ML19259A762
| ML19259A762 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 01/03/1979 |
| From: | Goodwin C PORTLAND GENERAL ELECTRIC CO. |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| TAC-08291, TAC-8291, NUDOCS 7901100273 | |
| Download: ML19259A762 (24) | |
Text
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- Ell PortlandGeneralSectricCompany Chm es Goode.n. Jr A: wint '. e Re9M January 3, 1978 Trojan Nuclear Plant Docket 50-344 License NPF-1 Director of Nuclear Reactor Regulation ATTN:
Mr. A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors U. S. Nuclear Regulatory C amit. icn Washington, D. C.
20555
Dear Sir:
Your letter of November 28, 1978 regarding Containment purging during normal Plant operation rs quired several actions of us.
We have completed those actions as follows.
Currently, we are limiting Containment purging to not exceed 90 hr per year while the Plant is in operations.1 Modes 1, 2, 3 and 4.
We have established administrative controls to ensure this is accomplished.
A Plant Administrative Order and Radioactive Discharge Permit procedures have been revised to ent,ure that purging is limited to no more than 90 hr per year. Purgin g is restricted to those occasions when purging is deemed necessary for personnel exposure reduction and will not be conducted for temperatere or humidity control. Operation of the Hydrogen Vent System during Mode s 1, 2, 3 and 4 for Containment pressure control is not included in this purge limitation. provides i. detailed description of the basis and desira-bility for Containment purging during Modes 1, 2, 3 and 4, taking into account operational capabilities, radiological controls, respiratory requirements and safety considerations.
We have completed a review of the purge isolation circuitry and have determined that a problem similar to that experienced at Millstone Unit 2 and Salem Unit 1 could not occur at the Trojan Nuclear Plant. Capability for manual override of the high radiation signal of the Containment 0
09 \\
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POttland General BOCiliCCOn4XNif Mr. A. Schwencer January 3, 1978 Page two purge valves does not exist.
Furthermore, a Containment Isolation Signal (CIS) will cause the Containment purge valves to close under all conditions. There is no easy menner in which the Containment purge valves could be overridden to permit purging when a high radiation level exists in the Containment.
In addition, these valves can not be opened if either the high radiation signal or the CIS are present.
We have performed a detailed evaluation on the issues relating to purg-ing during normal operation, as required by Standard Review Plan 6.2.4, Revision 1, and the associated Branch Technical Position CSB 6-4.
This evaluation is provided as Attachment 2.
The responses are based on information contained in the Trojan Fiaal Safety Analysis Report (FSAR).
We have considered valve operability, performance of the ECCS, and radiological consequences during purge operations simultaneous with the design basis loss-of-coolant accident. Two items not provided at this time are our responses to Items 1.g and 5.b of Branch Technical Position CSB 6-4 which require additional analyses. We plan to respond to these two items no later than April 1,1979; and, if necessary, to propose a change to the Trojan Technical Specifications at that time.
We have also completed a review of all Engineered Safety Features Actu-ation System (ESFAS) circuits incorporating a manual override feature to ensure that overriding of one ESFAS does not also cause the bypass of any other ESFAS signal. Attachment 3 is a table listing the manual overrides for the ESFAS circuits.
The description, switch location for each override, and annunciation status is listed. There are no instances when overriding of one ESFAS circuit will cause the bypass of any other ESFAS signal. As a result of this analysis, we do not feel it is neces-sary to implement any additional administrative controls to prevent improper manual defeat of ESFAS signal as a part of our Plant operation.
In conclusion, we feel information provided by this letter and its attachments should provide a sufficient basis for you to justify Con-tainment purging at the Trojan Nuclear Plant for up to 90 hr per year in operational Modes 1, 2, 3 and 4, pending completion of the NRC review.
Sincerely d
s C. Goodwin, Jr.
Assistant Vice President Thermal Plant Operation &
Fbintenance CG/GAZ/4kkSA19 Attachments
ATTACHMENT 1 BASIS FOR CONTAINMENT PURGING Loss of the capability of Containment purging or even a restriction to 90 hr per yeet during operation in Modes 1, 2, 3, and 4 have serious implications upon the Trojan Nuclear Plant. These irplications are further described as follows:
1.
