ML20197J631
| ML20197J631 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 05/17/1986 |
| From: | George Thomas PUBLIC SERVICE CO. OF NEW HAMPSHIRE |
| To: | Noonan V Office of Nuclear Reactor Regulation |
| References | |
| SBN-1054, NUDOCS 8605200103 | |
| Download: ML20197J631 (25) | |
Text
_
George S. Thomas Vice Presider t-Nuclect Production PutWC Service of New Hampshire New Hampshire Yankee Division May 17, 1986 SBN-1054 T.F. B7.1.2 United States Nuclear Regulatory Commission Washington, DC 20555 Attention:
Mr. Vincent S. Noonan, Project Director PWR Project Directorate No. 5
References:
(a) Construction Permits CPPR-135 and CPPR-136, Docket Nos.
50-443 and 50-444.
(b) USNRC Letter dated March 13, 1986, "Seabrook Technical Specifications," V. S. Noonan to R. J. Harrison (c) PSNH Letter dated April 22, 1986, "Seabrook Proof and Review Technical Specifications," G. S. Thomas to l
V. S. Noonan
Subject:
Response to NRC Questions on Seabrook Station Technical Specifications
Dear Sir:
As requested, enclosed please find our response to staff questions concerning the Seabrook Station Technical Specifications. Should you have any questions regarding this matter, please contact Mr. Warren J. Hall at (603) 474-9574, extension 4046.
Very truly yours, d
George S. Thomas GST/cj b Enclosures cc: ASLB Service List 1
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8605200103 860517 ADOCK 05 % g3 PDR A
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b P.O. Box 300 + Seabrook,NHO3874 Telephone (603)474-9521 II
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Enclosure to SBN-1054 1.
REQUEST PSNH asked that the pressure for determining containment integrity be changed from 46.8 to 48.7 psig but did not provide a basis for the change in the comments on the Technical Specifications. The applicant is considering submitting the results of a further re-analysis to justify accepting a higher temperature in the refueling water storage tank that would result in a higher containment pressure.
RESPONSE
The initial change in the pressure, determining containment integrity, from 46.8 to 48.2 psig was provided to the staff in Amendment 58 which was submitted under SBN-999 dated April 11, 1986.
Further re-analysis has been performed to justify changing the RWST temperature to a higher valve. This re-analysis also resulted in a higher containment pressure. Attachment I to this enclosure contains marked up FSAR pages that are affected by the re-analysis and includes the changes in RWST temperature and containment pressure. A marked up copy of the affected Technical Specification changes with the appropriate corrections are being submitted with SBN-1055, dated May 17, 1986.
2.-
REQUEST PSNH asked that for the containment air lock the Technical Specifications be changed to allow entry by opening the outer door when the inner door is inoperable for the purpose of repairing the inner door. The staff 42 agreed to the change with a modification to limit the cumulative time to one hour per year for opening the outer door with and inoperable inner door.
RESPONSE
NHY agrees with the staff request to add the cumulative time of one hour per year for opening the outer door with an inoperable inner door. A marked up copy of the affected Technical Specification changes with the appropriate corrections are being submitted with SBN-1055, dated May 17, 1986.
3.
REQUEST PSNH agreed to provide justification for the operational limits on containment internal pressure.
RESPONSE
This information was transmitted to the staff by SBN-1035 dated May 7, 1986. w
4.
REOUEST Secondary system containment isolation valves have not yet been added to Table 3.6.2 CONTAINMENT ISOLATION VALVES.
RESPONSE
Secondary containment isolation valves have been identified and will be added to the Technical Specif! cation table. A marked up copy of the affected Technical Specification changes with the appropriate corrections are being submitted with SBN-1055, dated May 17, 1986.
5.
REQUEST PSNH asked that the time required to produce the design basis negative pressure in the annulus be increased from 1 to 5.4 minutes. The staff understood that offsite dose calculations assumed a time of 3.6 minutes.
PSNH agreed to determine the correct value.
RESPONSE
The significance of secondary containment pulldown time to a negative pressure of (-)0.25" W.G. lies in the calculated radiological effects of a LOCA. The analysis summarized in Appendix ISB of the FSAR assumes that all containment leakage (not categorized as bypass leakage) and other post accident releases within the secondary containment boundary (e.g.
ECCS Leakage) are filtered by the emergency exhaust system as soon as a negative pressure of (-)0.25" W.G. is established. Until recently, the pulldown time utilized for this analysis was 4 minutes.
Per Amendment 58 of the FSAR, certain parameters utilized in the post-accident radiological analysis were changed. Among the parameters changed was the secondary containment pulldown time which increased from 4 to 8 minutes. A UE&C analysis concludes that 2.6 minutes of exhaust fan / filter operation will be required to reduce the initial secondary containment positive pressure caused by primary containment expansion back to atmospheric.
Because primary containment expansion will obviously not be simulated during the pulldown test, we have subtracted this 2.6 minutes from the 8 minute required pulldown time to arrive at the Technical Specification value of 5.4 minutes.
6.
REQUEST The applicant agreed to verify specified dif ferential pressure for the containment spray pumps.
RESPONSE
NHY has performed a review of the UE&C design of the Containment Spray System. This review has determined that the value of 262 psid was calculated using the design parameters of the system and has been veri-fled to be within the parameters of the system analysis. A marked up copy of the affected Technical Specification changes with the appropriate corrections are being submitted with SBN-1055, dated May 17, 1986.. -
7.
REQUEST Resolve the discrepancy between the Overtemperature AT and overpower AT equations in FSAR Section 7.2.1.1 and Technical Specification Table 2.2-1.
RESPONSE
There was no discrepancy found.
A typographical error was noted in the Proof and Review Technical Specifications transmitted by Reference (b).
