ST-HL-AE-1674, Forwards Comments & marked-up SER (NUREG-0781) Re Operation of South Texas Projects 1 & 2
| ML20211D828 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 06/09/1986 |
| From: | Wisenburg M HOUSTON LIGHTING & POWER CO. |
| To: | Noonan V Office of Nuclear Reactor Regulation |
| References | |
| CON-#286-562, RTR-NUREG-0781, RTR-NUREG-781 OL, ST-HL-AE-1674, NUDOCS 8606130161 | |
| Download: ML20211D828 (79) | |
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June 9, 1986 ST HL-AE-1674 l
File No.: G7.3 Mr. Vincent S. Noonan, Project Director PWR Project Directorate #5 U. S. Nuclear Regulatory Commission Washington, DC 20555 South Texas Project Units 1 and 2 Docket Nos. STN 50-498, STN 50-499 Comsg a on the Safety Evaluation Report _(SER), NUREG-0781 bear Mr. Noonan:
The South Tc.xas Project (STP) has completed its review of the "Ssfety Evaluation Report related to the operation of South Texas Project, I? nits 1
& 2, NUREG-0781". Attached please find our general comments regarding the subject document, t
In Attachment I we have included written comments on the SER while in Attachment 2, we have provided miscellaneous editorial commenta on marked-up pages of the SER as; well as marked up pages as referenced by comments described in Attachmont 1.
If you should have any questions on this matter, please contact Mr.
M. E. Powell at (713) 993 1328.
Very trulff'yours, r
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M. R. His nburg
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Manager,1 telear Lice n i
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MEP/bl Attachments 1: Written Comrtents on the STP SER, NUREG-078L 2: Mark-up of SER pages I G606130161 860609 \
PDR ADOCK 00000498 j Ll\NRC\aba i1 .
_ , .. -_ .. - - n lionston Lightirig & Power Company ST-HL AE-1674 Pile No.: 07.3 Page 2 cc:
Hugh L. Thompson, Jr. , Director Brian E. Betwick, Esquire
- Division of PWR Lic&asing - A Assistaht Attorney General for Office of Nucitar Reactor Reg:11ation the State of Texas U.S. Nuclear Regulatory Commission P.O. Sox 12548, Capitol 3rstion Washington, DC 20555 Austin, TX 78711 Robert D. Itartin Lenny A. Sinkin
. Regional Adminiatl'ator, Region IV Christic Institute Nuclear Regulatory Cosmission 1324 horth Capitol Street 611 Rya.n Plaza Drive, Suite 1000 Washington, D.C. 20002 Arlington, TX 76011 Oreste R, Pirfo, Esquire N. Pragad Kadarbi, Project Manager Hearing Attorney U.S. Nuclear Regulatory Commission Office of the Executive legal Director 7920 Norfolk Avenue U.S. Nuclear Regulatory Commissi6n Bethesda, MD 20814 Washington, DC 20555 Claude E. Johnson Charles Bechhoefer, Esquire Senior Resident Inspector /STP Chairman, Atomic Safety &
c/o U.S. Nuclear Rc6ulatory Licensing Board Ccmmissine. U.S. Nuclear Regulatory Ccmmission P.O. Box 910 Washington, DC 20555 Bay City, TX 77414 Dr. James C. Lhmb, III M.D. Schwarz, Jr. , Esquire 313 Voodhavah Poad Baker & Botta Chapel Hill, NC 27514 One Shell Plaza Houston, TX 77002 Judge Frederick J. Shen Atomic Safety and Licensing Board J.R. Nowman, Engyire U.S. Nuclear Regulatory Commission Newman & iloltzinger, P.C. Washington, DC 20555 1615 L Street, N.W.
Washington, DC 20035 Citizons for Equitable Utilities, Inc.
c/o Ms. Peggy Buchorn Director, Office of Inspection Route 1, Box 1684 and Enforcement Brazoria, TX 77422 U.S. Nuclear Regulatory Commission Washington, DC 20555 Docketlug & Service Section Office of the Secretary T.V. Shockley/R.L. Rai.ge U.S. Nuclear Regulatory Commission Central Power & Light Company Washington, DC 20555 P.O. Box 2121 (3 Copies)
Corpus Christi, TX 78403 Advisory Committee on Reactor Safeguards li.L. Peterson/C Pokorny U.S. Nuclear Regulatory Commission City of Austin 1717 H Street P.O. Box 1088 Washington, DC 20555 Austin, TX 78767 J.B. Poston/A. vonRcsenberg City Public Service Board P.O. Box 1771 San Antonio, TX 78296 Ll/NRC/aba Revised 5/22/86
h Attrchment 1 ST-HL-AE-1674 File No.: G7.3 Page 1 of 18 South Texas Project Units 1 & 2 Docket Nos. STN 50-498, STN 50-499 Comments on the STP Safety Evaluation Report, NUREG 0781 Chapter 1 o Page 1-4/Section 1.2.2 (Para, on RHR) - Note that RHR cut-in is dependent on both temperature and pressure, not just temperature.
o Pg. 1-5& 1-6/Section 1.2.2 - The RHR Ex removes heat from the low head SI lines during the recirculation mode only.
o Pg.1-6/Section 1.2.2 - The AFWS consists of 4 pumps - 3 motor driven and 1 turbine driven. The existing description implies that STP only has 2 AFU pumps.
o Pg. 1-9/Sec'fon t 1.2.2 - The Emergency Operations Facility _(EOF) per NUREG-0737 is referred to as the Emergency Operations Center (EOC) at STP, This same comment applies throughout the SER. .
o Pg. 1-9/Section 1.2.2 - The volume of the diesel fuel oil storage tanks is 67,000 g.allons, o Pg, 1-9/Soction 1.2.1.2 para . 1, 2nd sentence - Should read: "
...from tte auxiliary fuel oil storage tank."
o Pg. 1-9/Section 1.2.1.2 para. 2, 2nd to last sentence - Should read:
...until the EOC is activated."
o Pg. 1-9/Section 1.2.1.2 para. 3, last sentence - Should read:
"...approximately 1/4 mile east..."
o Section 1.3.2/Pg, 1 This section implies that the RHR pumps are used for SI, This is not true for STP. The only portion of the RHRS that is used during SI is the RHR Hx which is used to' remove heat from the LHSI system during SI, o Section 1.3.2/ Pg. 1-10 The containment analysis considers the operation of 3 KCFC units, which is 1.5 trains not 3 trains, and 2 of 3 CSS traina operating, o Pg 1 11/Section 1.3.5 - The QDPS also performs RCS hot leg temperature averagit)g, This should be added.
o Pg 1-17/ Table 1.1 - A response to TMI Action Plan Itaas II.K.1,5 and II.K.1.10 was provided to the NP.C via HL&P letter ST-HL AE.1636 dated 4/17/86, o Pg 1-18/Tablo 1.1 - The correct SER Section reference for TMI Item II.K.3.12 is 7.2.2.4 L1/NRC/a
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1 A Att:chment 1 ST-HL-AE-1674 File No.: G7.3 Page 2 of 18 o Pg 1-19/ Table 1.2 implies that STP has 9 SI pumps. This is not true.
STP has 3 HHSI and 3 LHSI pumps. The charging pumps and RHR pumps are l I
not used for SI as in other y designs. In other y designs, during SI, the charging pumps, RHR pumps and SI pumps draw suction from the RWST J during injection. During Recirculation, the RRR pumps and SI pumps draw from the containment sumps while the charging pump draws suction from the RHR pump discharge. We suggest the following changes:
STP g SNUPPS
- of IRSI pumps 3* 0 0
- of HHSI pumps 3* 0 0
- of Charging pumps 2 2* 2*
- of RHR pumps 3 2* 2*
- of Intermediate SI pumps 0 2* 2*
- - used during SI r
j An alternative change to the Table 1.2 could be the following:
! STP CP SNUPPS High Pressure Pumps 2"I 2 2 Intermediate Pressure Pumps 3 Low Pressure pumps 3"3 23 "I 22 "3 l This will allow keeping the table intact and just deleting the SI designators.
It should be noted that the classifications of the subject pumps are based on shutoff heads. The footnotes being added are,
'"3 "3
Charging only i
Dedicated LHSI pumps; 3 separate dedicated RHR pumps also Low head SI/RHR function shared o Table 1.4 - For open item #11, the correct SER reference is Section 9.5.1.7
! o Table 1.5 - For confirmatory item #30, the correct SER reference is i
Section 17.5.4 4
Chapter 2 i
o Pg 2-1/Section 2.1.1 - The value listed for the site acreage and the distance of the site from Matagorda is not consistent with SER Section 1.2.1.1 and the STP FSAR.
o Pg. 2-1/Section 2.1.1 - The coordinates of Unit 1 given in the SER are incorrect. The correct coordinates of Unit 1 are 28', 47', 42" north latitude and 96*, 02', 53" west longitude (reference STP FSAR Section 2.1.1).
