ML20147G499
| ML20147G499 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 12/20/1978 |
| From: | Novarro J LONG ISLAND LIGHTING CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| SNRC-351, NUDOCS 7812260221 | |
| Download: ML20147G499 (19) | |
Text
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LONG ISLAND LIGHTING COM PANY FLt'O,f j SHOREHAM NUCLEAR POWER STATION sfe P.O. DOX 618, NORTH COUNTRY RO AD e WADING RIVER, N.Y.11792 December 20, 1978 SNRC-351 Mr. Ilarold R.
Denton, Director Office of Nuclear Reactor Regulation U.
S.
Nuclear Regulatory Commission Washington, D.
C.
20555 SIIOREHAM NUCLEAR POWER STATION - UNIT 1 DOCKET NO. 50-322
Dear Mr. Denton:
We enclose, herewith, fifteen (15) copies of the following:
1.
NRC requests for additional information contained in Mr. Steven A.
Varga's letter to Mr.
A.
W.
- Wofford, dated August 31, 1978.
2.
Structural Assessment of the Sacrificial Shield Wall Under the Effects of Annulus Pressurization.
3.
Revised response number 23, previously submitted on March 31, 1978 (SNRC-272).
The above listed information corresponds to Outstanding Issues No. 6 (part 3), No. 6 (part 2) and Confirmatory Issues No. 18, respectively.
,Very~truly yours,
)). ) Jc, a vt c 3
,J.
P. Nova ro,
' Project Manager Shoreham Nuclear Power Station BRM/cl Encls.
2 s i n c o. u l-s(
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STRUCTURAL ASSESSMENT OF THE SACRIFICIAL SHIELD WALL UNDER THE EFFECTS OF ANNULUS PRESSURIZATION The design of the sacrificial shield wall was based on a uniform pressure distribution and was completed prior to consideration of asymmetric annulus pressurization.
The design has been assessed for the transient pressure profiles presented in Figures 6.2.1-38 of the FSAR as discussed below.
The structural assessment for asymmetric annulus pressurization consists of two parts.
The first part is the dynamic structural analysis of the reactor building complex.
This analysis is I
performed to determine the overall building response in terms of internal forces and moments and also amplified response spectra (ARS).
The second aspect of the assessment is the evaluation i
of the local effects of the asymmetric pressure load applied directly to the sacrificial shield wall.
1 The dynamic analysis for overall building response is performed using the ' lumped mass' structural model developed for seismic analysis and described in Section 3.7.2A of the Shoreham FSAR (Fig. 3.7.2A-1).
The asymmetric pressure distribution is integrated to obtain the resulting lateral loads on the RPV and shield wall at each of the time steps used to define the transient annulus pressurization loading.
The horizontal force time histories are then applied to the dynamic structural model and the solution is obtained by modal analysis.
The results include the maximum overall shear across the shield wall and the maximum overall bending moment.
These result j
respectively in shear and meridional membrane stresses of inconsequential magnitude required to resist those maximum internal loads.
The local stress analysis of the shield wall inner shell in the region of the maximum asymmetric pressure (node 16 of Fig. 6.2.1-36 for the feedwater break) is performed using a finite element structural model as described in Section 3.8.3 of the FSAR (Fig. 3.8.3-6) and confirmed analytically.
The maximum local stress due to the local maximum pressure differential across the shield wall of 86 psi (Fig. 6.2.1-38A) is combined with the stress level at the location due to all other loads in the critical load combination.
In this critical combination, the total stress does not exceed acceptable limits as specified in FSAR Sections 3.8.3 and 3.8.4.3.4.
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1 A ten percent increase in the local pressure differential in the break node (node 16 of Fig. 6.2.1-36 for the feedwater break) to account for the higher results obtained by the staft, can be accomodated by the present shield wall design.
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REQUEST:
23.
RESPONSE TIME TESTING, SECTION 14
)
The applicant has not submitted a description of the methods that will be used to measure the response times of the reactor protection system and primary containment isolation system including sensors and process to sensor coupling (i.
e.,
instrument lines).