This error was corrected by marked up comments returned to the staff by Reference (c).
8.
REQUEST Resolve discrepancy on use of span limit trip setpoint of less than 3%.
RESPONSE
Discussions with Westinghouse indicates that the 3% span margin concern is not appropriate for Westinghouse supplied equipment. Data gathered to date indicates that the response of the supplied equipment (both transmitters and process racks) is essentf ally the same near the upper and lower limits of the hardware range as at mid-span, i.e., the calibration and drift characteristics at the upper and lower limits for the range are within the allowances made for these parameters in the setpoint calculations. The only protection function which has a theoretical trip setpoint within 3% of the measurement span limits is Steam Generator Water Level - Low Low.
This is due to the conservative value used for the environmental allowance (EA). EA includes a combined uncertainty for both elevated temperature and radiation exposure. The transient event this function is used for does not result in any signifi-cant radiation exposure.
It is therefore expected that the actual trip setpoint would occur before reaching 3% or less of the measurement spray.
9.
REQUEST Are all Category I variables included in Table 3.3-10?
RESPONSE
We have included all design Category I type A, B and C variables in the previously proposed Technical Specifications. We will revise Technical Specification Table 3.3-10 to include all Category 1 variables. Amend-ment 59, scheduled for submittal to the staff in early June, revises the FSAR to reflect these changes. A marked up copy of the affected Technical Specification changes with the appropriate corrections are being submitted with SBN-1055, dated May 17, 1986.
10.
REQUEST Resolve discrepancies between FSAR Table 7.2-3 and Technical Specification Table 3.3-2.
. RESPONSE Our review of these tables has found three inconsistencies. These items are listed below along with the necessary corrective actions to resolve the inconsistencies.
1.
Power Range, Neutron Flux High Negative Rate - The response time value of _< 0.5 seconds given in Table 3.3-2 is correct. Item 5 in FSAR Table 7.2-3 will be revised to reflect this change. Attachment 2 to this enclosure provides a marked up copy of the affected FSAR page.
2.
Turbine Trip - The response time value of N.A. shown in Table 3.3-2 is correct. Item 15 in FSAR Table 7.2-3 will be revised to reflect this change. Attachment 2 to this enclosure provides a marked up copy of the affected FSAR page.
3.
Overtemperature AT and Overpower AT - The response time value of j[ 4 seconds given in Table 3.3-2 is correct and in accordance with the analysis.
Items 6 and 7 in FSAR Table 7.2-3 includes the 4 seconds listed in the Technical Specifications plus the unmeasurable transport time of 2 seconds associated with the RTD bypass loop flow rate and the thermal lag of the piping system. The RTD bypass loop flow rate verification is required by surveillance requirements in Technical Specification Table 4.3-1.
Additionally, it should be noted that the Technical Specification value is in the more conser-vative direction than the values in the FSAR. Thus, no change is required.
11.
REQUEST Resolve discrepancy between FSAR Table 7.3-1 Item 2a and Technical Specification Table 3.3-3 Item 2a - Containment Spray Manual Initiation.
RESPONSE
FSAR Table 7.3-1 Item 2a is incorrect. The Number of Channels should be 2 and the Number of Channels to Trip should be "I with 2 coincident switches." This is consistent with the Technical Specification table. to this enclosure provides a marked up copy of the affected FSAR page. The footnote in Table 7.3-1 in itself is written correctly.
-FSAR Figure 7.2-1, Sheet 8, shows the functional diagram for Containment Spray. Please refer to this figure for a clarification of the aforementioned footnote.
12.
REQUEST
. Verify that the 1.58 square inch vent is consistent with the FSAR as reviewed by the staff.
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RESPONSE
The Cold Overpressure Mitigation analysis discussed in FSAR Section 5.2.2.11 discusses the fact that overpressure protection at low temperatures is provided by the operation of a single PORV. Calculations performed by YNSD shows that 1.58 square inches is the area of a single PORV. The basis for one PORV is consistent with the FSAR. The analysis supports the FSAR and the FSAR has been reviewed by the staff.
13.- REQUEST
~Need assurance that backup trips (i.e...second trip in accident sequence) are included in the harsh environment analysis - diverse trip functions should include any associated environmental allowances as a part of the setpoint calculations.
RESPONSE
We have learned that the NRC staff has identified this concern with regard to other recent licensees.
We also understand that the staff's position is that this concern is not generic in nature and should be 1
based on the plant specific nature.of the environmental errors that should be included. Westinghouse has previously docketted a response to this concern during the licensing of the SNUPPS plants (Docket Nos.
STN-50-482 and STN-50-483). We would like to bring this response to your
. attention, portions of which we have included'and updated in response to your concern with regard to Seabrook. Westinghouse respectfully i
disagrees with the staff's contention that the matter is plant specific. While it is true that the magnitude of the environmental i
errors (EA) is plant specific (based on the environmental conditions present during and after an event, and the specific vendor and type of transmitters); the concept of including an environmental error in the determination of a trip or actuation setpoint for a protection function noted as a backup or diverse trip is a new and radical interpretation of the GDCs and Regulatory Guides pertaining to protection function operability requirements. The staff has noted that the grounds for the requirement of inclusion of the EA in backup trip setpoints is GDC 22, citing that the failure to include an EA in the backup trip setpoint potentially invalidates compliance with the GDC because it would then be impossible to claim that the backup trip would perform as a diverse function. Westinghouse disagrees with this interpretation of GDC 22 for the following reasons.
I.
GDC 22 states that:
l "The protection system shall be designed to assure that the effects of natural phenomena, and of normal operating, maintenance, testing and postulated accident conditions on redundant channels do not result in loss of the protection function, or shall be demonstrated to be acceptable on some other defined basis. Design techniques, such as functional diversity in component design and principles of operation, shall be used to the extent practical to prevent loss of the protection function."