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9 AttrchmInt 1 ST-HL-AE-1674 l File No.: G7.3 Page 3 of 18 o Pg. 2-1/Section 2.1.1 (2nd paragraph) - The site in SER section 1.2.1.1 should be 8.0 miles northwest of Matagorda instead of 8.5 miles.
o Pg. 2-1/Section 2.1.2 - The last sentence of the first paragraph should read: "No people reside within the exclusion area. The applicant has acquired all of the surface estate within the site boundary as well as most of the mineral interests within the site boundary. As the result of the acquisition of this surface estate and these mineral interests the applicant has the authority required by 10CFR100 to determine all activities within the exclusion area."
o Pg. 2-2/Section 2.1.3 - HL&P expects the population within 50 miles of the plant to increase to 299,443 by 1990 and 528,081 by 2030 (reference ER-OL Figures 2.2-2 and 2.2-6).
o Pg. 2-2/Section 2.1.3 - The inconsistencies in the population distribution have been rectified by Amendment 9 to the ER and Amendment 53 to the FSAR. The information in these two documents is also consistent with the Emergency Plan.
o Pg. 2-3/Section 2.1.4 - The subject inconsistencies have been rectified per the previous comment.
o Pg. 2-3/Section 2.2.1 - Delete reference to "the picnic area". The applicant has no plans for a picnic area at the site, o Pg. 2-3/Section 2.2.1 - 3057, 2668, and 1095 should be referred to as FM 3057, FM 2668 and FM 1095. In addition, the EZB is 1430 m not 1432 m.
o Pg. 2-7/Section 2.3.1 - The # of thunderstorm - days per year is 50 which is based on 46 at Victoria, TX. which is located approximately 65 miles from the site (ref. FSAR Section 2.3.1.2.5) o Pg. 2-10/Section 2.3.3 - The first paragraph addresses the old ,
meteorological tower and should be rewritten in past tense to avoid confusion with the upgraded meteorological system.
o Pg. 2-10/Section 2.3.3 - The height of the meteorology tower is 197 feet (60m) not 195 feet.
o Pg. 2-11/Section 2.3.3 - The first sentence of paragraph one should be rewritten as follows: "The meteorological measurements system was calibrated quarterly during the period of record from July 21, 1973 to January 3, 1978."
o Pg. 2-14/Section 2.4.1.1 - The sixth paragraph on the page refers to "301 miles upstream". The staff should indicate if this measurement is in river miles.
o Pg. 2-14/Section 2.4.1.1 - FSAR sect identifies the ColoradoRiverbasjntobe41,800mi}on2.4.1.2.1.1.1, (of which 90 mi is downstream of STP) not 40,800 mi as stated in the SER, t
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Attcchment 1 ST-HL-AE-1674 File No.: G7.3 Page 4 of 18 o Pg. 2-14/Section 2.4.1.1 - FSAR section 2.4.1.2.1.1.1 identifies an uncontrolled drainage arpa between the Mansfitld Dam and the South Texas Project of abou 3700 mi . The SER indicates,_an~ uncontrolled drainage area of 3580 mi o Pg. 2-15/Section 2.4.1.1 - The embankment can impound 202,600 acre-feet of cooling water for _ normal plant operation as authorized in Texas Water Commission Water Supply Permit No. 3233.
o Pg. 2-15/Section 2.4.1.1 - The ECP has a maximum surface area of 47 acres at an elevation,of 26 feet asl and impounds about 388 acre-feet.
These are the maximum limits authorized in Texas Water Commission Water Supply Permit No. 3233.
o Pg 2-15/Section 2.4.1.1 - Makeup to the ECP is via the MCR. Makeup is not directly provided from the Colorado River to the ECP.
o Pg.2-14and2-16/Section)4.1.1,2.4.2.3- The drainage area numbers l for Little Robbins Slough arE not in agreement.
o Pg. 2-16/Section 2.4.2.3 - All references to Little Robbins Slough on this page should be clarified to Relocated Little Robbins Slough.
o Pg. 2-19/Section 2.4.4.2 - In se and paragraph, last sentence: Should be changed to " . . .by piping. . . " fro.n " . . .of piping. . . "
o Pg. 2-21/Section 2.4.8 - The operating water level in the ECP is between 25.5' and 26.0' msl, which is 1.0 to 1.5 feet lower than the surrounding grade, o Pg. 2-23/Section 2.4.10 - The last portion of the 1st full paragraph is not precisely consistent with input provided in HL&P letter ST-HL-AE-1532 l
(12/16/85). We suggest that the last three sentences be revised to state I
"Should anomalies which require the remedial action procedures to be implemented, administrative controls will be initiated to keep the waterright doors normally closed and knockout panels on the Mechanical Auxiliaty Building in place. These requirements will be addressed by the applicant *in the MCR operating procedures."
o Pg 1-8 stat 6s t. hat the circulating water system supplies 907,400 gpm of cooling; water from the MCR to the main, condenser, while Pg. 2-23 (Section 2.4.11.1) states the value as 1.8 x 10 gpm. Ve suggest changing pg.
2-23 to clarify that this is the value for 2. unit operation.
i o Pg. 2-24/Section 2.4.11.2 - The Section states that the UHS consists of
! the MCR and the ECP, The ECP is the only UHS at the STP, HIAP does not
! consider the MCR as part of the UHS. In addition, the storage volume is 364 acre-feet not 343.8 acre-feet. Note also that in the preceding paragraph that GDC-44 does not apply to the MCR.
i o Pg. 2-25/Section 2.4.11.2 - At elevation 25.5 ms?, the ECP contains about 364.4 acre-feet.
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, Attrchment 1 ST-HL-AE-1674 File No.: G7.3 Page 5 of 18 o Pg. 2-26/Section 2.4.12.1 - The hydraulic copductivity shown of 13.6 cm/sec is based on a value of 200 gal / min /ft asreporteginFSARsection i 2.4.13.1.2. The units in the FSAR should be " gal / day /ft " and are being corrected in an upcoming amendment. The value of 13.6 cm/sec should be revised accordingly, o Pg. 2-34/Section 2.5.1.2 - FM 581 should be FM 521 o Pg. 2-34/Section 2.5.1.2 - The last line in the second full paragraph on this page should be revised from "... site area.." to "... site..." (i.e.,
delete area) to be consistent with definitions in Section 2.5.1 of FSAR.
A similar comment applies to page 2-37 in the last line of 2nd full paragraph. "...in the vicinity of the site." should be revised to "in ,
the plant site."
o Pg. 2-35/Section 2.5.1.2 (3rd paragraph) - HL&P does not believe the hypothesis is widely accepted that the growth faulting in the site area is developed as a consequence of the rise of salt bodies. We therefore suggest deletion of "...widely accepted..." in the 1st sentence.
o Pg. 2-44/Section 2.5.2.5.4 - In the last 2 paragraphs of this section, HL&P would like to note that STP does not consider sudden subsidence to be a generic problem with growth faults. In addition to the staff's reasons for sudden movement not to be a problem, sudden movements are also associated with edges of fields which experience very large fluid withdrawals. STP site does not experience very large fluid withdrawals.
o Pg. 2-45/Section 2.5.3 (3rd paragraph) - Figure 2.5.2-1A should be revised to Figure 2.5.1-1A. Sixth line from bottom, "... south of the site)" should be revised to "... south of the plant site)" to be consistent with definitions in Section 2.5.1. Sixth line from bottom -
, for evaluating the effect of the faults to the plant site, HL&P assumed the worst case of the faults extending to the surface. There is not evidence to suggest that the faults do rise to the surface. Therofore we suggest "...apparently extend to the surface." be revised to "...are assumed to extend to the surface."
l o Pg. 2 49/Section 2.5.4.1.3 - The second sentence under " Deformation, Compressibility and Consolidation" gives the wrong impression of how the tests were conducted - See FSAR Section 2.5.4.2.3.2.1. We suggest that the sentence be split as "The clays under the site are generally overconsolidated. The consolidation testing..."
o Pg. 2-49/Section 2.5.4.1.3 - In the second to last paragraph, odometer j
should be spelled oedometer.
o Pg 2-60/Section 2.5.4.5.4 - The settlement data in the FSAR has been I updated through 1985.
i o Pg. 2-63/Section 2.5.5.2 - Last sentence does not make sense; it appears l words are missing.
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. Attcchment 1 '
ST-HL-AE-1674 File No.: G7.3 ,
Page 6 of 18 l
I 1
o Table 2.11 - It should be notad that differential settlement (between l buildings) as discussed in the FSAR design criteria (Section 2.5.4.11) is applicable after pipe connections are made. SER Table 2.11 is misleading l in that it compares the FSAR design criteria to building movements measured well prior to pipe connections actually being made.
