to utilities an approach to mitigate the effects of channel bowing. (3) The following general guidelines minimize the potential for and detect the onset of channel bowing: 1. Channels shall not reside in the outer row of the core for more than two operating cycles.
- 2. Channels that reside in the periphery (outer row) for more than one cycle shall be situated in a core location each successive peripheral cycle which rotates the channel so that a different side faces the core edge.
- 3. At the beginning of each fuel cycle, the combined outer row residence time for any two channels in any control rod cell ,
l shall not exceed four peripheral cycles. After core alterations (i.e., reload) and before reaching 40% thermal power, a control rod drive friction test
- is recommended for those cells exceeding the above general guidelines. After the technical specification scram speed surveillance test on each rod, as required by BWR/6 Standard Technical Specification 4.1.3.2.a, each control rod meeting the above JSC: csc: rm/112211*-2 1/21/83 l l l
t 1 conditions will be allowed to settle a total of two notches, one notch at a time, from the fully inserted position. ; Total control rod drive friction is acceptable if the rod settles, under its own weight, to the ne).t notch within approximately ten seconds. If the rod settles too slowly, a rod block alarm will actuate, indicating possible impending channel box-control blade interference. The results l of this test will be considered acceptable if no rod block alarm is received. This testing will give an early indication of this interference and will prompt an investigation into the source of the friction. If necessary, corrective action will be completed before startup after the t next core alteration. I In lieu of friction testing, fuel channel deflection measurements may be ' used to identify the amount of remaining channel lifetime.
*This control rod settling ft :-tion test provides an equivalent level of the tests described in NEDO-213; This test provides adequate assurance of the scram function. The amount of friction detectable by this test is $250 lbs. Control Rod Drive Tests indicate that the CRD will tolerate a relatively large increase in driveline. friction (350 lb) while still remaining within technical specification limits. The control blade is in its most constrained, highest friction location when it is fully inserted.
The ability of the blade to settle from this position demonstrates that the total drive line friction is less than the weight of the blade (*250 lbs). JSC: csc: rm/I12211*-3 1/21/83
~
l . QUESTION 490.05 l In GESTAR-II (NEDE-24011), which is the primary support document for the fuel system for your proposed 238 nuclear island design, you have not provided a discussion of fuel assembly liftoff for normal operation and
" abnormal transients" which are separate and distinct from our concerns l
regarding the seismic-and-LOCA-loads liftoff. As indicated in Item II.A.1(g) of Section 4.2 of the SRP, however, worst-case hydraulic loads for normal operation should not exceed the holddown capability of the fuel assembly. Although your letter from Gridley to Eisenhut, dated July 11, 1977, addresses this issue for plants and fuel designs of 1977 vintage, it is not evident that assembly liftoff will be precluded for normal operation, including anticipated operational occurrences or abnormal transients in your proposed 238 nuclear island. Accordingly, provide a discussion of your analysis of this issue. The design basis i rnd limits aspects of this issue should be addressed as part of your l response to Questions 490.01 and 490.02. ) 1 RESPONSE < l Design changes which have occurred in the GE BWR fuel bundle since the July 11, 1977 letter from Gridley to Eisenhut have not changed the conclusions reached in that letter. It is therefore considered appli-cable to the fuel to be used in the 238 Nuclear Island design. GE perfoms two analyses for the affect of hydrodynamic loads on our fuel bundles. The results of the analysis are provided in the plant specific tiew Loads Report. The first analysis is for loads on the bundle resulting from Upset Conditions (i.e. loads from the combination of normal operation plus OBE plus scram plus SRV actuation). For these operating conditions, including Anticipated Operational Occurrences, GE does not calculate any fuel bundle separation from the core support piece. The second analysis is for loads on the bundle resulting from Faulted Conditions (i.e. loads from the combination of normal operation, LOCA, SSE, SRV Actuation and
. Mark III containment hydrodynamic conditions). The method of analysis used for these Faulted Condition loads is provided in NEDE-21175-3-P "BWR Fuel Assembly Evaluation of Combined SSE and LOCA Loadings." These faulted loads bound the loads calculated for Upset Conditions.