(underlined for emphasis)
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It is Westinghouse's position that this GDC is met in its entirety by the current Westinghouse design and analysis basis. GDC 21 is satisfied with regards to single failure criteria. CDC 24 is satisfied with regards to control / protection separation, and IEEE 323-1974 is' satisfied with regards to transmitter operability and environmental qualification.
It is therefore reasonable to assume the proper functioning of redundant primary trip functions.
2.
Prudent engineering suggests that backup functions of some nature be provided and in the Westinghouse design they are.
Westinghouse ensures that where necessary and prudent, one or more additional trip functions are available to backup the primary trip functions.
However, because explicit inclusion of backup trips is outside the design and analysis basis, Westinghouse does not determine a value for a trip setpoint based on a backup function to a primary trip.
Reactor Trip and ESF setpoints are determined solely on explicit analysis assumptions and where no explicit assumptions are made, engineering judgment is used to make a determination as to an adequate setpoint value. These assumptions are noted in the plant specific setpoint studies which Westinghouse has provided for NRC staff review since June, 1978, for each plant since D. C. Cook Unit II.
Therefore, to ensure that a backup function will be available, if necessary, Westinghouse has determined that in the lack of specific NRC guidelines and requirements, the requirement of survivability should be used. That is, if an event takes place which generates adverse environmental conditions and a backup trip function is prudent, that function's transmitter should be qualified for the conditions that exist at that time.
If the transmitter will not experience adverse conditions, qualification is not necessary. This position assures that CDC 22 is met because there is no loss of necessary protective action. The action may be delayed but it is reasonable to assume that in any postulated major event, action will be initiated such that 10CFR100 limits will be satisfied.
3.
Regulatory Guide 1.105, Rev. 1 (11/76) notes in the Introduction that 10CFR50.36 is the basis for determination of a setpoint.
10CFR50.36(c)(1)(ii)(A) states"...Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting shall be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded...".
(underlined for emphasis) Westinghouse has conserva-tively assumed for the determination of a trip setpoint that satis-fying the Safety Analysis' Limit.(the assumed value for protection function actuation) assures that the Safety Limit (as defined by 10CFR50.36) is satisfied.
If a protection function is not used in the Safety Analysis for a particular event, then no Safety Analysis Limit exists for that function.
It is then impossible to determine a Limiting Safety System Setting (as defined by 10CFR50.36) except by the use of engineering judgment. As noted previously, Westinghouse uses the most limiting Safety Analysis Limit for a primary protection.
function in determining a trip setpoint.
In the event a protection function is not used in the Safety Analysis, as noted in setpoint study and of which there are several instances, engineering judgment is used to make a reasonable determination.
4.
ISA Standard S67.04, 1982, notes that inclusion of EA is necessary only in certain circumstances. Section 4.3.1(6) states " Environ-mental effects on equipment accuracy or time response characteristics caused by anticipated >perational occurrences or accidents for those systems required to mitigate the consequences of such events."
(underlined for emphasis) The only systems required to function in such conditions are primary trip functions, since those are the only functions assumed to operate. The NEC staff was represented when this standard was written. The staff has endorsed this standard in Regulatory Guide 1.105, Rev. 2.
In summary, Westinghouse believes the following:
1.
Based on readings of the noted GDCs, Regulatory Guides, and industry standards, the interpretation of GDC 22 espoused in the referenced NRC letters is new and generic in scope.
2.
The Westinghouse setpoint methodology used (which has been approved on many previous plant applications) is consistent with the require-ments of GDCs 21, 22 and 24, and Regulatory Guide 1.105, Rev. 2.
3.
With the lack of previous NRC guidance and requirements, Westinghouse has determined that the assurance of survivability of a backup or diverse trip function is adequate to ensure the health and safety of the general public.
In the event this requirement is deemed inadequate, generic level discussion, clarification, and opportunity to comment is in order. As a demonstration of the Westinghouse position on survivability a table was prepared noting the primary and
~
backup trips for the events discussed in the Safety Analysis (Chapter 15). The table was based on Table 15.0-6jof the SNUPPS FSAR but is characteristic of and applicable to plants utilizing the protection system provided to SNUPPS (similar plants lare Byron, Braidwood, Seabrook, Millstone, etc.).
That table rjhowed that, where necessary, backup trip functions are environmentally qualified to a level sufficient to assure survival of the environmental conditions and thus continued ability to serve as a backup function.
14.
REOUEST Page 3/4 3 Functional Units 14 and 15.
Should channels to trip be "l/ bus instead of 2.1/ bus both busses"?
RESPONSE
The terminology for the number of channels to trip is acceptable as either "l/ bus both busses" or 2-1/ bus."
We would prefer to have it remain as "2-1/ bus.~
15.
REQUEST Justify the changes to Action Statement 5 on page 3/4 3-6.
RESPONSE
Our interpretation of the reason for the inclusion of valve numbers in
' Action Statement 5 on page 3/4 3-6 is to prevent an unmonitored boron dilution.in the event no source range nuclear instrumentation is available. We will propose valve isolation to prevent an unmonitored boron dilution. A marked up copy of the affected Technical _ Specification page 3/4 3-6 with the appropriate corrections are being submitted with SBN-1055, dated May 17, 1986.
16.
REQUEST Justify why a monthly surveillance of the Boron Dilution Alarm Setpoint is not performed as part of Action 9 on page 3/4 3-14.
RESPONSE
The Boron Dilution Alarm is provided by the Shutdown Monitor. The Shutdown Monitor is a microprocessor which monitors reactor power from qualified excore neutron detectors, and provides an alarm on neutron flux doubling in the source range. A description of this system is provided in Section 7.6.11 of the FSAR. During power operation the Shutdown Monitor will not normally be in alarm hence Action Statement 9 of page 3/4 3-9 is not appropriate.