Chapter 3 o Pg. 3-6/Section 3.4.1 - A statement is made that " major tanks containing liquids in the MEAB are housed in watertight compartments that are designed to retain the contents of the tanks". Not all tanks are housed in watertight compartments (reference STP FSAR Section 3.4 through Am.
53) o Pg. 3-6/Section 3.4.1 - In the fourth full paragraph on the page the second sentence is not precisely consistent with the input provided to the NRC on 12/16/85 (ST-HL-AE-1532). We suggest adding the following to the end of the sentence "...while remedial action is taken."
o Pg. 3-12/Section 3.5.2 - A reference should be added regarding the previous SER issued by the NRC on the IVC probabilistic analysis. In fact, we recommend that this SER be included as an appendix to NUREG-0781.
o Pg. 3-14/Section 3.6.2 - There are some exceptions taken regarding the augmented inservice inspection for those portions of piping within the break exclusion region. These are described in FSAR Section 6.6.
o Pg.3-14/Section 3.6.2 - We have changed the CUF to 0.4 based on ANSI /ANS 58.2. This information was transmitted to the NRC via letter ST-HL-AE-1611 dated 2/28/86, o Pg. 3-15/Section 3.6.2 - A statement is made in the SER that "All safety-related equipment near the eliminated break locations is environmentally qualified for the nondynamic effects of a nonsechanistic pipe break with the greatest consequences on the equipment." This is not consistent with HL&P's previous commitments on arbitrary intermediate breaks (AIB's). Recall the following two (2) references on AIB's:
- 1. HL&P Letter ST-HL-AE-1115 dated August 20, 1984, Elimination of Arbitrary Intermediate Pipe Breaks
- 2. NRC Letter ST-AE-HL-90682 dated August 13, 1985, SER for the Elimination of Arbitrary Intermediate Pipe Breaks Reference 1 above requested NRC approval of the elimination of AIB'c and submitted the technical justification for the change. As noted on Page 2 of the HL&P letter, no change was made to the breaks postulated for equipment qualification purposes. Specifically, the following statement was made:
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. Attschm:nt 1 ST-HL-AE-1674 File No.: G7.3 Page 7 of 18
" Application of the above criteria will not affect environmental qualification (EQ) of equipment. Breaks will continue to be postulated non-mechanistically at the previous arbitrary intermediate break locations for EQ purposes."
Reference 2 above determined the change was acceptable and transmitted the Safety Evaluation Report for the subject change.Section III of the SER contains the following basis for the staff decision for both Class 1 and Class 2 and 3.
piping: "2) all safety-related equipment in the vicinity of Class 1 (2 and 3) piping systems have been environmentally qualified for the non-dynamic effects of a non-mechanistic pipe break with the greatest consequences on the equipment."
As noted in NRC Question response 210.35N, the project has proceeded with its design on the basis of the commitment contained in Reference 1, i.e.,
previously established locations for non-mechanistic breaks.
This same comment applies to Pgs. 6 & 9 of SER Appendix G.
o Pg. 3-17/Section 3.7.3 - Clarifications regarding the seismic subsystem analysis are identified on SER pg. 3-17 in Attachment 2 to this letter, o Pg. 3-32/Section 3.9.2.4 - Per the response to NRC Question 210.20, RCL break dynamic effects have been eliminated from the STP design (including asymmetric load) - reference HL&P letter ST-HL-AE-1326.
o Pg. 3-39/Pg. 3-40: Section 3.10.1 - 3.10.2 - Information to respond to the NRC request was submitted to the NRC via HL&P letters ST-HL-AE-1422, 1390, 1435, 1425, 1401, 1510, 1426, 1483, and 1493 which were incorporated in Am. 53. In addition, per discussions with the NRC, a proposed date for the SQRT and PVORT is the week of October 6, 1986.
o Pg. 3-42/Section 3.10.2 - The information to respond to Item (f) was transmitted to the NRC via HL&P letter ST-HL-AE-1597 dated 1/28/86.
o Pg. 3-43,44/Section 3.11.3 - Information to respond to the NRC request was submitted to the NRC via HL&P letters ST-HL-AE-1408, 1422, 1435 &
1404.
Chapter 4 o Pg. 4-1/Section 4.1 - STP only uses Hafnium control rods not a combination of Hafnium and Silver-Indium-Cadmium (P. 4-14/ Sect. 4.2.3.1)
(See SER Pg. 4-24 for consistency) o Pg. 4-35/Section 4.2.5 - The primary boration pathway is via SI not the CVCS.
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.. Att:chment 1 ST-HL-AE-1674 File No.: G7.3 Page 8 of 18 Chapter 5 o Pg.5-3/Section 5.2.2.1 - A statement is made that the pressurizer PORV actuation circuitry is control grade. The actuation circuitry associated with manual operation of the pressurizer PORV's and the cold overpressure mitigation system is fully Class lE. The only portion of the actuation circuitry that is control grade is that portion concerning automatic operation of the pressurizer PORV during normal operation (reference FSAR Figures 7.2-17A&B) . Additionally, a statement is made that the pressurizer safety valves relieve at 2560 psig. The subject valves relieve at 2485 psig as indicated in FSAR Tables 5.4-15 and 5.4-16.
o Pg. 5-4/Section 5.2.2.1 - A statement is made that the resulting peak pressurizer pressure is 2560 psia. This should be 2485 psig for consistency with the previous comment, o Pg. 5-5/Section 5.2.2.2 - The normal seal flow is 8 gpm per pump as stated in FSAR Section 9.3.4.1.2.1.
o Pg. 5-13/Section 5.2.5 - Flow and/or temperature indication is not provided on the leak off lines.
o Pg. 5-27/Section 5.4.7.2 - To meet the commitments contained in BTP RSB 5-1, plant operating procedures will require periodic boron measurements during cooldown. We do not intend to place this requirement in the Technical Specifications, o Pg. 5-32/Section 5.4.12 Item (5) - Change " shutdown panels" to
" shutdown panel". The existing statement implies that there are several auxiliary shutdown panels which is not the case.
Chapter 6 o Pg. 6-1/Section 6.1.1 - The concentration range of borated water in the RWST is 2500-2700 ppm (reference FSAR Table 6.3-1),
o Pg. 6-1/Section 6.1.1 - The percent weight of sodium hydroxide is 30% to 32% as identified in HL&P letter ST-HL-AE-1421 dated 10/31/85.
o Pg. 6-3/Section 6.2.1.1 - This section stytes that the net free ume of the Containment is 3,410,000 ft . The correct value is 3.56 x vo}ft*
10 as identified in FSAR Table 6.2.1.1-3 and SER Section 6.5.2 (Pg.
6-30).
o Pg.6-4/Section6.2.1.1.1-UnpertheheadingSecondarySystemBre_aks, reference is made to a 0.14 ft break size. This should be 1.4 f t'.
o Pg. 6-7/Section 6.2.1.2 - Each steam generator compartment encloses two (2) SG's and 2 RCP's.
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. Attachment 1 ST-HL-AE-1674 File No.: G7.3 Page 9 of 18
- o. Page 6-9/Section 6.2.2 - Under the heading Containment Spray System, the SER states that the design flow rate is 2900 gpe. Per FSAR Table 6.2.2-1, the design flow rate is 1900 gpm.
o Pg. 6-11/Section 6.2.4 - The MSIV's no longer close on SI o Pg. 6-18/Section 6.3.1 - After approximately 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />, the operator
- manually realigns the system for hot-leg recirculation (reference FSAR Section 15.6.5.1) o Pg. 6-20/Section 6.3.2 - The RHR pumps are not used for SI/ECCS for the STP.
o Pg. 6-25/Section 6.3.5.3 - Reference or implication is made that small breaks are classified as Condition IV events. Small break LOCA's
, are classified as condition III events and therefore, per ANSI N660 1
(draft), a 10 minute operator action time is appropriate.
l o Pg. 6-27/Section 6.4 - A statement is made that the capability to alarm in the control room on the detection of high concentrations of either hydrocloric acid, acetic acid or naptha at the outside air intake is not explicitly stated in FSAR Sections 6.4 and 9.4.1. HL&P has
, provided this information to the NRC via letter, ST-HL-AE-1520 dated November 13, 1985. This was subsequently incorporated in FSAR Amendment 53.
j o Pg. 6-27/Section 6.4 (2nd paragraph) - A clarification should be made
- that only two trains of control room KVAC are required, o Pg. 6-29/Section 6.5.1.1 - The function of the control room HVAC makeup and cleanup filtration system is to supply filtered air not necessarily
- l nonradioactive air. Also, note that the third unit is automatically l
started on an SI signal and requires operator action to turn it off. The <
l CR makeup system supplies a minimum of 2000 cfm. .