The staff requires these measurements during preoperational testing for all instrument channels for which response time verification is required during plant life by the plant technical specifications.
(The technical specifications do not require response time measurements of all these channels prior to startup).
The applicant has not completed developing this portion of the test program and has committed to submit a des-cription of these tests at least six months prior to fuel loading.
The staff will review this information prior to issuance of an operating license and report our findings in a supplement to the Safety Evaluation.
RESPONSE
The applicant reaffirms its committment to submit descriptions of these tests at least six months prior to fuel loading (i.
e.,
Preoperational tests).
As required by the Shoreham Technical Specifications,
. response-times for the reactor protection system and the primary containment isolation system will be obtained.
Although these two systems perform separate functions, the primary sensors used to measure the process variable (i.
e.,
pressure, level, etc.) are similar and may be tested using the same test methods and equipment.
Methods and equipment used to test the primary sensors are given by primary. sensor type as follows:
1.
Pressure Sensor.- the pressure sensor response test directly Eleasures the sensor response time
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by comparing the sensor being tested to a high speed sensor when both are simultaneously subjected to a ramp input.
The ramp is generated by pressurizing an air cylinder, then bleeding this pressurized air through a needle valve to slowly pressurize a second cylin' der which is l
partially filled with water.
As this second l
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cylinder is pressurized it transmits the pressure, j
in the form of a ramp, to the tested sensor and the high speed comparison sensor.
By recording the output of each sensor on a brush recorder, the times at which each sensor crossed the trip point can be measured and the response time determined.
The hydraulic signal generator used to yield the ramp input is commercially available from Nuclear Services Corporation and all other necessary equipment such as brush recorders and pressure indicators is available at Shoreham.
2.
Level Sensors - The level sensors used in the reactor protection system and the primary containment isolation system utilize differential pressu'res to give their indications and trip functions and thus are testable using the method described under pressure sensors.
3.
Flow sensors - similarly, the flow sensors that utilize differential pressures are testable using the method described under pressure sensors.
4.
Temperature Sensors -
a.
RTD Sensors - These sensors are tested using the self-heating test which is a steady-state
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version of the loop current step response method.
The self-heating test utilizes the dependence between temperature rise in an RTD during steady-state internal ohmic heating and the sensor-to-fluid heat trans fer resistance.
The normal sensing current in the RTD is ex-ternally increased from 1 to 5 milliamps to a significantly higher value of 20 to 100 milliamps.
The change in the RTDs resistance is measured such that a characteristic resistanco versus current curve can be constructed whose slope is related to the sensor response time.
The equipment needed to perform this test, consisting of a power supply, a bridge and a brush recorder, is all available at Shoreham.
b.
_Thermocpuple Sensors - the testing method for thermocouple sensors is currently being developed.
Q 5.
Radiation Monitors - by Technical Specification definition, detectors are exempt from response time testing.
The response times will be measured from the first electronic component in the channel using a current source to simulate detector outputs.
All required test equipment is presently available at Shoreham.
6.
Valve Stem Position Switches - These devices change state coincidentally with their first set of contacts in the logic string, and a brush recorder is all that is required to determine the response time.
For all the types of primary sensors given above, only the sensor response time itself was covered.
Per Technical Specifications, the response time out to an end action will also be determined either using more channels on a brush recorder, as in the case of the reactor protection system; or by using several procedures, each of which allows determination of a portion of the total channel response time, as in the case of the primary containment isolation system.
' Process to sensor coupling (i.e.,
instrument lines) will also be included for,$_p and pressure instruments.
This time shall be calculated on a worst case basis (longest instrument line) and shall be added to the other parameters previously described to obtain the total instrument channel response time.
4.
Full scrams will F-erformed to verify proper operation of the entire syst.