NEDE-21175-3-P provides a discussion of the margin between the calculated fuel movement as a result of the Faulted Condition loadings and that movement required to disengage the bundle lower tie plate from the fuel support piece. JSC: csc: rm/I12211*-4 1/17/83
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+ QUESTION 490.06 j You state internal inpressure. gas Section A.4.2.1.1.6 of GESTAR-II that there is no limit for The internal pressure is used in conjunction with other loads on the fuel rod cladding in calculating cladding stresses. The results of such calculations which are provided in Section 2.5.1 of NEDE-24011, show that the calculated cladding stresses can be accommodated. ,
- j Although this analysis may satisfy our acceptance criteria for cladding
{l stress (Item II. A.I.a of Section 4.2 of the SRP), it does not satisfy our acceptance criterion for rod internal pressure (refer to Item II. A.1.f of l j SRP Section 4.2 and Question 490.01) because this criterion involves more 4 than stress limits on the cladding. The rod internal pressures used in your cladding stress calculations are well in excess of the nominal coolant system pressure. Accordingly, justify operation under these
- , conditions and explain why the absence of an internal gas pressure limit
,: does not appreciably decrease the margin of safety in calculating fuel system damage. 1-i,
RESPONSE
i Fuel and poison rod internal pressure increases with increasing burnup ' and at end-of-life the total internal pressure, due to the combined effects of the initial helium fill gas and the released fission gas, is at a maximum. This maximum internal pressure is used in conjunction with other loads on the fuel rod cladding to calculate cladding stresses. While there are no limits on internal gas pressure stated in GESTAR II or elsewhere, the maximum internal gas pressure actually is limited by the stress limits in GESTAR II consistent with the following criterion: The stress in the c.ladding resulting from differential pressure will not exceed the stress limits specified in Section 2.5 of GESTAR II. GE has performed evaluations for P8x8R and barrier P8x8P using the GESTR(M) mechanical model, in which the fuel rod cladding creepout rate was calculated and compared with the fuel pellet irradiation swelling rate, during late life operation when the fuel rod internal pressure is highest. The results of this evaluation demonstrate that the cladding creepout rate is less than the fuel pellet irradiation swelling rate. This indicates that the fuel cladding gap is not expected to increase under the maximum planned normal operating conditions. GE has accumulated significant operating experience with all fuel types. This experience includes the operation of Lead Test Assemblies to higher than design exposures. Based on numerous post-irradiation inspections performed, it is concluded that the fuel is adequately designed relative to fuel rod internal pressure since cladding creepout due to fuel rod internal pressure has not been observed. JSC: rm: csc/112211*-5 1/21/83
ATTACHMENT NO. 7 s DRAFT RESPONSES T0 11ECHANICAL ENGINEERING BRANCH QUESTIONS
QUESTION 48 Which operational transients will be used for preoperational testing of de non-dSSS piping systems? Which system will be monitored and what
.._ __ locations will be instrumented?
RESPONSE
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3.9.2,1.2 PREOPERATIONAL TESTING OF NON-NSSS PIPING l 3.9.2.1.2.1 PREOPERATIONAL VIBRATION TESTING This subsection defines the general requirements for vibration testing of piping systems as specified in Regulatory Guide 1.68 "Preoperational and Initial Startup Test Programs for Water-Cooled Power Reactors". Specific vibration testing requirements are defined in ANSI /ASME OM3-1982,
" Requirements for Preoperational and Initial Startup Vibration Testing of Nuclear Power Plant Piping Systems". An outline of that standard is given here. Preparation of detailed test specifications by the Applicant will require consulting the complete standard.
Ingrumentation locations will be provided by the Applicant in accordance with vibration monitoring group selection (Subsection 3.9.2.1.2.1). Piping systems to be tested are classified into three vibration monitoring groups, according to required degree of test sophisications. Vibration Monitoring Group 1 (VMG1) requires precise test instrumentation plus some degree of mathematical analysis, where as VMG3 may require only visual observations by competent personnel. VMG2 permits simpler instrumentation than VMG1, such as hand-held on temporarily mounted displacement meters or accelerometers. 3.9.2.1.2.1.1 VIBRATION MONITORING GROUP SELECTION In selecting monitoring groups for various systems or piping configuration, a general rule is that the most rigorous testing (VMG1) thould be applied to systems where vibration has the greatest safety implication. Ad an example, VMG1 is applied to all Safety Class 1 Piping. VMG1 may also be applicable to Safety Class 2 and 3, if vibration-producing elements 1 (pumps, compressors, relief valves, etc) are present. Table M 4 lists systems being preoperationally tested for vibration. Th. ., P classification for vibration testing shall be made by applicant. Prelimina y testing may require changes in group selection if unforeseen problems develop.