I o Pg. 6-30/Section 6.5.1.2 - 1st paragraph, " vent stock" should be
" vent stack".
l o Pg. 6-34/Section 6.6.3 - The NRC has stated that the PSI program t
must include a volumetric sample of Class 2 welds in the CVCS.
i 10CFR50.55a(b)(2)(iv) requires that ASME Code Class 2 piping welds in the RHR, ECC and CHR systems shall be examined. The South Texas Project (STP) did not include the CVCS in the Class 2 examinations since the CVCS does not serve an ECCS function. The charging pumps on STP are not used
=
, head and low head safety injection pumps that perform the ECCS function (reference STP FSAR Sections 6.3 (ECCS) and 9.3.4 (CVCS)). Based on this, HL&P believes that our previous position which excluded the CVCS welds constitutes an acceptable position and that the NRC should reconsider their position. The STP is in conformance with NRC
, requirements.
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e Att chment 1 ST-HL-AE-1674 File No.: G7.3 Page 10 of 18 Chapter 7 o Pg. 7-1/Section 7.1.2 - Bechtel Corporation should be Bechtel Energy Corporation.
o Pg. 7-2/Section 7.1.2 (Item 2(d)) - The last sentence should state that the T-Hot average signal is used for overpower Delta-T trips, o Pg. 7-2/Section 7.1.4.1 (Items (3)(b) & (c)) - HIAP has provided information to respond to these items via HIAP letters ST-HL-AE-1311 (08/02/85) and ST-HL-AE-1367 (10/04/85) o Pg. 7-9/Section 7.2.2.3 - A statement is made that the modified design of the RCS temperature measurement system results in a better response time for temperature measurement. This is not true, in actuality the response time is slower but is within the limits of the accident analysis. In addition, maintenance problems are significantly reduced and the ability to verify operability is greatly enhanced, o Pg. 7-12/Section 7.3.1 - The list of initiation signals for the ESFAS function of main steam isolation is incomplete. Items (c), (d) and (e) should be added as follows:
(c) high-2 containment pressure (2/3)
(d) low compensated steamline pressure (2/3 in any steam line)
(e) low-low compensated Tcold (2/3 in any loop) interlock with P-15 o Pg. 7-14/Section 7.3.1.1 - The ESF load sequencer verifies the nonexistence of a Mode III signal not Mode II.
f o Pg. 7-17/Section 7.3.1.8 - This Section states that the volume of the Auxiliary Feedwater Storage Tank is 500,000 gallons. This should be changed to 525,000 gallons for consistency with other Sections of the SER (e.g. , Section 10.4.9/Pg.10-17, Section 9.2.6/Pg. 9-15 (reference HIAP letter ST-HL-AE-1627 dated 3/26/86).
, o Pg. 7-18/Section 7.3.1.12 - The CR makeup system supplies a minimum of 2000 c,fm not a maximum, o Pg. 7-19/Section 7.3.2.1 - A statement is made in the SER that "The l analog signals are converted into digital signals and then fed through solid-state voting logic to determine the actuation status." Note that there are no analog signals within the SSPS. They are converted to digital signals in the 7300 series Process Protection System and in the '
Nuclear Instrumentation System before they are transmitted to the SSPS input relays.
o Pg. 7-24/Section 7.3.2.12 - A statement is made which implies that all of the safuty-related electrical drawings are not available. Note that the l
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- e. Attcchment 1 ST-HL-AE-1674 File No.: G7.3 Page 11 of 18 subject drawings were reviewed with members of the NRC Instrumentation &
Control Systems, Branch (ICSB) during the ICSB audit conducted in March 1985. In addition, copies of the nonproprietary electrical and I&C drawings were submitted to the NRC via HIAP letters ST-HL-AE-1253 (05/17/85) and ST-HL-AE-1278 (06/17/85),
o Pg. 7-27/Section 7.4.1.2 (2nd paragraph, 2nd sentence) - As a clarification, change "...feedwater pump, which are located on the ASP."
to "...feedwater pump, for which transfer switches are located on the ASP."
o Pg. 7-27/Section 7.4.1.4 (Item 3) - The transfer switches are on the transfer switch panels, o Pg. 7-28/Section 7.4.2.1 - The QDPS controls the flow into the SG's through the AFW panel during safety injection. Also, position indication is provided by indicating lights on the main control panel and the ASP.
o Pg. 7-31/Section 7.5.1.1 - The QDPS displays safety parameters; however, the QDPS does not perform the SPDS function described in NUREG-0737.
o Pg. 7-31/Section 7.5.1.1 - The three demultiplexers do not provide outputs to the computer and annunciators.
o Pg. 7-32/Section 7.5.1.2 - The SER Section makes no reference to the use of ERFDAD's for post-accident monitoring. We recommend this information be included (reference FSAR Section 7.5) o Pg. 7-37/Section 7.5.2.9 - This section states "The adequacy of the QDPS will be addressed in a supplement to this SER." Based on several discussions with the Staff over the past two years, we believe that the adequacy of the QDPS is not an issue as implied by this statement. If it j was intended to state that the evaluation of the remainder, other than l the SGWLCS, of QDPS would be addressed in a supplement to the SER, it
! should have been so stated. Section 7.1.4.1(3)(d) would more I appropriately read " evaluation of QDPS."
o Pg. 7-39/Section 7.6.2.2 (Item (2)) - The accumulator isolation valve interlock opens automatically when the main control board switch is in the Auto position not the Open position (reference FSAR Section 7.6.3).
o Pg. 7-48/Section 7.7.2.2 (Item (1)) - As a clarification we recommend changing the following. Change "...to an adverse environment are 1
isolated..." to "...to an adverse environment have been environmentally l qualified. In addition, the PORV's are isolated..."
l o Pg. 7-49/Section 7.7.2.3 - A statement is made that "the applicant shall inform the staff when the modification has been completed." The subject modification has been completed and the staff was informed via HIAP letter ST-HL-AE-1572 dated 1/10/86.
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. Attachment 1 ST-HL-AE-1674 File No.: G7.3
' age r 12 of 18 Chapter 8 o Pg. 8-2/Section 8.2.1 - A statement is made that "There is adequate i
spacing between these sets of towers to allow complete failure of one
. without jeopardizing the other." This statement is not correct for all i
of the towers. Per STP FSAR Section 8.2.1.1 there is adequate spacing between the middle and western towers to allow complete failure of one without jeopardizing the other. For the purpose of analysis, the right-of-way is considered as two independent rights-of-way.
o Pg. 8-3/Section 8.2.1 (paragraph at top of page) - As a clarification we recommend changing the first complete sentence. Change "... current transformers and voltage devices." to "... current transformers and fused i branch circuits from a set of voltage devices." (See marked up page in Attachment #2) o Pg. 8-3/Section 8.2.1 (paragraph at top of page) - Note that the negative terminals are normally tied (see marked up page in Attachment 2).
o Pg. 8-3/Section 8.2.1 (4th full paragraph) - The first sentence should be clarified to identify two of the three preferred offsite power sources (see marked up page in Attachment #23 . '
i
, o Pg. 8-4/Section 8.2.2 - We recommend that the third sentence be rewritten 1 as follows to more accurately describe the STP grid analysis: "The i applicant's steady-state (load-flow) and transient stability analysis using criteria in FSAR Table 8.2.3 and the results shown on FSAR Figures 8.2.6 through 8.2-12 demonstrate that outages of critical generators and l faulting of critical buses will not endanger the supply of offsite power to the ESF electrical systems, nor will they result in overloaded transmission circuits which would hinder the availability of the offsite power supply."
o Pg. 8-5/Section 8.3.1 (2nd full paragraph) V should be 480V o Pg. 8-5/Section 8.3.1 - The 480V motor control centers supply motors of
'less than 100 hp not less than 150 hp (reference STP FSAR Section 8.3) o Pg. 8-6/Section 8.3.1 - The channels for UPS are incorrect - see markup in Attachment #2.
o Pg. 8-7/Section 8.3.1 (3rd full paragraph) - The undervoltage relay setpoint analysis is discussed in response to NRC Question 430.20N and has been submitted to the NRC via HL&P letter ST-HL-AE-1599 dated 1/28/86.
- o Pg. 8-7 & 9 8/Section 8.3.1 - The Technical Specifications have been submitted to the NRC via HL&P letter ST-HL-AE-1549 dated 1/15/86. The information requested has been included.
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. Attcchment 1 ST-HL-AE-1674 File No.: G7.3
- Page 13 of 18 o Pg., 8-14/Section 8.3.3.3.1 - A statement is made that the centerline-to-centerline separation between adjacent electrical penetrations within a given train or channel is 4 feet. Per FSAR Section 8.3.1.4.4.11 the Containment electrical penetrations are physically separated and located at different elevations of the Containment wall.