Operational Capabilities Fuel element failure concurrent with primary leakage in the Containment could result in airborne concentra-tions of radioactivity that exceed the MPC limitations of our respiratory protection equipment (iodine and parti-calates) or that result in significant external dose rates (noble gases).
Lack of the ability to purge would then require waiting until the Plant could be cooled to Mode S before attempts could be made to repair and recover the Plant. This would lead to a significant increase in Plant inavailability, a loss of an important energy source in the Pacific Northwest.
A second operational limitatie.. could occur during sched-uled outages. A lack of ability to purge could increase the amount of time to cool the Plant down to Mode 5.
Purging the Containment could help to reduce the ambient temperature and decrease the amount of time required for cooldown.
Longer cooldowns add costs and delays to the critical path for scheduled outages.
2.
Radiological Controls Inability to purge could increase the airborne activity levelt and could result in an increase in exposures and contamination levels within the Containment. This is in direct conflict with and contradiction to existing ALARA commitments, which is not tolerable from a health physics standpoint.
3.
Respiratory Requirements The lack of ability to purge would increase airborne concentrations to the point where use of respirators could be required for all Containment access. This would conflict with requirements of 10 CFR 20 and NUREG-0041, which require every attempt be made to " engineer out" the use of respirators in all possible situations by proper ventilation and purification techniques. Thus the inability to purge would require us to violate other regulations and guidelines, suffering a less of personnel efficiency and experiencing significant increases in the cost of maintaining radiation protection equipment.
e 4.
Safety Considerations Containment purges are sometimes necessarily performed to provide fresh air and cooling for the safety and health of personnel. Without purging, workers would be required to work in full anti-C's and respirators in areas with temperatures in excess of Il0*F.
No t only is this a poor practice from a health and safety viewpoint, but it also causes a loss in efficiency.
For example, Scott air packs have a 30-min use time limit. With 10 min required for ingress and egress from a work station in the Containment, there would be a direct 30 percent loss in work time.
Fur the rmore,
the overall loss of efficiency of the workers is at least 50 percent when respirators are worn.
The Containment was purged a total of 251 hr during 1977 while in Modes 1, 2, 3 and 4.
Although that number could have been reduced, it should be realized that 1977 was a relatively trouble-free year.
Since Trojan would have had trouble meeting the 90-hr purge limit in a relatively trouble-free year, this could present difficulties when and if probleas develop.
Therefore, a 90 hr per year purge limit for the life of the Plant could cause serious operational problems and be a costly requirement.
GAZ/4kkSA22 ATTACHMENT 2 RESPONSES TO REQUIREMENTS OF BRANCH TECHNICAL POSITION CSB-6.4 Branch Technical Position The system used to purge the Containment for the reactor operational modes of power operation, startup, hot standby and hot shutdown; i.e.,
the on-line purge system, should be independent of the purge system used for the reactor operational modes of cold shutdown and refueling.
Response
The Containment Purge System (used for cold shutdown and refueling) is independent of the Hydrogen Vent System (the on-line purge system), with separate blowers, filters, Containment isolation valves and suction ducting.
Common exhaust ducting downstream of the blowers is utilized.
Refer to FSAR Figure 6.2-48.
The Containment Purge System may be used iafrequently during power operation, startup, hot standby and hot shutdown for reduction of noble gas dose rates to personnel entering Containment.
1.
The on-line purge system should be designed in accordance with the following criteria:
a.
The performance and reliability of the purge system isolation valves should be consistent with the operability assurance program out-lined in MEB Branch Technical Position MEB-2, Pump and Valve Operability Assurance Program.
(Also see SRP Section 3.9.3.)
The design basis for the valves and actuators should include the buildup of containment pressure for the LOCA break spectrum, and the purge line and vent line flows as a function of time up to and during valve closure.
, Response The performance and reliability of the purge system isolation valves is consistent with the operability assurance program outlined in BTP MEB-2, Pump and Valve Operability Assurance Program, with respect to the ability of the valves to close under the pressure and flow conditions for the LOCA break spectrum.