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alt For ASME Class 1 piping: S FS #351. alteMC 22s' 1 Where C 2
= Secondary stress index as defined in ASME Code.
K2 = L cal stress index as defined in ASME Code. M = Maximum zero-to-peak dynamic moment due to vibration only, or in combination with other loads as required by system design specification. F = Factor of safety applicable to class 1 piping (=1.3).
= Stress factor (=0.8 for carbon steel, = Ccde for s+ambss 34ee.I )
Sg= Endurance limit (S,) from Figure I- or I-9.2 of Section III of the ASME Code 2 = Section modules of pipe For ASME Class 2 and 3, and ANSI B31: S dt =C 22E kk 5 *C b E Where 22 C K =21, i = Stress intersification factor, as defined in : Subsections t NC and ND of the ASME Code or B31. For transient vibrations, the and maximum alternating stress should be limited as follows. For ASME Class 1 piping, S,= allowable alternating peak stress value from Figure I-9.1 or I-9.2 using N ywhere N y= (EVLC) EVLC = Equivalent number of maximum anticlpated vibratory load cycles. U y = Unused usage factor = 1-U U = Cumulative usage factor from ASME Class I analysis, which excluded vibratory code. The maximum alternating stress intensity S dt shall be limited to 0.8 SEL for carbon steel, or 0.6 S EL f r stainless steel. For ASME Class 2, 3, and B31 piping, the stresses shall be evaluated the same as for steady state vibration. C. 3.9.2.1.2.1.3 {EPTANCE CRITERI A -- VMG-2 For testing utilizing deflection measurements, acceptance is based on the following equation. S allow = M EL gn toooo C Where S gjo = allowable zer o Kto, peak deflection limit
l S, = Value of deflection cbtain:d all other symbols are the same as given for VMGl. For testing utilizing volocity measurement, acceptance is based on 'the following equations. 3.64 x 10 3 Y allow *- C3C4 3 EL C C E 3 - 2 2 I Where V,33,, = Allowable velocity, in/second C 1
= Correction factor to compensate for the effect of concentrated weights along the characterestic span (sec. Fig.10 of OM3-1982)
, C 3 = a correction factor accounting for pipe contents and insulation
=
1.0 + W + W F INS
- W W Where W= weight of the pipe per unit length (Ib/ft)
W p= weight cf the pipe contents per unit length (Ib/ft) W INS = the weight of the insulation per unit length (1b/ft)
=
1.0 for pipe without insulation and either empty or containing steam C = 4 correction factor for end conditions different from fixed ends and for configurations' different from straight spans
=
1.0 for a straight span fixed at both ends, but conservative for any practical end conditions for straight spans of pipe
=
1.33 for cantilever and simply supported pipe span
=
0.74 for equal leg Z-bend
=
0.83 for equal leg U-bend Appendix D of OM-3 presents examples of correction factorsyC and C for 4 typical piping spans along with a combination of these factors to provide an initial screening method, c. 3.9.2.1.2.1.4 $EPTANCE CRITERI A -- VMG3 The acceptability of piping is determined by visual observation, or employing simple devices such as rules, optical wedge, or spring scale. If the level of vibration is too small to be perceived and the possibility of damage is judged to be minimal, the system is acceptable. The judgement as to acceptability can be made only by evaluation of all the following facts as to their effects on piping stress. a) Vibration magnitude and location l
b) Proximity to sensitive equipment c) Branch connection behavior d) r spability of nearby component supports Any unique operational characteristics of the systems shall be considered in the evaluations. If unacceptable vibration levels are indicated by the nethod listed above, the system must be reclassified as either VMG2 or VHGl. 3.9.2.1.2.1.5 CORRECTIVE ACTION Should the piping vibration exceed the acceptance criteria, correction action must be taken to make the system acceptable. This action may consist of adding supports, reducing forcing functions, determining and modifying resonant sections, or changing operating conditions. After corrective action is taken, additional testing shall be performed to determine if the vibrations have been sufficiently reduced to satisfy the acceptance criteria. 3.9.2.1.2.2 PRE 0PERATIONAL THERMAL EXPANSION AND DYNAMIC TESTING Preoperational thermal expansion and dynamic testing is provided in Subsection 14.2.12.1.75.