The vertical and horizontal separation distances between redundant separation groups are not less than the minimum acceptable separation distances of 3 ft horizontally and 5 ft vertically for general plant areas.
I o Pg. 8-15/Section 8.3.3.3.3 - The list sf totally enclosed raceways should be revised based on FSAR Section 8.3.1.4.4.7.
o Pg. 8-16/Section 8.3.3.3.4 (last paragraph) - The non-safety raceways are not Seismic Category I (reference STP FSAR Section 8.3.1.4.4.12).
Chapter 9 o Pg. 9-5/Section 9.1.3 - Per STP FSAR Table 9.1-1 the abnormal pool temperature remains below 158 F not 150 F.
o Pg. 9-7/Section 9.1.4 - A statement is made that the fuel handling system is designed to seismic Category I requirements. Per FSAR Section 9.1.4.1 the subject equipment will not fail in such a manner as to damage seismic Category I equipment during an SSE.
o Pg. 9-13/Section 9.2.3 - A statement is made that there are three 100t capacity DMWS pumps. The pumps are not each 100% capacity (see in attachment #2).
o Pg. 9-13/Section 9.2.4 - A statement is made that the potable water is supplied by onsite wells and is filtered and chlorinated by equipment in an adjacent building. The phrase ". . .by equipment in an adjacent building" should be deleted as it is not true. It appears that the potable and sanitary waste system (PSWS) for the plant has been combined in interpretation with the chlorination facility for the Training Center, o Pg. 9-18/Section 9.3.2.1 - A statement is made that the sampling lines and components of the process sampling system conform to the classification of the system to which each sampling line and component is l
connected. Note that a class break is taken at the root valve unless the sample is inside containment. In this case the class break is taken at the outside containment isolation valve.
i o Pg. 9-20/Section 9.3.2.2 - Reference is made to R.G. 1.3 This R.G. is l applicable to BWR's. We believe the correct reference should be R.G.
1.4.
l o Pg. 9-20/Section 9.3.2.2 (2nd paragraph) - the following change is recommended:
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. Attcchment 1 ST-HL-AE-1674 File No.: G7.3 Page 14 of 18 Delete "...by a grab sample..." and add the phrase "...via the on-line determination with the PASS panel using ion chromatology with a grab sampled backup completed..."
1 This will allow for consistency with the STP FSAR.
4 o Pg. 9-22/Section 9.3.4 - The positive displacement charging pump is not considered an " essential" portion of the CVCS. It is not necessary for normal operation. The primary purpose of the positive displacement pump is that of hydrotesting the RCS. However, it can also be used to provide reactor coolant pump seal injection flow and reactor coolant boration capability for the abnormal condition when both centrifugal charging i
pumps are out of service.
o Pg. 9-22/Section 9.3.4(2nd paragraph) - Suggest changing "The charging seal injection portions. . ." to "The charging and seal injection portions..." (i.e., add 'and' between charging and seal).
J j o Pg. 9-23/Section 9.4.1 - A reference is made to SER Section 12.3 for a discussion of the TSC HVAC system. This appears to be an incorrect reference.
o Pg. 9-27/Section 9.4.1.4 - As a clarification, note that FSAR section 9.4.1.4 commits only to periodic inspection, maintenance and testing for those systems which are subsystems of the Essential Chilled Water System.
The Technical Specifications on those subsystems should be adequate to address NRC concerns.
o Pg. 9-28/Section 9.4.2 - This Section implies that the PASS cooler is safety-related and supplied with safety-related cooling water. This is not true. The PASS supplementary cooling system is not safety-related and is provided with cooling water from a non-safety source (reference
( STP FSAR Section 9.4.2.2.2).
! o Pg. 9-31/Section f 4.3 - A statement is made that a single failure will i only affect one tr in of safety-related equipment. Note that in some cases redundant cooling is provided to the same room to mitigate the j consequences of a single failure in the cooling system. These rooms i contain more than one train of safety-related equipment (reference FSAR Figure 9.4.3-3),
o Pg. 9-43/Section 9.5.1.6 - A statement is made that STP complies with Section III.G.2 of Appendix R inside the RCB. Actually, STP is not in i full compliance with Section III.G.2 of Appendix R. Amendment 3 to the STP FHAR revised the FHAR to show a special hazard, manually actuated, deluge water sprays in containment above and below the Train B cabling which is provided to protect Train B. This spray also provides vertical
. separation between the Train A and Train C cabling. Since all three
! trains have detection, this provides alternate compliance with the l
Appendix R requirement for detection and automatic suppression.
o Pg. 9-45/Section 9.5.1.6 and Pg. 9-57/Section 9.5.4.2 The volume of the diesel fuel oil storage tanks is 67,000 gallons.
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. AttCchment 1 ST-HL-AE-1674 j File No.: G7.3 Page 15 of 18 o Pg. 9-54/Section 9.5.3 - HL&P letter ST-HL-AE-1663 clarifies the STP position on de lighting testing.
o Pg. 9-55/Section 9.5.4.1 (2nd full paragraph) - This paragraph should be 1
rewritten to reflect the results of the DG exhaust stack opening probabilistic analysis. In addition, the exhaust stack no longer extends beyond the edge of the DGB. The NRC was provided with revised design information via HL&P letter ST-HL-AE-1608, dated 2/17/86. This also applies to Pg. 9-64/Section 9.5.8 (4rd full paragraph) o Pg. 9-57/Section 9.5.4.2 - This section provides a discussion of the emergency diesel engine fuel oil storage and transfer system. However, i the opening paragraph talks about the emergency diesel engine lubricating ,
oil system. We suggest that this paragraph be moved to be the first )
paragraph of Section 9.5.7 on Pg. 9-62 (Section 9.5.7 provides a l discussion of the emergency diesel engine lubricating oil system) . ;
Chapter 10 ;
o Pg. 10-1/Section 10.2 - There are three double flow low-pressure turbines at STP not two (reference STP FSAR Sections 10.2.2 & 10.2.2.5).
o Pg. 10-1/Section 10.2 - The electro-hydraulic control for the turbine is analog not digital (reference STP FSAR Section 10.2.2.7).
o Pg. 10-3/Section 10.2 - STP does not have any plans for a Technical Specification on turbine valves, nor have we committed to one in FSAR
! Section 10.2.4. The subject is open for discussion with the appropriate Staff reviewers.
I o Pg. 10-9/Section 10.4.1 - The cathodic protection is used to protect the water boxes not the titanium tubes and aluminum bronze tube sheets (reference STP FSAR Section 10.4.1.2).
o Pg. 10-11/Section 10.4.3 - The condenser exhausts to the outside atmosphere not the TGB atmosphere (reference STP FSAR Section 10.4.3-1) o Pg. 10-18/Section 10.4.9 - The volume of the AFST should be 525,000 gallons for consistency with Pg. 10 17 (the value listed on pg. 10-18 is 518,000 gallons which is the minimum AFST volume required by the Technical Specifications - reference HL&P letter ST-HL-AE-1627, dated 3/26/86). This should be clarified.
o Pg. 10-19/Section 10.4.9 (2nd paragraph) - We suggest changing
"... periodic monthly tests..." to "... periodic monthly valve verifications..."
. Chapter 11 l
! o Pg. 11-1/Section 11.2.1 - 1st paragraph 3rd sentence: Change
, " publishing" to " polishing".
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Attrchment 1 ST-HL-AE-1674 File No.: G7.3 Page 16 of 18
! o Pg. 11-2/Section 11.2.1 - 4th paragraph: Change to read: " cartridge filter, and either an evaporator or mixed bed domineralizer..."; change to read: " waste monitor tank".
J o Pg. 11-2/Section 11.2.1 - 5th paragraph: Add the sentence, " Alternately depending on waste activity levels, waste may be filtered and transferred to a waste monitor tank where it is sampled and discharged from the plant i via the open loop auxiliary cooling water syse m."
o Pg. 11-2/Section 11.2.1 - The condensate regenerant waste value is 4650 gpd per FSAR Table 11.2 - 7.1 and SER Table 11.1 not 4500 gpd.
o Pg. 11-7/Section 11.4.1 - We suggest deleting everything after j
... transferred to..." and adding: "... vendor supplied solidification or dewatering equipment to process liquid waste and spent resins, respectively."
i ~
o Pg. 11-8/Section 11.5 - Redundant radiation monitors are not installed in the main exhaust vent (reference STP FSAR Figure 9.4.3-4 an_ Section 11.5.2.3.3)
Chapter 12
- o Pg. 12-2/Section 12.1.2 - The li ht8 switches are located, whenever possible, outside the radiation area or in areas of low radiation.