Both the 54-in. Containment Purge System and the 8-in. Hydrogen Vent System inlet and exhaust ealves were tested by the valve manufacturer for closure under a 75 psig static pressure condition and were shown to meet the required closure time criteria (see FSAR Table 6.2-1).
because of the design of a butterfly valve (in-line center pivot point),
the dynamic forces generated are negated, and this method of testing is acceptable.
In addi-tion, the valves have been adequately designed to close against bearing friction created by a 60 psi design basis LOCA differential pressure.
b.
The number of purge and vent lines that may be used should be limited to one purge line and one vent line.
Response
The Containment Purge System at the Trojan Nuclear Plant contains one 54-in.
Lameter purge (exhaust) line and one 54-in. diameter vent (supply) line.
The Hydrogen Vent System, which is used for low-flow rate purges for Containment pressure control, consists of two redundant trains (140 cfm each) with 8-in. diameter exhaust lines and B-in.
diameter supply lines (see FSAR Figure 6.2-48).
Both trains may be operated simultaneously.
c.
The size of the purge and vent lines should not exceed about eight inches in diameter unless detailed justification for larger line sizes is provided.
Response
The Hydrogen Vent System, which is used for low-flow rate purges for Containment pressure control, has purge and vent lines of 8-in. diameter. The Containment Purge System contains 54-in. purge and vent lines.
The justification for this system is presented in the responses to the other items in this attachment, and in Attachment 1.
d.
The containment isolation provisions for the purge system lines should meet the stendards appropriate to engineered safety features; i.e.,
quality, re-dundancy, testability and other appropriate criteria.
Response
The Containment isclation provisiens for the purge system lines meet the standards appropriate to engineered safety 'tatures.
The design bases for the Containment isolation components include pro-vision for the following:
(1) A double barrier at the Containment penetration in those fluid systems that are not required to function following a design basis event. The Containment Purge System and the Hydrogen Vent System have redundant valves on both intake and exhaust lines.
Piping that forms part of the Contain-ment isolation boundary is designed to ANSI B31.7, 1969 Nuclear Power Piping Code. Valves are designed and fabricated to Class 2 requirements of the Draft ASME Code for Pumps and Vc tes for Nuclear Power, November 1968. All automatic valves that are activated by a CIS also
% <e handswitches in the control room for manual actuation to close in the event the valve fails to go to the closed position.
All Containment isolation components were designed, fabricated, and tested under quality assurance requirements in accordance with 10 CFR 50, Appendix B.
Containment isolation valves are designed to Seismic Category I requirements.
The valves are capable of operation during and after seismic loadings.
(2) Automatic, fast, and efficient closure of those valves required to close for Containment integrity following a design basis event to minimize release of any radioactive material.
(3) A means of leak testing all barriers in fluid systems that serve as Con-tainment isolation. All piping systems penetrating the Containment have been provided with test vents and test con-nections or have other provisions to allow leak testing as required by Appendix J to 10 CFR 50.
(4) The capability to periodically test the operability of Containment isola-tion valves. Those automatic isola-tion valves with air or motor operators that do not restrict normal plant operation are periodically tested to ensure operability, according to Tech-nical Specific ation 3/4.6.3.
e.
Instrumentation and control systems provided to isolate the purge system lines should be indepen-dent and actuated by diverse parameters; eg, containment pressure, safety injection actuation, and containment radiation level.
If energy is require to close the valves, at least two diverse d
sources of energy shall be provided, either of which can affect the isolation fuaction.
Response
Containment purge line isolation occurs on a Containment Isolation Signal (CIS), manual operation or Containment high radiation level.
The CIS is actuated by high Containment pressure, high differential pressure between steam lines or pressurizer low prcssure concident with low water level. There are four independent Contain-ment pressure detectors connected to seven pressure switches.
Three of the pressure switches are combined to provide the signal for Containment isolation if the high-pressure setpoint is recched.
Containment high pressure setpoint is established at 5 psig Containment pressure.
In addition to
, diverse modes of operation, channel separation also maintained. This also ensures that the single-failure criterion is met (see FSAR Figure 7.2-1 (Sh 8) and Table 7.3-1 and -2].