Tablo 3.9-24 , PIPING SYSTEMS TO BE VIBRATION TESTED / INSPECTED (PREOPERATIONAL) , Reactor Feedwater System Reactor Water Cleanup System Standby Liquid Control System Residual Heat Removal System Reactor Core Isolation Cooling System Reactor Recirculation System Controlled Drive Hydraulic Systems low Pressure Core Spray System i High Pressure Core Spray System Fuel Pool Cooling and Cleanup System Leak Detection System Li.ould and 1;alid Radwaste Sygtems, , Neutron Mon'itoring' System Offgas System Upper Pool Storage System N I Chilled Water System Demineralized Water and Condensate Distribution Essential Service Water System Heated Water Distribution HPCS Service Water System Suppression Pool Make-up System Suppression Pool Cleanup Essential Bldg. Chilled Water RHR Service Water Systems Condensate System Res eko< Proka.chas ykem
QUESTION 50 Due to a long history of problems dealing with inoperable and incorrectly l installed snubbers, and due to the potential safety significance of failed snubbers in safety related systems and coeponents, it is requested that maintenace records for snubbers be documented as follows: Pre-Service Examination A pre service examination should be nde on all snubbers listed in Tables 3.7-4a and 3.7-4b of Standard Technical Specifications 3/4.7.9. This examination should be made after snubber installation but not more than six months prior to initial system pre-operational testing, and should as a minimum verify the following:
- 1. There are no visible signs of damage or impaired operability as result of storage, handling, or installation.
2.. The snubber location, orientation, position setting, and configuration (attachments, extensions, etc.) are according to des'ign drawings and specifications.
- 3. Snubbers are not seized, frozen or jammed.
- 4. Adequate swing clearance.is provided to allow snubber movements.
- 3. If applicable, fluid is to be recommended level and is not leaking from the snubber system.
- 6. Structural connections such as pins, fasteners and other connecting hardware such as lock nuts, tabs, wire, cotter pins are installed correctly.
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If the period between the initial pre-service examination and initial
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system pre-operational tests exceeds six months,due to unexpected situations, l reexamination of items 1, 4, and 5 shall be performed. Snubbers which i are installed incorrectly or otherwise fail to meet the above requirements must be repaired or replaced and re-examined in accordance with the above criteria. Pre-Operational Testing During pre-operational testing, snubber thermal movements for systems whose operating temperature exceeds 250*F should be verified as follows:
- a. During initial system heatup and cooldown, at specified temperature intervals for any system which attains operating temperature, verify the snubber expected thermal movement.
- b. For those systees which do not attain operating temperature, verify via observation and/or calculation that the snubber will accommodate the projected thermal movement.
- c. Verify the snubber swing clearance at specified heatup and ccoldown intervals. Any discrepancies or inconsistences shall t:e evaluated for cause and corrected prior to proceeding to the next specified sn4crva) .
The above described operability program for snubbers should be included and documented in the pre-service inspection and pre-operational test programs. The pre-service inspection must be a prerequisite for the pre-operational testing of snubber thermal motion. This test program should be specified in Chapter 14'of the FSAR. 1 ( FSH:pab/J12117 12/11/82 .
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C duCs r T6ET Frot. qqog go 3.9.3.4.1 Piping (C h inue e tests are conducted at various tempera-tures to ensure operability over the specified range; e peak test loads in both tension and compres-sion are required to be equal to or higher than the rated load requirements: and e the snubbers are tested for various abnormal environment conditions. Upon completion of the abnormal environmental transient test, the snubber shall be tested dynamically at a frequency with a specified frequency range. The snubber must operate normally during the dynamic test.
)
(d) Snubber Installatien Requirements An installation instruction manual is required by the pipe support design specification. This manual is required to contain instructions for storage, handling, erection, and adjustments (if necessary) of snubbers. Each snubber has an installation location drawing which contains the installation location of the snubber on the pipe and structure, the hot and cold settings, and additir nal informa-I Q $15%Cl~ tion needed to install the particular snubber. s
-s' (45 Struts - The design load on struts includes those lcac. s caused by dead weigh , thermal expansion, seismic fcrces (i.e., operating basis earthquake and safe shutdown earthquake), hydrodynamic loads, system anchor displace-ments, and reaction forces caused by relief valve dis-charge, turbine'stop valve closure, etc. )
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examinationehowMhmadeaftegubberinstallationbut six months prior to initial system pdNp'e'r* anon $ YestN ei ;; a.ainieemtverify the following:
~
. 1. There are no visible signs of damage or impaired operability as
] result of storage, handling, er installation.
2.. The snubber location, orientation, position setting, and configuration (attachments, extensions, etc.) are according to des'ign drawings and specifications.