The existing wording is misleading (reference STP FSAR Section 12.1.3.1) i o Pg. 12 3/Section 12.2.1 - There should be no noticeable N-16 in the secondary side
- o Pg. 12-5/Section 12.3.2 (3rd full paragraph) - The STP FSAR currently states that the shielding design bases for normal operation are 4,100 MWt
! and 1 percent failed fuel. Exception may be taken to the 1 percent failed fuel fraction in those cases where, due to the advanced stage of construction, the modifications would require excessive shielding changes. In this instance the design basis may change from 1 percent
. failed fuel to 0.25 percent failed fuel on a case by case basis. The wording in the SER is misleading. We suggest using the wording in the STP FSAR.
Chapter 13 o Pg. 13-2/Section 13.1.1.1 - This section should be clarified to state
' that Westinghouse supplies the " initial" fuel assemblies (reference STP
- FSAR Section 13.0) o Pg. 13-2/Section 13.1.1.2 - Include the Manager of Nuclear Security as reporting to the V. P. Nuclear Plant Operations (reference STP FSAR Section 13.1.2.6) o Pg. 13-3/Section 13.1.1.2 - Item (7) should be revised based on the experience history of Mr. Guthrie (reference STP FSAR Table 13.1.-1).
The experience history reflects that of Mr. Phillips who is no longer the Manager, Safeteam.
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. Att:chment 1 ST-HL-AE-1674 File No.: G7.3 Page 17 of 18 o Pg. 13-7/Section 13.2 "LP-8.1" should be "IP-8.1".
o Pg. 13-18/Section 13.5.1.5 - Change to : ...which has responsibility for..."; Remove: " ... lead responsibility" o Pg. 13-18/Section 13.5.1.8 - HL&P letter ST-HL-AE-1650 provides a clarification of our commitment to ANSI 3.1-1981 (meeting the intent of);
we suggest that paragraph be reworded accordingly, o Pg. 13-18/Section 13.5.1.8 " Prerequisite and preoperational test procedures are prepared in accordance with the Startup Administrative Instructions", not the " Plant Procedures Manual".
o Pg. 13-18/Section 13.5.1.8 - Item (2): Note that acceptance testing is for non-safety related systems only, o Pg 13-22/Section 13.5.2.4 - A statement is made that the applicant has not provided the program and related procedures to minimize reactor coolant and radioactive gas leakage outside the containment (reference TMI Item III.D.1.1). This information has previously been transmitted to the NRC via HL&P letter ST-HL-AE-1387 dated 10/29/85 and subsequently incorporated in amendment 53 to the STP FSAR.
Chapter 15 o Pg. 15-1/Section 15.0 - In the introdactory, ACCIDENT ANALYSIS, section, the following correction needs to be made.
- 1) In the large middle paragraph on page 15-1, the sentence starting "The analyses assume a 2-second de'ay between turbine trip and LOOP..." should read, in part, "The analyses assume a 2-second delay between reactor trip and LOOP. . ." This is documented in FSAR Section 15.3.3.2.
These uncertainties are the same as those noted on page 15.0-6 of the FSAR.
o Pg. 15-12/Section 15.2.8 - On page 15-12, the fourth paragraph discusses the "results of the analysis"; the sixth paragraph is redundant, also discussing the identical "results of the analysis",
o Pg. 15-18/Section 15.4.7 - The third paragraph indicates an uncertainty of St. This value is not referred to in the FSAR, Section 15.4.77 o Pg. 15-23/Section 15.6.3 - A statement is made that if credit is taken for PORV operation in the SGTR analysis, then the PORV actuation circuitry will have to be upgraded to safety-grade standards. The actuation circuitry associated with manual operation of the pressurizer PORV's and the cold overpressure mitigation system is fully Class 1E.
The only portion of the actuation circuitry that is control grade is that portion concerning automatic operation of the pressurizer PORV during normal operation (reference FSAR Figures 7.2-17A&B).
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. Att:chment 1 l ST-HL-AE-1674 l File No.: G7.3 '
Page 18 of 18 l Chapter 17 o Pg. 17-1/Section 17.2 (2nd paragraph) - We suggest deleting "all" after
"... verifying compliance with..."
o Pg. 17-1/Section 17.2 (4th paragraph) - Item (1) is now performed by :
Nuclear Training. '
o Pg. 17-1/Section 17.2 (5th paragraph) - Delete " Nuclear Training Section reporting to the Plant Manager."
o Pg. 17-3/Section 17.3 (4th paragraph) - Last sentence: Change to:
... checklists that are reviewed and approved by the QA organization."
o Pg. 17-3/Section 17.3 (7th paragraph) - We suggest adding
"... verification or..." between " perform and followup".
1 Appendix C 4
o Pg. 5 - HL&P has notified the NRC of the use of a revised CUF. The CUF now being used is 0.4. This information was submitted to the NRC via letter ST-HL-AE-1611 dated 2/28/86.
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(48.27 km) northwest. Essentially no hazardous material is transported on the roads in close proximity to the site, except for that used at the South Texas Project. Because of the separation distances between the major highways and the site, accidents that may occur on these roads would not pose a threat to I the safe operation of the plant.
No main railroad lines run within 5 miles (8.05 km) of the site. The app 1tcant examined the potential hazards of materials shipped on the Missouri Pacific, the Santa Fe, and the Southern Pacific lines that operate within the general area. The Missouri Pacific and the Santa Fe are located about 7 miles (11.26 km) north and east of the site, respectively, and the Southern Pacific is about 10 miles (16.09 km) west. Both the Missouri Pacific and the Santa Fe have rail spurs that terminate at the Celanese Chemical Company plant about 4.8 miles (7.72 km) from the site. Smaller amounts of hazardous material are carried on the railroads than are stored at the Celanese plant, which is the rail line's closest approach to the site. Only railroad cars consigned to the South Texas Project, which are operated by the Missouri Pacific railroad, will be allowed on the rail spur leading into the South Texas site.
The Colorado River, which runs generally in a north-south direction and passes within 2.75 miles (4.42 km) east of the site, is the principal waterway in the area. There are no locks or dans on the river in the vicinity of the site.
i Approximately 1200 barges annually transport both raw and finished materials, including petroleum products, up the 15-mile (24.14-km) stretch of the river from the Intercoastal Waterway near the Gulf of Mexico to the turning basin in
- the vicinity of the Port of Bay City. The applicant analyzed the largest ship-ment of petroleum products expected to be transported on the river past the .
l site. Assuming a 15,000-barrel-capacity barge with a 10% gasoline-air mixture, the applicant's calculations indicate that the peak overpressure at the plant from such an explosion would be less than 0.1 psi. The staff considered the possibility of such an accident, evaluated the potential consequences, and agrees with the applicant's conclusion that the risk associated with such an explosion tW does not present a hazard to the plant.
l The Colorado River is the source of makeup water for the onsite re rvoir used for cooling water. Accidents involving river traffic striking the intake struc-ture or spilling corrosive chemicals or other material into the ri.ver would not prevent the safe shutdown of the plant because the available plant cooling water supplied by the onsite reservoir is sufficient for cooldown.
The applicant concludes that the separation distances of the railroads, highways, and waterways in the area, as well as the nature of the roads near the plant and the types and quantity of hazardous materials transported on these routes, are sufficient to preclude adverse effects on the plant in the event of an acci-dental explosion or release. After reviewing the data and evaluating the con-sequences, the staff agrees with the applicant's conclusion.
2.2.2 Nearby Facilities There are no military bases, bombing ranges, munitions plants, or missile in-sta11ations within 10 miles (16.09 km) of the site. No airports are located within 5 miles (8.05 km) of the site, but there are two relatively small airports within 10 miles. C-Level Farm is 9.5 miles (15.29 km) west-northwest, and South Texas SER 2-4
ATTACHMENTIL ST HL AE- /b PAGE c2 0F Collegeport Airfield, which is not in active use, is 8.5 miles (13.68 km) south- I west of the site. A replacement airfield with a single runway has been built f about 0.25 mile (0.4 km) east of the Collegeport runway. In addition, there ,
l' are 18 individual runways within about 10 miles of the site. Except for the '
3700-foot (1128-s) turf runway at C-Level Farm and the new 2800-foot (853-m) l runway at Collegeport, practically all of the other runways are small 1800- to
, 2600-foot (549- to 793-e) grass strips. Essentially all of the aircraft uti-lizing the ground facilities in this area are light weight, used principally for
- crop dusting or other agricultural purposes. During the peak growing season, '
there may be as many as 25 to 100 takeoffs and landings daily at these locations.
The William P. Hobby Airport about 78 miles (125.5 km) northeast and the Houston Intercontinental Airport about 92 miles (148 km) east-northeast are the closest major airports near the plant.