The inboard 54-in. purge isolation valves are motor-operated, and the outboard 54-in. purge isolation valves are air-operated. All of the 8-in. Hydrogen Vent System isolation valves are motor-operated. Th t, air-operated isolation valves fail to the closed position in the event the air supply is lost or fails or upon receipt of a signal to close; no power is required. All motor-operated valven, ret aive power during normal conditions from the normal power scurce. Under loss of normal and preferred power conditions, power is supplied from the standby power source.
f.
Purge system isolation valve closure tLnes, including instrumentation delays, should not exceed five seconds.
Response
lhe Containment Purge System isolation valve closure times are 3 sec for the air-operated valves and 5 see for the motor-operated valves (see FSAR Table 6.2-1).
Instrumentation delays in the Engineered Safety Features Actuation System (ESFAS) will be a maximum of I sec (see FSAR Section 7.3.1.2).
The Hydrogen Vent System valves closure times are 5 see for all valves.
These design closure times were verified by test-ing the valves for closure under a 75 psig static pressure condition (see response to Item la).
g., Provisions should be made to ensure that isola-tion valve closure will not be prevented by debris which could potentially become entrained in the escaping air and steam.
Response
Review of this item has not been completed.
2.
The purge system should not be relied on for temperature and humidity control within the containment.
Resporise The Containment Purge System is not used for temperature and humidity centrol within the Containment.
The Hydrogen Vent System is used for Containment pretsure control.
3.
Provisions should be made to minimize the need for purging of the containment by providing containment atmosphere cleanup systems within the containment.
Response
Fission product cleanup and control capabilities within the Containment are available duricg normal operation from two non-ESF Seismic Category II 4000 cfm recirculating filter units,ployiny prefilters, HEPA filters, and carbon adsorbers, which are available for intermittent or continuous operation to contrcl the build-up of airborne halogens and particulates which result from small primary coolant system leaks within the Containment.
The objec-tive of these cleanup filters (in conjunction with the Containment purge system) is two-fold.
The first is to maintain Containment airborne fission product levels below the concentration limits of 10 CFR 20 for occupational exp.osures, thus permitting safe access to the Containment as required. The second is to reduce fission product releases to the environment to levels as low as practicable (meeting the intent of 10 CFR 50 Appendix I) when Cantain-ment purging is required for access.
This system is not ef fective in removing noble ga; so-topes from the Containment atmosphere; if required for personnel entry, noble gases are reduced by operation of the Containment Purge System.
4.
Provisions should be made for testing the availability of the isolation function and the leakage rate of the isola-tion valves, individually, during reactor operation.
Response
Subsequent to initial plant operation, Containment isola-lation components will be periodically operated or tested as outlined in Technical Specification 3/4.6.3.
Test connections and pressurizing means are provided to test each isolation valve or barrier that is required to be Incally leak tested per Appendix J of 10 CFR 50.
Either nitrogen, air or water is used as the pressurizing medium, depending on the physical location and service of each line. Leak testing of Containment Purge System and liydro-gen Vent System valves will be accomplished by:
(1) pres-sure decay, or (2) Containment integrated leak test.
Containment isolation valves are also functionally tested as per Technical Specification requirements (see Techni-cal Specification 3/4.6.3.1.1 and Table 3.6-1).
Current proposed amendments to the Trojan Technical Specifications will require functional testing as per the requirementa of 10 CFR 50.55a(g), as part of the Trojan pump and valve In-Service Inspection Program (ISI).
- 5.. The following analyses should be performed to justify the Containment purge system design:
a.
An analysis of the radiological consequences of a loss-of-coolant accident.
The analysis should be done for a spectrum of break sizes, and the instrumentation and setpoint that will actuate the vent and purge valves closed snould be identi-fied.
The source term used in the radiological calculations should be based on a calculation under the terms of Appendix K to determine the extent of fuel failure and the concomitant release of fission products, and the fission product activity in the primary coolant. A pre-existing iodine spike should be considered in determining primary coolant activity. The volume of contain-ment in which fission products are mixed should be justified, and the fission product < from the above sources should be assumed to be released through the open purge valves during the maximum interval required for valve closure. The zadio-logical consequences should be within 10 CFR 100 guideline values.