..(
- 3. Snubbers are not seized, frozen or jammed.
- 4. Adequate swing clearance is provided to allow snubber movements.
- 5. If applicable, fluid is to be recommended level and is not leaking from the snubber system.
- 6. Structural connections such as pins, fasteners and other connecting hardware such as lock nuts, tabs, wire, cotter pins are installed '
correctly. If the period between the initial pre-service examination and initial system pre-operational tests exceeds six months due to unexpected situations, reexamination of items 1, 4, and 5 she+1pb perfonned. Snubbers which are installed incorrectly or othenvise fail to meet the above requirements l be' repaired or replaced and re-examined in accordance with the above i criteria. . \ ( .- _ - _ - - - - - - . - - - - - - . - - . - - - _ . _ _ - - - - - - - -
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y - The purpose of this test is to verify that the(non-NSSS) safety-related . piping, designated as AShi Class 1, 2, or 3, is free to expand thermally as designed.--d tSt : : -ient-hdeceeyty. -ui.ima stnhrteady state- vibratitrer
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&g Frerequisites
'-~- The system piping to be tested is supported and restrained in conformance with the design drawings.
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'- Instrumentation has been installed and calibrated.
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_ ,-N - During initial system ~~ bestup, piping thermal movements at selected points will be instrumenteJ, ~~~~ monitored, and recorded. Accessible pipe hangers and snubbers not
-{~- instrumented will be visually inspected. ~
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$4 Acceptance Criteria
- 1. There shall be no evidence of blocking of thermal expansion of the piping systems or components other than by design. --
- 2. The measurec thermal movement shall be within 125 percent of the ,
analytical value or 10.25 inch, whichever is greater.
- 3. Spring hanger movement shall r.emain within the hot and cold set pointse k r" ,o . . ._--._.. ..... .. .... . ...., .~. .._ .. .. u. ____
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. _ _ , [_ gnubber thermal movements for systems f a o those operating temperature exceeds 250*i should be verified as follows:
Nin l fJ_ ! a. During initial system heatup and cooldown, at specified temperature 5* l _ intervals for any system which attains operating temperature, verify j the snubber expected thermal movement. __( I d
- b. For those systems which do not attain operating temperature, verify og~
via observation and/or calcu'lation that the snubber will accommodate
- 2. C the projected thermal movement.
i i c. Verify the snubber swing clearance at specified heatup and cooldown l intervals. Any discrepancies or inconsistences shall be evaluated
- f. .. for cause and corrected prior to proceeding to the next specified l
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WRC Regulatory Guide 1.70, Item 3.9.2.1 ,
"Preoperational Vibration and Dynamic Effects Testing on Piping" says the preoperational piping vibration and dynamic effects testing during.
. _{ _. startup functional testing will be conducted on safety-related
~ ASME class 1, 2, and 3 piping systems including their supports and restraints. The following test program is intended to comply with
- that requirement.
14.1.11,l .~15.1.1 M ' ;PR55is,dUj[ SIT C Vibration tests shall not be made before all piping and supports have been inspected and determined to be properly installed and hydrotested. __ { _ 14. Ell.i.~iSAJ. W%nATton Tuv 'h ea oisi+ L g - Dr?NIPT!O:4 OP. .THE CST--PiCGRXN- The vibration test is designed to be run with the reactor and associated system in either the hot or cold condition. The test program is divided into two phases, t e*M tof'A*j- N.' v- S'a,4.** 3 N a -- to & ~ p
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g.'t.rt.t.15. M b e . w h b m e.o Itcw.*CN . Phase I - The dynamic response of the system is noted by observation ~ and visual instrument measurement. Piping with less than allowable deflections requires no further evaluation and can be approved to ~ have met the requirements of section 3.9.2 of ?.egulatory Guide 1.70. Allowable deflections should be developed af ter completion of stress . analysis. Piping exceeding Phase I acceptance limits will be ~ treated as described in Phase II. , Phase II - Take remedial action (add or relocate supports, etc) or proceed with time history analysis. 1 ,. .s
~
Apply time history analysis to determine whether additional corrections are required. . i
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- PHASE I All safety-related piping shall __
i
,be subjected to preliminary vibration measurements. These measurements - l l shall be taken during pre-operational tests, with machinery and fluid , l systems operating under test conditions. Any indication of persistent -
l vibration shall be followed by recorded measurements for subsequent analysis. l-Special attention shall be given to piping M. TEST 7 CONDITIONS ,, attached to pumps, compressors, and other rotating or reciprocating equipment. Measurements shall be taken near isolation valves, pressure- l control valves, and other locations where shock or high turbulence cay be present.