These scrams may be coordinated with the CRD preoperational test, and will include reactor protection system response time measurements for those sensors listed in Table 3.3.1-2 of the Technical Specifications.
The total channel response times will take into account the process to sensor coupling (i.e., instrument lines).
5.
Proper functioning of all appropriate annunciators will be demonstrated.
6.
All auxiliary actions associated with a trip of the RPS will be veiified out to the end action:
for example, closure of the appropriate isolation valves.
7.
It will be verified that no scram occurs when only one trip system is lost.
Acceptance Criteria 1.
The applicable general acceptance criteria, as listed in Section 14.1.3.6, will be met.
2.
MG set time from stop initiation to 5 percent under-voltage or under-frequency shall be greater than or equal to the minimum design required value.
3.
.The MG set under-frequency trip point shall meet the design required value.
4.
Reactor protection system response times (i.e.,
the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until doenergization of the scram pilot valve solenoids) shall be less than or equal to the values listed in Table 3.3.1-2 of the Technical Specifications.
5.
The scram reset time delay shall meet the minimum design required value.
14.1-44
1 SHOREHAM - RESPONSE TO NRC REQUESTS OF AUGUST ll,1978 REQUESTS
)
221.30 The GE analytical methodology used to determine the short term mass and energy release rates for annulus pressurization analyses is presently under review.
In the interim, it is requested that confirmatory calculations for the feedwater line and recirculation line breaks be performed with the RELAP 4/ MOD 5 computer progran.
You should perform the RELAP 4/M00 5 calculations using the following assumptions:
Use the Henry /Fauske break flow model during subcooled a.
blowdown and the Foody break flow model during saturated blowdown.
b.
Do not model momentum flux between the reactor vessel and its connected nozzles (oiping).
Do. lot model the loss coefficients in the feedwater c.
sparrer, do not assume sa;urated R.V. fluid condition's, justify the amount of subcooling.
d.
For the feedwater line break, model the entire feedwater system, includino the feedwater pumps.
221.31 For the feedwater and recirculation line breaks, provide a table of break area and nass flux calculated for each side of the break (if applicable) as a function of time.
Also, oraphi-cally overlay the RELAP a results with the results previously calculated by the GE method.
221.32 Figure 6.2.1-5 indicates that for the recirculation outlet break the pipe is severed between the vessel and the shield wall. Verify this interpretation.
However, if the severed plane is within the shield wall, provide the flow area between the shield wall and the recirculation pipe.
Provide a detailed graphic description of the break location.
221.33 Provide a listing of the PELAP 4 input data used in the feedwater and recirculation line breaks.
RESPONSE
Annulus Pressurization refers to the loading on the shield wall and reactor vessel caused by a postulated pipe rupture at the reactor pressure vessel nozzle safe-end to pipe weld.
The pipe break assumed is l
an instantaneous ouillotine rupture which allows mass / energy release into the dry well and annular reoion between the biological shield wall i
and the reactor pressure vessel (RPV).
h
RESPONSE (cont.)
The mass and eneray released during this postulated pipe rupture causes-A rapid asymmetric decompression acoustic loading of the 1.
annular region between the vessel and shroud from the pipe break at or beyond the vesse] nozzle safe-end weld.
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A transient asymmetric Jifferential pressure within the 2.
annular region between the biological shield wall and the reactor pressure vessel (annulus pressurization).
i A jet stream release of the reactor pressure vessel inventory 3.
and the impact of the ruptured pipe against the whip restraint attached to the biological shield wall.
The results of the mass and energy release evaluation are then used to produce a dynanic structural analysis (force time history) of the RPV The force time history output from the dynamic analysis and shield wall.
The is subsequently used to compute loads on the reactor components.
followinp is a more detailed description of the annulus pressurization calculation performed for the Shoreham fluclear power station.
The GE nethodology for calculating the mass-energy release from a recirculation line break which results in an annulus pressurization event was provided the NRC's Mr. Denwood F. Ross, Assistent Director for Reactor Safety via GE letter dated May 2, 1978 from Mr. E
- 9. Fuller of BWR Licensing.