Response
The radiological consequences of a Loss-Of-Coolant Acci-dent (LOCA) if the 54-in. Containment purge air supply and exhaust valves were open at the t ime of the accident were evaluated for the design basis LOCA (double-ended pump suction break).
The valves were conservatively assumed to be full open for 2.0 sec (1.0 sec for genera-tion of a high Containment pressure alarm (5 psig) and subsequent Containment Isolation Signal (CIS), plus a maximum CIS response time of 1.0 sec).
In accordance with the valve operation time specified for Containment isola-tion in FSAR Table 6.2-1, the valves were assumed to close in 5.0 sec.
This is a conservative assumption, since the outer, air-operated purge valves are designed to close in 3.0 sec.
The open area was assumed to be reduced by rotation of the valve stems as a linear func-tion of time.
The analysis was performed using the CONTEMPT-PS computer code.
Purge valve leakage rates were calculated using an ideal gas model for both suberitical and critical flow.
The following additional assumptions were made in the analysis:
(1) Double-ended pump suction guillotine break blowdown data was used (FSAR Table 6.2-9).
(2) Containment net free volume was that used in the DBA analysis (FSAR Table 15.4-3c).
(3) A flow coefficient of 1.0 assumed during the time that the valves were open.
(4) The blowdown mass was assumed to instan-taneously mix with the initial Containment atmosphere.
(5) The Containment initial conditions were:
Pressure, psia 14.7 Inside Temperature, 'F 120.0 Relative Humidity, %
53.0 The results of the analysis were that 11,409 lb of Contain-ment atmosphere would be released to the environment before the Containment is isolated. Releases of radioactive material through the purge isolation valves after this LOCA has been evaluated with the following equation:
I P
A=A L(t) f(t) dt
)
o wh ere A = Activity release through valves, Ci L(t) = Fractional leakage rate through valves, sec-1 f(*) = Cumulative fraction of reactor coolant ac t ivi ty released into Containment Initial inventory of activity in reactor A
=
coolant, Ci Time valves are closed after LOCA, 7 sec.
T =
The applicable source term for this analysis is the activity of radiciodine and noble gases in reactor coolant at the time of initiation of the LOCA.
Inventories of these isotopes in reactor coolant are given in Table 1.
The noble gas concentrations are based on steady-state opera-tion with I percent fuel defects while the iodine concen-tration is based on the Technical Specification limit of 60 pCi/gm dose-equivalent I-131 for a pre-existing iodine spike.
The mass of reactor coolant is 522,700 lb.
Tran-sient fuel cladding f ailures would not occur until approxi-mately 28 sec after the LOCA (see FSAR Table 15.4-2) so no fuel rod or gap activity would be released during the 7 see that the purge valves are open.
Fractional leakage rates through the purge isolation valves have been computed from the previously described Contain-ment analysis.
These leakage rates are based on the mass of air initially present in the Containment atmosphere and the mass rate of flow of air out the valves.
Values of F(t) and L(t) are listed in Table 2.
Using these values, an estimated 1.47 percent of the initial inventory of activity in reactor coolant is released through the purge valves prior to closure.
Offsite doses at the Exclusion Area Boundary (662m), from the leakage of primary coolant from the open purge valves were evaluated using the models and assumptions of Regula-tory Guide 1.4.
The fifth percentile atmospheric dispersion
-4 3
factor (X/Q) was used (4.26 x 10 sec/m ), and the
-4 3
breathing rate was assumed to be 3.47 x 10 m /sec.
The resulting 2-hr doses are 45.7 rem thyroid and 0.013 rem whole body (beta + gamma).
Since the Appendix K analysis results in peak fuel tempera-tures well below the melting point of the fuel rods, only the fuel rod gap inventory could be released to the Con-tainment following Containment isolation. The 2-hr Exclusion Area Boundary doses from the a LOCA with gap release (with Containment isolation) are 1.05 rem thyroid and 0.39 rem whole body (see FSAR Table 15.5-7).