17.
REQUEST Justify the omission of P-12 as an ESF interlock.
RESPONSE
The P-12 interlock does not perform an ESF function for Seabrook, it is not used in the mitigation of any FSAR Chapter 15 analysis, and hence has been deleted from the Technical Specification 3/4.3.2.
P-12 is used only for operation of the steam dump valves which are not safety related.
18.
REQUEST Does the Loss-of-Offsite Power initiating signal and function actuate more systems / components than listed in Table 3.3-5?
RESPONSE
On a loss of offsite power, loads are shed from the emergency busses, the diesel generator starts and equipment is loaded back on the busses in accordance with the loading sequence of the emergency power sequencer as described in Section 8.3 of the FSAR. However, the only engineered safety features actuated by the Loss of Offsite Power are the start of the Emergency Feedwater pumps and isolation of the nonsafety portion of the Service Water System. A marked up copy of the affected Technical Specification Table 3.3-5 with the appropriate corrections are being submitted with SBN-1055, dated May 17, 1986.
19.
REQUEST Justify the omission of analog and logic calibrations during the 18 month shutdown reriod for the ECCS.
RESPONSE
We have not deviated from the Standard Technical Specifications with regard to the requirement for channel calibration every refueling.
20.
REQUEST Verify chat all ESFAS slave relays are testable at power. List any that are not.
RESPONSE
All of the ESFAS salve relays are testable at power.
For further discussion of testability refer to FSAR Section 7.3.2.2.e.
21.
REQUEST Justify the omission of a periodic surveillance of the actuation circuitry used to isolate an EFW supply lime on high flow.
RESPONSE
A surveillance for the actuation circuitry used to isolate an EFW supply line on high flow will be proposed for specification 4.7.1.2 A marked up copy of the affected Technical Specification page with the appropriate corrections are being submitted with SBN-1055, dated May 17, 1986.
22.
REQUEST Verify that the secondary steam supply pressure in 4.7.1.2(a).2 is correct.
RESPONSE
The secondary steam supply pressure proposed in specification 4.7.1.2(a).2 has been verified to be correct.
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-23. - REQUEST Not all ESFAS described in the FSAR appear to be in Technical Specifica-tion Table 4.3-2.
Validate agreement.
RESPONSE
All of the functions which are performed by the Engineered Safety Feature Actuation System as listed in Section 7.3 of the FSAR are included in the testing required by Table 4.3-2 of the Technical Specification. The P-12 interlock is included in FSAR Section 7.3 because it is included in the solid state protection cabinet;-however, this interlock is not an q
Engineered Safety Feature for Seabrcok.
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3.1.5 Reactor containment 3.1.5.1 Criterion 50 - Containment Design Basis The reactor containment structure, including access openings, penetrations, and the containment heat removal system shall be designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and, with sufficient margin, the calculated pressure and temperature conditions resulting from any loss-of-coolant accident. This margin shall reflect consideration of (1) the effects of potential energy sources which have not been included in the determination of the peak conditions, such as energy in steam generators and energy from metal-water and other chemical reactions that may result from degradation but not total failure of emergency core cooling l
functioning, (2) the limited experience and experimental data available for w
defining accident phenomena and containment responses, and (3) the conservatism of the calculational model and input parameters.
RESPONSE
The design of the containment is based on two containment design basis accidents. One assumes a double-ended rupture of the largest reactor coolant pipe (LOCA); the other the rupture of a main steam line inside containment.
The maximum calculated atmospheric pressure and temperature reached within _g4 sig and % the maximum the containment during the LOCA are @ ined during a 6ain steam Itne
~2/7;1 7sf atmospheric pressure and temperature atta rupture is 34.5 pais and 3700F. A containment design pressure of 52.0 psig has been selected to provide ample margin to allow for increased energy The peak liner design temperature was selected equal to the sources.
maximum calculated LOCA atmospheric temperature, 271 F.
Although the con-tainment atmospheric temperature following a MSLB is higher than that following a LOCA, the containment liner temperature will not exceed 271 F, since a lower heat transfer coefficient will result under the superheated j
atmospheric condition during the MSLB.
See Subsection 3.8.1 for containment loading combinations and 6.2.1 for design evaluation.
The containment electrical penetrations are designed so that the containment l
structure can, without exceeding the design leakage rate, accommodate the calculated pressure, temperature and other environmental conditions result-ing from any loss of coolant accident. Section Section 8.3 for discussion of containment electrical penetrations and protection of containment elec-trical penetrations.
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SB 1 & 2 Amendment 56 FSAR November 1985 n HachmeJ I 5.
Abnormal Loads Pcuf e 2 e4 11 Abnormal loads are those generated by postulated high energy ruptures, particularly a rupture in the reactor coolant system resulting in a loss of coolant accident (LOCA). Post LOCA contairusent flooding is also considered. The maximum level of flooding, however, is 5'-4" above the top of the fill mat (-26 feet); this depth of water causes negligible loading on the containment structure.
(a) Accident Pressure (Pa)
A transient pressure load is used for the design of the containment. The maximum calculated internal pressure gf,6 associated with the DBA is W JTpsig. le r r:f:::: 21:5 th; ::quir r t: ef IM"50, C -^rrl Seri,2 Criterirr 50, thic - rr ice erred by 2 2ppr prict: r:r;;in, 10!, te detr--Err the ritir-- ::; ir:d d::i;;; pr;;;;.;;, 5 0. af f.8 /, fo
- .paa6-- @ Provide $ additional $marginfthe design pressure b4evhecA is r : ::t:51ich:d 2: 52.0 psig. The pressure-transient curve for the containment is shown on Figure 3.8-9, Sheet 1.