There are no low-level military aircraft training routes near the site. A low-level route (08-19), previously used for bombing and navigation training by the Air Force and Navy, is no longer used. ,
Two low-level Federal Al mays (V-70 and V-20) are within a 10-mile radius of the site. At their closest points, the centerlines of V-70 and V-20 are approxi-mately 5 and 9 miles (8.05 and 14.48 km) northwest of the site, respectively.
On the basis of its review of the applicant's assessment of aircraft hazards at the site, the staff has concluded that the probability of an aircraft crash causing radiological consequences in excess of the guidelines of 10 CFR 100 is within the acceptance criteria of SRP Section 2.2.3 (less than about 10 7 per year) and is, therefore, acceptable.
( Several industrial facilities are within approximately 7 miles (11.26 km) of l! the site. Some of these facilities are operating and some have ceased operations or have shut down coupletely. The Celanese Chemical Company produces, stores, and ships chemicals from its plant, which is about 4.8 miles (7.72 km) north-northeast of the site. The DuPont Co. High Density Polyethylene Plant is about The Crysen Teminal (formerly Bay Tex),
9 '0 ,5.f siles (10.46 km) east of the site.about 4.8 miles (7.72 km) north-northeast, maint '
storage capability of 120,000 barrels. Both Crysen and Celanese utilize the public wharf at the Port of Bay City. This wharf, about 4.8 miles (7.72 km) from the site, handles barge shipments of gasoline, diesel oil, chemicals, and nylon salt. The production equipment at the K and K Compression facility (for-merly the Big Three Industrial Gas and Equipment Co.) located 5.5 miles (8.85 km) north-northeast has been dismantled; there are no plans for future industrial production. Parker Brothers, a docking facility for handling class and oyster shells, is 3.5 miles (5.63 km) east of the site. This facility has ceased oper-ations, and no plans for its future are contemplated. Except for the Parker Brothers facility, all of these industries are located in the vicinity of the Port of Bay City, which is accessible to water, rail, and highway transportation, and the area is highly undeveloped. No large industrial expansion is planned for this area in the near future.
Because of the distances involved, there is no danger of explosions or toxic gas releases at these facilities affecting the safe operation of the plant.
However, because of the anhydrous ammonia, ammonium hydroxide, hydrazine, and vinyl acetate that are stored in the area, redundant detectors for these chemi-( cals will be installed in the outside air intakes of the control room. Wnen these cheefcals are detected, an alarm will soura and the control room will be l
_ wm%,- ,-- ry,- .w. 9, p,, , -wv
ATTACHMENT .2>
ST HL AE &71 PAGE B OF Sg The staff concludes that there are no identified geologic structures or other hazards at the site or in the near vicinity of the site that represent a known safety hazard to the facility.
2.5.2 Vibratory Ground Motion In its review of the seismological aspects of the plant site, the staff has fol-lowed the tectonic province approach to determine the vibratory ground motion corresponding to the safe shutdown earthquake (SSE) (Appendix A of 10 CFR 100).
The staff concluded in the SER-CP that the 0.10g acceleration used as the high frequency input to the RG 1.60 response spectrum was adequate for the South Texas site. Since the conclusion of the CP-review, more information pertaining to the geological character of the site has become available, including more detailed information on growth faults near the site. The applicant assessed the effects on the South Texas structures of postulated movement along these growth faults using the currently available information together with a thorough ground and air survey of the area surrounding the plant. Details of the findings are '
discussed in subsequent subsections. On the basis of the review of*the geological and seismological data and other evidence provided by the applicant, the staff concludes that the South Texas seismic design spectrum used to analyze Category I structures, systems, and components is adequate.
t 2.5.2.1 Seismicity i The Gulf Coastal Plain (PSAR Figure 2.5.1-5) in which the site is located is an area of infrequent and low seismicity. This observation is substantiated by the work of Barstow et al. (1981), which contains an earthquake frequency map of the United States east of the Rocky Mountains. This map indicates that there l
have been less than four earthquakes with a modified Mercalli intensity (MI) greater than III within an area of 4517 min (11,700 km2 ) and a time span of 177 years (1800 to 1977). Barstow et al., Dewey and Gordon (1984), and others have relocated and/or reevaluated many historical events. In response to a staff request, the applicant has provided an updated listing of historical earthquakes making use of the most recent sources. This listing includes more earthquakes than originally' listed in the FSAR. However, it does not appreciably changetheseismicityoftheregioninwhichthesiteislocated. The appli-cant s Table 230.05N-1 indicates that within a 200-mile (321.9-km) radius of the site only 23 earthquakes re reported from 1873 to 1983. None of these GO M occurred withfH % miles km) of the site (applicant's Figure Q230.05N-2).
Table 230.05N-2 lists ear qua es from 1699 to 1983 and within 600 miles (965.6 km) of the site that had an M I equal to or greater than V and/or body wave magnitudes (a b
) equal to or greater than 3.5. In many cases, the inten- ,
sities and magnitudes listed are a result of reevaluations based on geophysical and seismological data that have become available as a result of recent studies l pertaining specifically to eastern U.S. seismicity. l 2.5.2.2 Tectonic Province At the CP review, the staff discussed the characteristics of the tectonic prov- '
ince (Gulf Coastal Plain tectonic province) in which the South Texas site is 10-cated. In particular, the staff recognized four seismic zones: (1) the Missis-sippi Embayment Earthquake Zone, (2) the Southern Cordilleran Front Zone, (3) the zone at the intersection of the Ouachita Tectonic Belt and the Wichita Structural System, and (4) the Gulf Coastal Plain Seismic Zone. '
l
. ATTACHMENT i
. ST HL E /6
,- . PAGE OF Sg variation of modulus with depth used for settlement and subgrade reaction anal-
) y:as is shown on Figure 2.11. Pelsson's rette varied from 0.25 to 0.30.
I The average compressibility inden (0.0014) of compacted structural backfill was computed using values of Young's modulus, also shown en Figure 2.11, and Poissen's ratio with the modulus values computed free CIO triaxial tests. ,
poisson's ratie of 0.25 for the backfill was determined free pubUshed data.
At Rest lateral Earth Pressure At-rest lateral earth pressure was estimated to determine lateral soil pressures acting against buried structures and to assist in selecting stress levels for laboratory soil testing. Three methods were used to determine in situ lateral pressure in the clay: (1) work of Brooker and Ireland (1965) relating K, to the plasticity index and everconsolidation ratio (average K,= 0.92), (2) consolida-
- tien tests (average K, = 0.93), and (3) triaxial compression tests -(Bishop and l Menkel,1962). An in situ K,value of 1.0 was used for design.
j For in situ sand, K, = 0.35 was computed using CIO triaxial test results (f' =
41 degrees). For clean in situ sands, the theoretical at-rest earth pressure 3 coeff'cient varies free about 0.35 for dense sands to about 0.5 for loose sands.
- [h.5 A.10.6.QThe at-rest pressure used for compacted .
l eration the lateral pressure in the upper few feet of the e all as a result of compaction. gy)
- Unit Weinht l Unit weight for in situ soils shown en F5AR Figure 2.5.4-49, was determined j free und'sturbed samples by dividing the wet weight by the esasured volume.
! Ceapaction tests (ASTM 0 1557-70) were perfereed en samples of clay to estab-
! 11sh maximum dry density and optious moisture. Maximum and minimum density J tests were used to establish the unit weight for structural backfill. These
- results are summarized in Table 2.5.
! re eiiity I Laboratory permeability tests were perfomed en undisturbed samples of clay from the site, en compacted clay samples from the ECP, and on compacted structural backfill material. Field tests were used te determine the permeability of the in situ send layers. ficients of permeability ranged from 10.e g, 10.s en/sec for in situ c1 , 10.s for tempacted clay, 16
- for in situ sand, and 10.s for structural backfil . -2 g g-E anne ic Seii prwerties l
1 (PSAt z .s. 4.z. (. .z) j Dynamic soil properties were determined fres $namic laboratory tests and from !
field geophysical measurements of shear-wave veloci . Dynamic laboratory tests i included stress controlled and strain-centrolled cyc ic triaxial tests en undis-
- i. turbed cohesive and cohestenless samples and resonant column tests on reconsti-tuted samples of backfill esterial. ,
i ,
l 1 ;
. ATTACHMEN T4.
- ST HL AE /fe#
PAGE 5 OF ST gesults of the stress-centre 11ed cyclic tests are given in PSAR Tables 2.5.4-6 through 2.5.4-11 strain controlled test results are in F5AR Tables 2.5.4-16 (in situ soils) and 2.5.4-17 (backfill), and resonant-colon results are in FSAR Table 2.5.4-18. Measured and design (sell-structure interaction analyses) shear-wave velocities are shown en Figure 2.12.