This nethodology was used in the adequacy assessment made for SNPS.
A description of the time aspects of the calculated mass and energy flow rates, followed by a description of the modeling for the feedwater line A comparison and separately for the recirculation line is provided below.
is made between GE's analytical method and the methud used in RELAP 4/ MOD 5.
Finally, both nraphical and numerical results of this comparison are provided to substantiate the conclusion that the resulting break flows using the GE nethods are much more conservative than those predicted by use of RELAP for SNPS..
The GE method for calculating the short term nass energy release assumes that the overall time for mass release may be div; Jed into two periods; the inventory period and the quasi-steady period.
The inventory period is defined as the time required to accelerate the pipe flui'd to steady-state velocities at which time the flow is assumed to choke at minimum flow cross sections.
During this time the mass flux is based on. initial In the quasi-steady thernodynamic condition existing within the pipe.
period, during which the flow is choked, the mass flux is based onFor both time thermodynamic conditions up-stream from the choke points.
periods the mass flux is determined from a graph of critical mass flux Each side J
versus enthalpy, as calculated by the Moody Slip Flow Method.
of the break is analyzed separately and the results summed to give the J
tj total mass release rate.
Comparison of General Electric Analysis to RELAP 4/M00 5 For the annulus pressurization event the '!RC has cuestioned General Electric's nethod for computing mass and eneroy flow rates followina a postulated LOCA from long lines containino subcooled fluid. A program was developed to expedite the licensina of the Shoreham Nuclear Power station to perforn RELAP analyses usina a,nprooriate assumptions and "to compare the results with those 'obtained usina General Electric's nethod.
RELAP 4/ MOD 5 is a aeneral computer procram which can be used to analyze the therral hydraulic transient behavior of a water-cooled nuclear reactor subjected to postulated accidents such as loss of coolant accidents.
The pronran simultaneously solves the fluid flow, heat transfer, and reactor kinetics ecuations describino the behavior of the reactor.
Numerical input data is utilized to describe the initial conditions and geometry of the systen beina analyzed.
This data acludes fluid volume, peometry, pump characteristics, power generation, heat exchanner properties, and nodalization of fluid flow paths.
Once the systen has been described with initial flow, pressure, temperature, and power level boundary conditions, transients such as a loss of coolant accident can be sinulated by control action inputs.
RELAP then computes fluid conditions such as flow, pressure, nass inventory and ouality as a function of time.
For the brief transients considered here (t < 0.5 seconds), appreciable sirplification of the overall thern31-hydraulic system, includino the reactor pressure vessel, is justified owing to the relatively lonner time-constants which apply for heat transfer.
Brief sunnaries of the modeling-approaches for feedwater and recirculation line breaks are given below.
The assumptions (or ground rules) applied to these analyses are as follows:
a.
Feedwater line:
1)
Shorehan RELAP deck as basis.
2)
Use Henry-Fauske-Moody flow nodel.
3)
Instant break openina.
4)
Elininate mass flux terns between vessel and break (short side),
b.
Recirculation line:
1)
Shoreham RELAP deck as basis.
2)
Allow for finite break opening time.
3)
Use Henry-Fauske-Noody flow model.
4)
Eliminate momentum flux terns in RELAP between vessel and break (short side).
9
Method for Feedwater Line Modeling The feedwater system for Shoreham Muclear Power station consists of the pumps, heaters, valves, and piping necessary for the transmission of hotwell condensate to the reactor vessel as part of the closed cycle cooling loop. -In the Shoreham Nuclear Power station there are two feedwater pumps.
The flow passes through a complex series of pipes and components from the feedwater pumps to.the reactor vessel.
The break location for the feedwater line break is the safe end to the pipe weld housed within the vessel to shield wall subcompartment.
For the feedwater line break instantaneous break opening is assumed.
Flow for the vessel side passes through the feedwater nozzles of the broken line and out the break.