(b) Accident Temperature (Ta)
The transient temperature increase of the liner was considered in the design of the containment. The maximum liner temperature (liner temperature spike) is 268.80F. However, a maximum liner temperature of 2710F has been used in the design. The temperature transient curves for the containment liner are shown on Figure 3.8-9, Sheets 2 and 3.
w The time-dependent thermal gradient through the concrete of the dome, cylinder and base mat was also considered in the design of the containment. When the accident pressure, Pa, is considered, the coincident thermal gradient is equivalent to the normal operating gradient.
Due to the higi...tsulating properties of the concrete, the pressure peak occurs before the temperatures within the concrete are appreciably altered. For design of the cylinder and dome the peak liner temperature and peak pressure were also considered to occur simultaneously.
This produced the most conservative design condition where responses to the loads are additive. Where responses are not additive, peak pressure was considered without the thermal loads. The thermal gradients used l
b for the design of the containment are shown on Figure 3.8-8.
For the design of the liner, the transient conditions of liner temperature and coincident accident pressure were 3.8-21 l
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SB1&2 Amendment 56 FSAR November 1985 6.2 CONTAINHENT SYSTEMS 6.2.1 Containment Functional Design 6.2.1.1 Containment Structure 1
a.
Design Bases The containment design bases are established by the requirement that the system safely withstand the consequences of postulated accidents in conjunction with simultaneous occurrences of adverse environmental conditions. The containment structure and the containment enclosure, together with the exhaust systeni, are designed such that the offsite doses from radioactivity released under accident conditions are less than the limits set forth in 10 CFR 100.
1.
Postulated Accident Conditions for Containment Design Accidents postulated to ' determine the containment internal design pressure and the containment design temperature include ruptures of the primary and secondary coolant system piping concurrent with a variety of single active failures. The simultaneous loss of of fsite power (LOOP) has also been assumed whenevet it results in more restrictive design conditions.
The detailed accident conditions for primary system pipe rupture are given in Subsection 6.2.1.3; those for secondary system pipe ruptures in Subsection 6.2.1.4.
The single active failures (SAFs) postulated for the primary system pipe ruptures include failure of a containment spray train and failure of a diesel generator. Those postulated for the main steam line breaks include failure of main feed-water pump to trip, a feedwater isolation or control valve, a l
main steam isolation valve, an emergency feedwater pump s
runout control, and a containment spray train.
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The calculated maximum internal containment pressure is 46=t psig, resulting from a (full) double-ended guillotine rupture of the primary coolant system pipe at the pump suction, with one of the two containment spray pumps failed at time of 2
containment a ray actuation, nominal initial containment y -
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pressure of
/ psig and design flow rate of g" fore pm tor the containment building spray system. This is, there the containment design basis (DB) accident. t :::::d:::: riS g[5gr r'=a=-=1
" r - i t e r i : r "O c f 10 C:'"
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i mH t: 52.0 psig, thus providing a M _ margin between
% ~ 4.J h the design and_ maximum calculated values./ ine siimi realculation f contai e'n t building ay system in icate the' f reduct' of the i
_ction flow r to 2930 gpm r the DB i
LO This red ion resulted
'n a slight in ease of th alculated m imum interna.ontainment pre sure to 4 psig.
Us of the maxim tech spic inji'al conta' ent pressu of 1.5 psig the analysip fesults i he Eurther inc se of maximt calculated cretainment essure to 47.7 psig. Henc the margin bItween the design and maximum f
(calculatedvalues is still 9.0%,
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i break type / size and SAF combination, namely Break No. 1 and SAF No. 2.
Overall, a total of six cases have been analyzed. The details of the calculation of the mass and energy releases for the six cases analyzed are given in Subsection 6.2.1.3.
The results obtained therein have been based on a temperature of 1200F for the safety injection water.
For Seabrook Station, Units 1 and 2, the maximum temperature of the injection water is-888 ~.
F
/00 *f The time for switchover from the injection mode to the recirculation mode depends upon the injection flow (charging pumps, high pressure safety injection pumps, and low pressure safety injection pumps),
spray flow and the quantity of water available in the RWST. Since one of the two spray trains has been assumed to fail in all cases analyzed, the recirculation times are calculated for the maximum safety injection (two' injection trains) and the minimum safety injection (one injection train) cases. With the injection and spray flow rates given in Table 6.2-2 and the available quantity of water in the RWST provided in Table 6.2-1, the recirculation times for the maximum safety injection and the minimum safety injection cases are calculated to be 1662 seconds and 2685 seconds, respectively.
The transient responses of the containment pressure, teoperature and sump water temperature for the six cases analyzed are shown in Figures 6.2-1 through 6.2-18.
The transients show that, following blowdown, and prior to refill, the containment pressure and containment temperature drop because the mass and energy released through the breaks ceases completely at the end of the blowdown period.
l However, the reflood and post-reflood mass and energy released from the break increase the containment pressure and temperature again. The l
containment pressure and temperature eventually drop due to decreases in the mass and energy release rates, and due to energy removal by containment spray and passive heat sinks. After the switchover i
from the injection mode to the recirculation mode, l
the containment spray water is taken from the containment sump through the containment spray heat exchanger, and is at a higher temperature than that of the RWST. The spray heat removal rate thus drops.
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SB 1 & 2 Amendment 56 a-FSAR November 1985
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water tank temperature of 500F, and concurrently provide. sufficient he.t into the enclosure area to maintain an enclosure temperature of 390F. No credit was taken for heat contributions from the SAT heaters. The site environmental condition for this design' evaluation assumed -170F and 30 mph winds, and are more conservative than the sinimum outdoor conditions listed I
in FSAR Figure 3.11(5)-1.