Adding the above doses gives a total dose from the DBA LOCA with the 54-in. purge valves open at the time of the accident of 46.8 rem thyroid and 0.40 rem whole body, which are less than the 10 CFR 100 guidelines.
The consequences of a LOCA with the 8-in. hydrogen vent system air supply and exhaust valves open was also examined.
Since the hydrogen vent valves are designed to close within the same time as the 54-in. purge valves, the consequences of a LOCA wion these valves open would be substantially less than calculated above.
b.
An analysis which demonstrates the acceptability of the provisions made to protect structures and safety-related equipment; e.g.,
fans, filters and ductwork, located beyond the purge system isolation valves against loss of function from the environment created by the escaping air and steam.
Response
Review of this item has not been completed.
c.
An analysis of the reduction in the Coatainment pressure resulting from the partial loss of Con-tainment atmosphere during the accident for ECCS backpressure determination.
Response
An analysis of the reduction in the Containment pressure was performed for the cases where (1) the Containment is isolated, and (2) the 54-in. purge air supply and exhaust lines are initially open and close on a Containment isolation signal. The sensitivity of the predicted Peak Cladding Tempera-ture (PCT) for a LOCA to the average Containment pressure during the period from the beginning of reflood until the time of PCT was used to deter-mine the impact on ECCS performance.
The predicted PCT as a function of the above average Containment pressure was obtained from SARs for five Westinghouse plants with different Containment volumes. The data abstracted for these five plants is presented in Table 3.
The change in PCT as a function of the change in Containment backpressure based on this data can be represented as a best-fit s traigh t line as:
A PCT = 21.28 x A P - 8.46 where APCT = increase in PCT from base case, *F.
AP = reduction in Containment back-pressure (average during time from beginning of reflood to PCT) from base case, psi.
The change in the Containment pressure for the case where the purge air supply and exhaust valves are initially open as compared to the base case where the Containment is isolated was deter-mined in a conservative manner using the CONTEMPT-PS code.
The specific conservatisms incorporated in the analyses were:
(1) The Containment DBA pressurization assump-tions were used instead of LOCA Containment backpressure assumptions (yielding a larger change in pressure because the higher Con-tainment pressures cause more leakage).
(2) The purge valves were closed in a step-wise manner at 5.5 see following the LOCA (0.5 cec for Containment isolation signal initiation and 5 see for isolation valve closure, which is the maximum time presented in FSAR Table 6.2-1 for these penetrations).
This is a conservative assuur tion, since the purge valves would close over the entire 5 sec interval, rather than in a stepwise manner at 5 sec.
Even if instrumentation delays (maxi-mum of I see) are included, a stepwise 5 sec closure time is conservative.
In addition, the outer, air-operated purge valves are designed to close in 3 sec, as opposed to 5 see for the inner, motor-operated valves.
The 5 sec closure time used in this analysis assumes failure of the air-operated valves in the open position.
),
s
The results from the CONTEMPT-PS analyses indicates that the impact on the pressure of having the purge valves initially open is a reduction in Containment backpressure of no more than AP = 2.5 psi. This translates into a maximum increase in PCT of 45'F.
A more realistic assessment of a LOCA with purging in progress would more than likely indicate that the impact on predicted PCT would be considerably less.
The current Trojan LOCA PCT is predicted to be 1974*F, which was calculated with the approved " October 1975" version of the Westinghouse Evaluation Model with appropriate corrections for upper head water temperature (results transmitted to the NRC by letter on November 31, 1977). When consideeing this PCT and the conservatively calculated increase of 45'F in PCT as a result of the purge valves being initially opened, there still is significant margin to the limiting PCT value of 2200*F.
d.
The allowable leak rates of the purge and vent isolation valves should be specified for the spectrum of design basis pressures and flows against which the valves must close.
Response
The allowable leak rate was specified for design basis differential pressure and flow in the pur-chasing Technical Specification. Manufacturer's Certificates of Conformance are part of the valves QA documentation.
The valves are tested as part of the Trojan Inte-grated Leak Rate Test procedure and the allowable leakage is specified as part of the total Contain-ment leakage as defined in 10 CFR 50, Appendix J.