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tr.pr. m = +. . +cewe e ta ws 4-w:.le 4.n-T __: ?.. L 1 ;i'?ew li.st-of-recomended A " 5. TEST INSTRUMENTATION. tas' ' - - * - " = . Preliminary measurements may be made with a - light-weight portable vibration meter. e9 C BsMcd:lt:1. Frem these measurements, the number and location of recorded measurement points shall be determined. , JAF"3" &* RECORDED MEASUREMENTS Every measurement record shall be - accompanied by a sketch showing the location of the measurement point,
- plus a description of the system operating conditions at the time of measurement.. Measured data shall include actual deflections and ,,
frequencies. Time duration of measurement shall be sufficient to indicate whether the vibration is continuous or transient. -
-e - JCCF .CPHASE II ACTION If the allowables are exceeded, two options are- ~
- available, whichever is deemed appropriate. ,
a Take remedial action (add or relocate supports, etc). - b Perform time-history test of the piping system. wwsp s (c.:Ge _.
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t I .1.I1. (.1%L V. m n ,.a 'T~ c u- fr L e1 0.>0 c (c N m .e ...u) p G, TIME-HISTORY TEST , hEstablishthetime-historyofthepipingsystem. A(r Perform stress analysis based on time-history and compare with code allowables. FC If the allowables are exceeded, take remedial action.
$ $.1. REMEDIAL ACTIONS Two basic methods of solving the problems are suggested, one or both of which may be used in a given case.
a
$A Change in the piping arrangement. This includes a number of possible changes, as - 's a Adding and/or relocation of piping supports.
b Rerouting of piping layout to eliminate fluid resonance characteristics. (S' Change in the flow moder of the system by - a Increasing opening or closing time of valves. b Addition of a device og a grid, strainer or damper, which _ _ , minimizes the forcing function of the vibration source.
~
These solutions require partial or full reanalysis of the affected _ piping system. i m I
/4 2-sos- 3
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QUESTION 59(3.9.3).
- Using the guidance of IRlREG-0609, provide the methodlogy.used and the results of the annulus pressurization (AP) analysis (asymmetric LOCA Ioads) for the reactor system and affected components. including the fellowing: .
- 1. reactor pressure vessel >and supports,
- 2. core supports and other reactor internals,
- 3. control rod drives, .
4
- 4. ECCS piping attached to the reactor coolant system,
- 5. .prIsary coolant piping, and
- 6. piping supports for affacted piping e The results of the above analysis should specifically address the effects of the combined loadings due to annulus pressurization and an SSE.
RESPONSE
The following is a brief description of the methodology.
- 1. Pressure-Time Histories 1
The pressure time histories in the annulus region between the RPV I and shield wall are generated from a feedwater line break and a f recirculation line break. The RELAP code using nodalized mass and energy balance is used in this analysis.
- 2. Concentrated Force-Time Histories e
FSH:pab/J12117 12/11/82 l g l . 1 _ _ _ - _ - _ _ - _- . _ - . - --- - - - - .--- - - - -
The farcing function of j:t impingement on the shiold wall is
)
. obtained from the break flow transient cause by a feedwater line j '
. break and a recirculation line break. Forcing functions of jet f reaction on RPV, jet impingement on RPV, and pipe whip restraint load on restraint anchors are obtained from the feedwater line 3 [ . . , ,
creak, the recirculation line break, and main steam line break.
- 3. Intagrated Dynamic Analysis Bean and shell models are used to integrate pressure-time histories I and concentrated force-time histories in determining the effects on
[ the sheild wall pedestal, vessel support, core support and internals, and control rod drives. These @namic analyses yield displacements, forces, stresses and moments.
- 4. Attached Piping Analysis )
I Acceleration time history from the integrated dynamic analysis is used to generate response spectra for the stress analysis of the attached piping. This analysis covers ECCS lines, primary coolant piping, and associated pipe supports.
- 5. Load Combination for Vessel and Piping i
Asymmetric LOCA loads in combination with SSE by the SRSS methodology are treated as a faulted condition for evaluation against the ASME l Code and functional capability requirements. This is described in ! Table 3.9-2. Tk As .I o e. a n n wk u .c y ywg 4 on awl i.f w s ( [ b-4. fVoV\ Mp \ caw
- h hM FSH:pab/J12117 12/11/82
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