For the above envircamental conditions, the heat loss from the enclosure building, including infiltration losses, is 158,000 BTU /hr. as compared to an KWST heating' panel capacity of 674,000 BTU /hr. Accordingly, freeze pro-taction is provided for all equipment. In addition, both tank low temperature and enclosure temperature alarma are provided in the main control room.
For Unit 2, the building geometry is different, and a separate analysis is being performed to verify freeze protection.
98,F.
- ~
The animum tempereture of the water in the ' tanks is :21:21 ; 2 R L O',^7, r8; ;1--- 1;::ifi: 2:t::::1:;i:21 d te, :::::in;
- f--
- 1;; i::: ;:in frf f 21_:: f the ;::til tier fr-. The heat of reaccios upon the mixing of thd bo'ic acid and the sodium hydroxide would raise the temperature of r
the tank contents a proximately 20F thereby raising the RWST maximum supply of temperature to Neither tank is protected against tornado missiles, l00i;,and a tornado and accident;are noe considered simultaneous events.
In the event of tornado damage to either tank, the affected unit would be shut down.
The SAT is, connected to the RWST'by two parallel lines each with an automatic motor operated valve. The valves are actuated and powered from separate sources to' insure that the NaOH solution can be added to the containment spray even in the event of a single active failure.
The method of addition of 20% NaOH solution in required concentrations to the borated water drawn from the RW3T immediately following a LOCA is primarily dependent on passive components, such as tanks, pipes and a baffled mixing chamber. The rate of addition is' dependent on the drawdown rate of the RWST and is based on principles of hydrostatics 3nd hydrodynamics. A description of tho system is contained in the following paragraphs.
i
^
L r
4 t
6.2-48a
g AHachmed /
0 N
SB 1 & 2 f
FSAR f
single failure does not prevent the transfer to the recirculation mode since each active component is duplicated by the dual train concept.
c.
Spray Effectiveness Each spray train contains 198 SPRAYCO 1713A hollow cone ramp bottom nozzles. Sixty-five (65) nozzles were randomly selected from a quantity of 325 to evaluate the performance of the nozzles and verify the required flow of 15.2 gpa at 40 psi differential pressure.
The average mean droplet diameter for the nozzles tested was 660 microns. This compares to a conservative value of 1250 microns used in the containment iodine removal analysis. The average mean drop diameter was arrived at by numerical averaging based on an instantaneous sampling of spray at design conditions.
Table 6.2-80 lists the percentage of sprayed volume and Figures 6.2-83 and 6.2-84 show the extent of overlapping of the sprays in plan for spray loops A and B, respectively. 'lhese figures show virtually 100% coverage of the containment at the operating floor level. Figures 6.2-85 and 6.2-86 show the spray loops A and B coverage pattern in elevation views.
Figures 6.2-87 through 6.2-89 sununarize the operating characteristics of the spray nozzles at 40 psi.
It should be noted that 99.99 percent of the drops are below a diameter of 1500 microns, have a terminal velocity less than 17.88 fps and contain over 99.99 percent of the total liquid volume.
It is assumed that, following a LOCA, the sprays are initiated at a time when the containment atmospheric temperature is 2660F, such that the air steam ratio is 0.
nds air per pound steam,and the initial spray temperature is
/dO F, The drop is considered to be a rigid sphere of radius to int ally at To in an air / steam atmosphere at Too, and Too does not change during the time the drop is falling.
Steam in contact with the drop will condense, leaving a boundary layer of air around the drop. This is equivalent to having an extremely large air-steam mass ratio, essentially 100 percent air and 0 percent steam.
It is also conservatively assumed that this boundary layer is created instantaneously. Heat is transferred to the drop through the boundary layer by convection, and water vapor dif fuses through the boundary layer.
Ranz and Marshall (Reference 19) provide correlations for both mass and heat transfer. After diffusing through the boundary layer, the steam will condense on the drop surface and the latent heat of condensation will act as a surface heat source.
The assumption of a rigid drop implies the longest time for the drop to heat up, thus providing the most conservative case.
6.2-51
AH:ehmed I tage M tt SB 1 & 2 Amendment 53 FSAR August 1984
(
Mn average molecular weight of boundary layer
=
diffusivity of water vapor D
=
heat transfer coefficient h
=
thermal conductivity k
=
Reynolds, Prandt1 and Schmidt numbers Re,Pr,Sc
=
Equation (1) describes the temperature behavior of the drop; equation (2) describes the mass transfer coefficient; and equation (3) the convective heat transfer coefficient at the drop boundary.
Paraly (Reference 20) has solved this set of equations numerically using a finite difference method for a range of drop sizes from 500 to 4000 microns in diameter, containment temperatures of 212oF and 2660F, and initial drop temperatures of 86oF,1220F and 176oF.
o Using these results for an avera ray nozzle height of 134 feet _g g and initial drop temperature o F equilibriation time is much shorter than the time required for the drops of essentially all sizes to reach the containment sump. Table 6.2-81 presents the parametric results.
It can be concluded from Table 6.2-81 that even the largest spray drops attain the containment temperature at times far shorter than the time required to reach the containment sump. As a result, the spray effectiveness value of 1.0 is fully justified in the case of a LOCA. The effectiveness of the sprays following an MSLB, when the containment atmosphere is superheated, is discussed in Appendix ISB.
d.
Net Positive Suction Head Available Adequate net positive suction head (NPSH) for the containment spray pumps is assured under all postulated operating conditions by analysis of the suction head available and vendor testing of the completed pumps.
Figure 6.2-81 shows NPSH available versus NPSH required over the range of flow. Maximum calculated flow under the most limiting NPSH conditions, i.e. during recirculation, is 3260 gpm.