The valves are incorporated in the Trojan Pump and Valve Inservice Inspection (ISI) and allow-able leak rates are specified in this program.
The Pump and Valve ISI program for the Trojan Nuclear Plant commences on September 20, 1979.
The design of a butterfly valve is such that the forces of flow are cancelled around the disc center-line pivot point.
The valves have been adequately designed to close against bearing friction created by a 60-psi design basis LOCA differential pressurc. SGG/c rw66.16B2
TABLE 1 REACTOR COOLANT INVENTORY Activity Isotope (Curies) 4 I-131 ^
1.42 x 10 I
Kr-83m 9.73 x 10 2
Kr-85 2.47 x 10 2
Kr-85m 5.60 e 10 2
Kr-87 3.20 x 10 2
Kr-88 9.84 x 10 2
Xe-131m 4.65 x 10 4
Xe-133 3.94 x 10 2
Xe-133m 6.32 x 10 3
Xe-135 2.73 x 10 1
Xe-135m 4.76 x 10 2
Xe-138 1.40 x 10
[a] Dose - equivalent I-131 SCG/crw66.16B17
TABLE 2 VALUES OF L(t) AND f(t) FOR 7 SEC PURGE VALVE CLOSURE t(sec)
L(t) f(t)
-3 0-0.75 4.97 x 10 7.46 x 10~
-3
-1 0.75-1.25 8.13 x 10 1.3T x 10
-2
-1 1.23-1.75 1.03 x 10 1.94 x 10
-2
-1 1.75-2.25 1.18 x 10 2.47 x 10
-2
-1 2.25-2.75 1.14 x 10 2.95 x 10
-2
-1 2.75-3.25 1.03 x 10 3.38 x 10
-3
-1 3.25-3.75 9.10 x 10 3.76 x 10
-1 3.75-4.25 7.79 x 10~
4.10 x 10
-3
-1 4.25-4.75 6.48 x 10 4.42 x 10
-3
-1 4.75-5.25 5.17 x 10 4.72 x 10
-3
-1 5.25-5.75 3.93 x 10 5.00 x 10
-3
~I 5.75-6.25 2.69 x 10 5.28 x 10
-3
~I 6.25-6.75 1.34 x 10 5.54 x 10 6.75-7.25 0.0 SGG/crw66.16B18
e TABLE 3, SENSITIVITY OF PEAK CLADDING TEMPERATURE TO CONTAINMENT SACKPRESSURE FROM SAR CALCULATED RESULTS FOR FIVE WESTINC110USE PLANTS WITH DIFFERENT CONTAINMENT VOLUMES Average Pressure Change in Change (From Average in PCT Beginning Pressure From Number Containment of Reflood Maximum From Base Base of Power Volume to PCT)
Pressure PCT Case Case Plant Loops (MW)
Fuel (ft3)
(psig)
(psig)
(*F)
(psi)
(*F) 6 Trojan 4
3411 17 x 17 2.165 x 10 28 37.75 1885
+7.25
-145 6
Diablo Canyon #2 4
3411 17 x 17 2.66 x 10 21.125 26.8 1931
+0.375
-99 6
Diablo Canyon #1 4
3411 17 v 17 2.66 x 10 20.75 26.1 2030 0
0 (Base Case) 6 Byron-Braidwood 4
3411 17 x 17 2.75 x 10 20.5 26.0 2060
-0.25
+30 (jd RESAR 3) 6 Seabrook 4
3411 17 x 17 3 x 10 14.25 21.2 2178
-6.5
+148 (NRC RESAR 3)
SGG/c rw66.16B19
ATTACHMENT 3 MANUAL OVERRIDES OF SAFETY SICNALS OR SYSTEMS Description Switch Location Annunciation Pressurizer Pressure SI-Train A and B C04 Yes Steamline SI -
Train A and B C04 Yes Containment Spray -
Train A and B C19 Yes Containment Isolation -
Train A and B C19 Yes Containment Ventilation -
Isolation - Train A and B C19 Yes Safety Injection -
Train A and 3 C19 Yes Feedwater Isolation -
Train A and B C19 Yes ANR/jf/5A10