Flow from.the system side passes through the feedwater piping network and out the break.
The feedwater check valves was conservatively neglected in the analysis.
The nodalization of the feedwater system is shown in Figure 1.
A series of 16 modes was selected after sensitivity studies were completed which demonstrated that a 16 node model was conservative relative to higher noded systems.
The broken feedwater leg to be analyzed was chosen by multiple RELAP.
runs to determine the limiting break location.
The critical astumptions in the analysis are as follows:
1)
The feedwater pumps are simulated as (constant) mass flow sources.
2)
The reactor pressure vessel (RPV) is an infinite reservoir at constant temperature and pressure.
3)
The temperature of the pump-side hydraulic network remains constant.
4)
Appropriate sections of the hydraulic network are combined by means of " ohm's law" expressions for series and parallel circuits, assuming constant fanning friction actions.
5)
The RPV thermodynamics state is subcooled at the prevailing tempera-ture in the lower plenum (532*F).
6)
The Hoody Critical Flow model is used for a saturated thermodynamic state, and the Henry-Fauske Critical Flow model is used for a 1
j subcooled thermodynamic state.
The break is modeled as an instantaneous guillotine pipe break with complete pipe offset.
Before the break occurs, a fully open valve connects volumes 10 and 12.
Closed valves connect those volumes to volume 11, an infinite sink at constant pressure and temperature (atmospheric conditions).
The break is initiated at time zero by closure of the valve between volumes 10 and 12 and simultaneous opening of the valves to volume 11.
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Method for Recirculation Line liodeling The recirculation system for Shoreham Nuclear Power station is similar to the recirculation system of other Boiling Water Reactors.
Flow is taken from the lower jet pump diffuser region, passed through 28" lines to a constant speed pump, then through a flow control valve to a header which feeds flow to five risers which provide flow to two jet pump nozzles each.
The nodalization for the recirculation line leak is shown in Figure 2.
The system has been modeled using 21 nodes.
The break is located at the vessel nozzle safe-end to pipe wel.d on the recirculation pump suction side.
The type of break considered here has a finite break opening time.
For this case the break opening is -complete af ter 33 milliseconds at which time the pipe offset longitudinal distance is 9.25 inches.
The break area is modeled as the surface area of an imaginary volume having a length of 9.25 inches and a diameter equal to that of the recirculation pipe 10.
This volume (#18) is connected by a valve (type 3) to an infinite reservoir (volume #19), and also by valves (type 3) to the vessel side volume (#17) and pump side volume (#21).
Another valve (type 1) also connects volumes 17 to 21.
It is normally open before the break, and at the initiation of the break, closes at the same rate as the other valves open.
The flow area of the type 3 valves connecting volumes 17 and 21 to volume 18 are equal to the pipe cross sectional area and the flow area of the type 3 valve connecting volume 18 to 19 is equal to the break area previously described.
This network of valves best represents the break with finite opening time.
Valves of type 3 are all opened at the same rate to insure that choking occurs at junctions 21 and/or 22 in addition to junction 23.
Junction 23 (having valve type 3) is in reality a fluid surface and choking cannot physically occur there.
Choking must at least occur at junctions 21 and/or 22 where the fluid is constrained by the pipe diameter.
Other assumptions in the analysis include:
1)
Negligible effects of core reactor kinetics on rated heat transfer to the core volume (volume 2).
4 2)
Constant flow of steam from the steam dome (volume 5) at rated conditions.
3)
Constant flow of feedwater at rated conditions.
4)
Recirculation pumps trip at time zero and are modeled via pump characteristic curves for coastdown.
5)
Jet pump hydraulics were modeled as one equivalent pump per re-circulation loop.
6)
The Moody Critical Flow model is used for a saturated thermodynamic state, and the Henry-Fauske Critical Flow model is used for a subcooled thermodynamic state.
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The mass release result for the G.E.