NPSH available at this flow is 22.31 feet versus a maximum required NPSH of 20.5 feet. The analysis of available NPSH conservatively assumes that each residual heat removal pump which shares a common g
suction on a train basis with each CBS pump is also operating at its maximum design runout flow of 4500 gpm and considers the flow path with the highest hydraulic resistance. The formulas and flow resistance data in Reference (21) were used along with the test data for the bell-mouth sump suction piping, to compute NPSH available.
s 6.2-53
-s 4 hchent I th e s,c n f
SB 1 & 2 Amendment 58 FSAR April 1986 The testing program will include a complete series of Type B and Type C tests as presented in Subsections 6.2.6.2 and 6.2.6.3, prior to fuel loading. During plant operation, each penetration requiring testing will be tested at least once every two years to ensure continued compliance with leakage limits.
If lM maintenance must be performed on a penetration because of suspected excessive leakage, it will be done upon completion of an initial leakage rate measure-ment. A second leakage measurement will be made upon completion of this maintenance to determine the "as left condition".
47 6.2.6.1 Containment Integrated Laakage Rate Test - Type A Test The initial containment integrated leakage rate test (Type A test) will,be performed af ter completion of construction of the containment structure and prior to initial fuel loading. The maximum allowable integrated leakage rate, La, at the calculated peak accident pressure, Pa, is 0.1 weight percent ll per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The calculated peak accident pressure, Pa, i psig. To 53 f4 allow some margin for deterioration between Type A tests, the acceptance N W,(,
criteria for the Type A test conducted at the calculated peak accident pres-sure (Pa) is that the upper 95% confidence limit of the measured leakage rate, Ian, be less than 0.75 La.
All penetrations will be installed and all systems penetrating the contain-ment will be complete, up to and including all automatic isolation valves external to the containment prior to the conduct of the initial Type A test.
Deviations from this schedule will be documented and properly consideren when reporting final leakage rate test results.
If the structural integrity test (SIT) precedes the initial Type A test, a
minimum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> will elapse from the time the containment was in excess of 85% Pa for the SIT and the commencement of a Type A test, to assure sufficient time for outgassing from the internal structure.
Prior to the conduct of a Type A test, a general inspection of the accessible interior and exterior surfaces of the containment structure and component's will be performed to uncover any evidence which may affect either the struc-tural integrity or leak-tightness.
If evidence of structural deterioration is noted, the Type A test will not be performed until corrective action has been completed in accordance with applicable repair procedures and tests specified in the applicable codes specified in 10CFR50.55a.
Such structural deterioration and corrective actions will be reported as part of the sub-sequent required test report.
In order that test results from the initial test be comparable to results obtained from future tests, all systems within the containment will be, as cLsely as is practical, in the condition they would normally be during future tests. Where possible, those systems which would be exposed to the contain-ment atmosphere during post-accident conditions will be drained and vented to containment atmosphere within the containment on the inboard side of isolation valves, and drained and vented to atmosphere outside the containment on the outboard side of the outboard isolation valves.
Systems that are required 6.2-82
AHackened 1 Paye 9ofll SB 1 & 2 Amendment 58 FSAR April 1986 to maintain the plant in a safe condition during the test will be operable 4
in their normal mode, and will not be vented.
Systems that.are normally filled with water and operating under post-accident conditions, such as the containment heat removal system, will not be drained and vented. Valves will be closed by normal operation without any adjustments or exercising for the purpose of improving leakage performance. Table 6.2-90 lists systems vented prior to and during the conduct of the Type A test. Table 6.2-91 56 lists those systems not scheduled to be vented and the justification thereof.'1 Those systems which are not vented and drained, and which could become exposed to containment atmosphere under postulated accident conditions, will be evaluated separately and their leakage contribution added to the measured Type A test results.
To perform the test, the containment will be pressurized with reasonably clean air which will be run through a cooler prior to admission to the containment, to assure that atmospheric stability will be obtained prior to the start of the test. Internal fan coolers will assure air mixing and l
temperature control during the conduct of the test. When the temperature is 56 stable, the pressure and humidity will also be stable. The temperature will be considered stable when the rate of change of the weighted average contained air temperature averaged over the last hour does not deviate by more than 0.50F/hr from the average rate of change of the weighted average contained air temperature averaged over the last four hours.
The containment leakage rate will be determined by the absolute method, i
utilizing a series of dry bulb temperatures, relative humidity measurements, (dew point temperatures) and the containment absolute pressure. Based upon the Perfect Gas Law, calculations of the contained mass of air with respect to time will be made. The test data vill be plotted and a least-squares fit of the data will be performed. The slope of this least-square line will be calculated to determine the leakage rate. A statistical analysis will deter-eine the acceptability of the measured results. The instrumentation sensi-tivity will be verified by a supplemental test upon successful completion of a Type A test. A controlled, metered leak will be established out of containment. This superimposed leak rate will be added to the existing containment leak rate and the composite leakage used to verify satisfactory response of the Type "A" instrumentation.
S Type A test acceptance criteria require that the upper 95% confiden limit 9 9 G, (UCL) of the measured test result, conducted at a pressure of Pa (/
psig),
be less than 0.75 La (0.11%/24 hrs). For Type B and C tests, the acceptance l
criteria require that the combined total B and C leakage, including the upper 55 error limit, be less than 0.60 La (0.09%/24 hrs). These acceptance criteria j
are in accordance with Appendix J.