Mass Energy Release method and the RELAP IV/ Mod 5 calculations are compared in Figures 3 and 4 for the postulated feedwater line break
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and recirculation line break respectively.
Also compared in tables 1 and 2 are break area and mass flux versus time for the Feedwater and Recirculation line breaks respectively.
The analyses show that the G.E. method is conservative relative to RELAP IV/ Mod 5 for both cases.
The ratio (r) of the G.E. method flow rates to those from RELAP Mod 5 is as follows:
l BREAK TYPE r(t=0.1 sec.)
r(t=1.0 sec.)
Feedwater 2.2 1.5 Recirculation Line 1.3 1.3 Additional conservatisms were included in the development of mass and energy release rates used in the assessment of the Shoreham shield wall and are listed as follows:
1.
No credit was taken for finite break opening time for the recirculation suction break.
Shoreham has assumed a linearly increasing blowdown rate from t=0.0 to t=10.0 msec which is considerably more conservative than the finite break opening case described above.
2.
For the recirculation suction break, an instantaneous isolation / SCRAM was assumed which increased the reservoir pressure during the transient.
3.
For the feedwater break, no choke point was assumed other than the feedwater line cross-sectional area at the break itself.
The post-inventory mass and energy release rates are somewhat more i
conservative than those during the inventory period and were applied from t=0.0.
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4.
At the time the feedwater analysis was done, a
20 percent margin was applied to the subcooled mass flow rates for the feedwater break due j
to uncertainty in the application of the Moody model to highly subcooled flows.
This margin has been shown to be unnecessary and represents j
conservatism.
8 The actual mass and energy release rates used in the pressurization analysis of the Shoreham shield wall annulus due to feedwater and recirculation suction breaks are given in Shoreham FSAR f
Tables 6.2.1-6 and 6.2.1-9 respectively.
The break locations and discussion of blowdown allocation are given in Section 6.2.1.1.4.2.
4
4 BREAK AREAS AND MASS FLUXES VS. TIME FOR SHOREHAM STATION FEEDWATER LINE BREAK (Safe End to Nozzie Weld)
Reactor Pressure Pump Side Vessel Side 2
2 2
Time (sec)
A(ft )
G(lbm/ft sec)
A(ft )
G(lbm/ft sec) 0.0 0.0 0.0 0.0 0.0 0.005 0.6674 3079 0.6674 4311 0.010 4631 4278 0.920 6119 4206 0.040 6437 4004 0.060 6823 4020 0.080 6770 4056 0.100 6399 4082 0.150 5175 4134 0.200 4435 6332 0.250 4219 7201 0.300 4218 7369 0.350 4218 8027 0.400 4216 8511 0.450 4216 8028 0.500 4216 8804 0.600 4215 9230 0.700 4213 10,160 0.800 4212 10,580 0.900 4212 11,310 1.000 4210 10,980 TABLE 1 i
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4 BREAK AREAS ANO MASS FLUXES VS. TIME
^
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FOR SH0REHAM STATION RECIRCULATION LINE BREAK (Safe End of Suction Line Weld)
Reactor Pressure Pump Side vessel Side 2
2 2
2 Time (sec)
A( f t )
G(1bm/ftsecl A(ft)
G(lbm/ft sec) 0.0 0.0 0.0 0.0 0.0 0.005 0.095 12,370 0.095
-5060 0.010 0.371 8,430 0.371
-3720 0.020 1.300 5,680 1.300
-3045 0.033 2.55 9,300 2.55
-6110 0.040 10,180
-6300 0.060 13,040
- 310 0.080 13,670 1620 0.100 13,940 585 0.150 13,880
- 250 0.200 13,990
- 735 0.250 13,960
-1155 0.300 13,780
-1340 0.350 13,770
-1260 0.400 13,720 1240 0.500 13,530
_ 910 0.600 13,540
- 550 0.700 13,490
- 540 0.800 13,370
- 660 0.900 13,250
_ 740 1.000 13,160
- 740 TABLE 2
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