S5 SS h
t 6.2-83 4
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Vs kh.ekm ent I P a rg e t o a 4 I I SB 1 & 2 Amendment 54 FSAR February 1985 TABLE 6.2-75 (Sheet 1 of 3)
CottfAlletENT SPRAY SYSTEM PARAMETERS Spray Additive Tank Quantity 1
Type Vertical cylinder Volume 10,700 gal Material Austenitic stainless steel Design code ASME Section III, Class 3 ANSI N18.2 safety class Class 3 Operating pressure Atmospheric Design temperature 1000F Maximum fluid temperature M C/$ Y 54 NaOH concentration 20% by weight Containment Spray Pumps Quantity 2
Type Centrifugal Horsepower 600 Hp Design flow 3010 gpm Operating flow 3010 gpm NPSH required 21 ft 0
Design temperature 300 F l
Design pressure 300 psig Design code ASME III, Class 2 i
ANSI N18.2 safety class Safety Class 2 t
i
v o N*W l Page it at it SE 1 & 2 Amendment 54 FSAR February 1985 TABLE 6.2-75 (Sheet 2 of 3)
Refueling Water Storage Tank Quantity 1
Capacity 475,000 gal Material Austenitic stainless steel Type Vertical cylinder Design code ASME III, Class 2 ANSI N18.2 safety class Class 2 Concentration of boron 1900 ppe boron (nominal)
Design temperature 1000F Maximum fluid temperature N$* 98 f +
54 Operating pressure Atmospheric Containment Spray Heat Exchanger Quantity 2
Type Shell and tube Design codes:
Shell side ASME III, Class 3 Tube side ASME III, Class 2 i
ANSI N18.2 safety class:
l Shell side 3
l Tube side 2
f Material:
I Shell side Carbon steel Tube side Austenitic stainless steel Design pressure:
Shell side 150 psig Tube side 300 psig Design temperature:
Shell side 2000F Tube side 3000F
- Maximum fluid temperature within the mixing chamber during the injection phase is(j5y[)due to the exothermal reaction between sodium hydroxide and boric acid. / 00'F' 54
r a
TABLE 7.2-3 (Sheet 1 of 2)
REACTOR TRIP SYSTEM INSTRUMENTATION Typical Trip Typical Time Reactor Trip Signal Typical Range Accuracy Response (sec) 1.
Power range high neutron 1 to 120 percent full power
+4.9 percent of span 0.5 l*
flux 2.
Intermediate range high 8 decades of neutron flux
+5 percent of span N/A neutron flux overlapping source range by
+1 percent of span 2 decades from 10-4 to 50 percent full power (1) 3.
Source range high neutron 6 decades of neutron flux
~+11.5 percent of span (1)
N/A l*
flux (1 to 106 counts /sec) g l 5.-.
g,,
4.
Power range high positive
' +15 percent'of full power
+1.4 percent of span (1)
N/A neutron flux rate y
5.
Power range high negative
-15 percent of fu11 power
+1.4 percent of span (1)
EEEC.
l M
neutron flux rate 6.
Overtemperature T:
Tn 530 to 6500F 6.1 percent of span 6.0(including l 510 to 6300F transport time %
TC 530 to 6300F TAV PPRZR 1700 to 2500 psig
(
F( Ap) -50 to +50 AT Setpoint 0 to 1000F 4
6.0(including l*h$s Eig 7.
Overpower T Tn 530 to 6500F 4.3 percent of syn transport time)*
TC 510 to 6300F gfg 530 to 6300F TAV AT Setpoint 0 to 1000F
_u g i$ f r (1) Reproducibility (see definitions in Section 7.1)
TABLE 7.2-3' (Sheet 2 of 2) 1 l
Typical Trip Typical Time Reactor Trip Signal Typical Range Accuracy Response (sec) 8.
Pressurizer low pressure 1700 to 2500 psig
+2.5 percent of span 2.0 l
Tcompensated signal)
+2.5 percent of span 2.0 l
9.
Pressurizer high pressure 1700 to 2500 psig Tnon-compensated signal) 10.
Pressurizer high water Entire cylindrical portion
+3.4 percent of span N/A l*
1evel of pressurizer (distance between taps) 11.
Low reactor coolant flow 0 to 120 percent rated flow
+2.6 percent of span 1.0 l*
within range of 70 percent to 100 percent
, 88 of full flow (1) e.
12.
Reactor coolant pump bus 61 to 87 percent nominal bus
+10.6 percent of span 1.5 l
N undervoltage voltage 13.
Reactor coolant pump 44 to 61 Hz
+1.0 percent of span 0.6 underfrequency l
14.
Low-low steam generator
+~6ft. from nominal full
+16.4 percent of span 2.0 water level load water level 56 15.
Turbine Trip 150 to 3000 psig
+1.5 percent of span
- I=IE/VM c
E;! E (1) Reproducibility (see definitions in Section 7.1)
NOe R3
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~s 0'b N Np
~
6% A 3 he / <f I SB 1 & 2 Amendment 56 FSAR November 1985 TABLE 7.3-1 INSTRUMENTATION OPERATING CONDITION FOR ENGINEERED SAFETY FEATURES No. of No. of Channels No.
Functional Unit Channels To Trip I'
1.
Safety Injection a.
Manual 2
1 b.
High Containment Pressure (Hi-1) 3 2
c.
Low Steamline Pressure (Lead-Lag 2 in any one compensated)*
12 (3/ steam line) steam line lsu d.
Pressurizer Low Pressure
- 4 2
2.
Manual **
& 2-e / all/h
(
s1. tanc be.$
b.
High-High Containment Swi1[Ap Pressure (Hi-3) 4 2
p Permissible bypass if reactor coolant pressure less than P-II.
Manual actuation of a train of containment spray is accomplished by actuating either of two sets (two switches per set). Both switches in a set must be actuated to obtain a manually initiated spray signal for the respective train. The sets are wired to meet separation and single failure requirements of IEEE Standard 279-1971. Simultaneous operation 5L of two switches is desirable to prevent inadvertant spray actuation.