Conclusions Laboratory precedures for the various tests for design were in accordance with ASTN standards or other generally accepted practices. The scope and results of '
the laboratory testing program and the recomended design parameters are cen-sidered reasonable and acceptable. Additional testing, perfereed either during er after construction, is described in subsequent sections. ,
1 2.5.4.2 Grounhater Condittens j EfD 2.5.4.2.1 General Site and regional groun heter conditions are discussed in FSAR 5ectie'ns 2.4.13 and 2.5.1.2.10 and in Sections 2.4.12.2 and 2.4.12.3 of this SER. fgesults of groundwater investigations in Matagorda County by the Temas Department of Water Resources and subsequent project investigattens, including the in sonitoring of pieneesters, were used to determine theconditions groundwater,pt at i
the site. The Beaumont Formation ogplies mest of the usabid ter in the
- county and extends from the ground surface to a depth of appres tely 1400 feet in the site area. Grounhater in the formation is confined stan l
pressure. The main producing aquifer sene lies below depths e to 300 feet j I supplies.
in the site area and is the only source for high-capacity water tefere significant groundwater withdrawals, which n in the 1940s, deep aqui-j for pfenseetric levels were.slightly above the surface in the vicinity of South Texas. Since that time, piescestric levels have declined uniferely 1 deep aquifer. Present static grounkster levels range free about t eet he' neath the site. It is estlested that elevations will be red an itional 87 feet between 1973 and 2020. Groun& ster sevements are westward. The deep aquifer sene 15 well below the influence of any construction activities. * -
SD 6 E0A shallow aquifer sene occurs above depths ranging fres 90 feet to 150 feet in the site area and is selpnented into lower and upper units ever most of the site.
This sene has been shown by several site pumping tests aise to be confined with somewhat different pieneestric heads in the two units. The shallow aquifer is 11eited in production capability; only stock and, occasionally, desestic wells pump fres it. piensestric levels are generally within 15 feet of the ground surface. Pienemeters were designed and installed to moniter either the upper er lower portion of the shallow aquifer sone esclusively. The shallow aquifer sens was tasperarily affected by plant construction dovetering but w111 he a11swed to return to the natural elevattens once the plant area subsurface cen-struction and backfill have been completed. Fisw in the shallow aquifer is in ,
a southeasterly direction.
2.5.4.2.2 Foundation Stability Considerations The influence of pieneestric pressures en the foundations and supporting soil was considered for the periods of escavatten, backfill, construction, and plant South Texas SER 2 51
_ _ _ _ _ _ _ _ _ _ _ ~ _ _ _ _ _ _. .______ _ _ _ . - - _ . - -
ATTACHMENT ob
. ST HL AE /G74 PAGE G OF 59 to layer M. The confleuration of the excavation and dewatering system is sho l
en FSAR Figure 2.5.4 64. h construction specification required that the pienemetric elevation be maintained at least 5 feet below the excavation bottor l
within all permeable strata in the shallow aquifer sene that subsequently woulc
> be supporting Category I structures, that is, the E. E, and H 1ayers. This requirement was met during all construction operations as shown en FSAR Fig-ures 2.5.4-65 through 2.5.4-66A. .
Torpede inclinemeter monitoring of the gCS FM local excavations indicated
- 3. lateral movements of in situ soil of approximately 0.1 inch for Unit I and .
g.4 inch for Unit 2 as campared with the allowable deformations of 6 inches l 4
(see Figures 2 14 and 2.15). h se observations documented that all slopes hac i
satisfactory stability against mass movement during construction and*that slope i within soll that would support Category I structures performed better than ex-pected. No adverse events occurred during excavation. Excavation was carried '
eut as unclassified, non-safety-related work.
Foundation preparation and Acceptance -
r
! The final escavation betten was inspected and verified to ensure satisfactory I conditions for the support of Category I structures as described in FSAR f
$ Appendix 2.5.A. Geologic mapping, described in F5AR Appendix 2.5# was also l' perfereed. S ,
h subgrade was prepared to obtain a suitable surface for placement of struc-tural backfill. Criteria established and complied with during subgrade prepa .
tien consisted of the following:
(1) Natural sell subgrade was prepared famediately before fill or backfill w.
placed. 7 i
% (2) Subgrade composed of cohesien soil was campacted to a minimum of 80%
relative density (ASTM D 204 . Layer E was compacted to a minimum dr; ;
i unit weight of 98 lb/ft8 4
- (3) 5derade cospesed of cohesive soil was compacted to 905 maximum dry densi-i j
(ASTM D 1557, mothed D--eodified Proctor).
1 (4) Where fill was placed adjacent to a sleping natural subgrade, the surface was excavated tarfire, undisturtied natural soil famediately before fill o ,
backfill was placed. Average and range of in-place density tests fres j segrade preparation are tabulated in Table 2.6. >
In same locattens, as shown en F54R Figure 2.5.4-116, a concrete sud seal was employed instead of segrade jeeparation by compaction. A minious of 6 inches '
of in site sell was carefully removed to provide an even, undisturbed surface; the exposed soil was inspected as part of the foundation verification program;
- and the sud seal was placed tenediately upon acceptance. The foundation verif
- cation program (F5AR Appendia 2.5. A) was taplemented during construction to det enstrate and document that the naturally deposited set 1s at final escavation i grade were the types anticipated and of a quality equal to er better than thos anticipated during the design phase. Implementation of the foundation verific tien field era followed the event diagram shown en Figure 2.16.
6 2 54 ;
~-. _ _ _ , ,
ATTACHMENT 4L.,
ST.HL AE MM PAGE 7 0F SR The strength end compressibility characteristics of the soils in layer A were considered inadequate for direct support of large and/or heavy buildings, and this layer was removed beneath the major power plant structures. Seil charac-teristics and foundation verification results for final conditions are summa-rised in F5AR Table 2.5.A.5-1 for layers B. C. D. and E.
During initial foundation verification of layer E of Unit 1, it was determined that some sells had a dry density that was less than anticipated: Results of subsequent standard penetration tests (154 to 26 feet below excavation levels) showed that the sells in layer E had a relative density that was lower than antic-1 pated. Low densities to a depth of appreminately 2 feet below encavation levels in the area of the RCS and to depths in encess of 2 feet in the northern portion of the FM were discovered. The lower density in the RCS area was attributed to excavation disturbance. Leesening of sells in the northern portion of the FHB was attributed to installation of casings by vibratory driving. Remedial action in the RCS area consisted of compacting the esposed silty send with a vibratory roller and removing the lessened 2 feet and then proof rolling the exposed foun-dation soil. Remedial action at the north end of the FNB (F5AR Section 2.5.4.12) consisted of vibreflotation of the area. procedures for this remedial work are presented in a Brown & Reet. Inc. report (1976; Rev.1.1978). Field density tests fellowing the remedial actions resulted in dry densities exceeding veri-fication criteria. FSAR Fleure 2.5.4-119 shows standard penetration test data before and after the vibreflotation remedial action.
The staff considers the dewatering escavation and foundation verification pre-gram reasonable and acceptable. On the basis of results presented in the FSAR.
condittens are equal to er better than foundation conditions anticipated for ,
the foundation engineering analyses and design. .
Backfill l I
As indicated in Section 2.5.4.3. the plant area escavatten ion elevations andincluding:
bine generator building area..wes backfilled to the foundat towithin18inchesofgradeadjacenttothestructureswithclean,well-araded, o medium to coarse sand import;ed free searces in the Eagle Lake area (55 elles i north of the plant site). The density criteria for Ca ey I strus;tural back- ;
fill were developed considering the %namic modulus and ing characterist.ics. ,
b cyclic strength pressures. and liquefaction potentiallemeeterescoedthesecriterIaearingc field construction acceptance cri- ,
teria req Wia mini- " relative density and at lent as aver lati l n The manious Isose nches in unWstr cted co- i 1 areas. andh e lifta were used M=S)Jft thickness wesE '
with'n restricted piecesent arves. I i
nic matter, particle shape, and fisid density tests. The este-en includes relative densnty, grain stas, durineeTres erge i
(g2 5 4 5 b h)r Tests
,g i ,,, g,,,
(A5TM standards) perfereed construct J sient of uniformity greater than 4. Actual grGdstion of the Category I s'.ruc- i
@gw tural backfill free samples in the placement areas is shown en Figure 2.13. I In-place density test resolts (statistical) '/or Units 1 and 2 are shown en Fis- l ures 2.37 and 2.18 respectively. The mean of the in-place density distrisutlen :
escoeded GEE relative densuy (July 1990). :
In response to a Show Cause 6 der issued on April 30,1900, by the NRC, an espert semelttee was retained by the esplicant ta perfere an indopth study of the struc-tural beckfill. Nine tasks were originally eutilmed for review covering the ,
\
South Tenas SER t-55
ATTACHMENT ST HL AE /6
_PAGE f OF following generalized subjects: (1) structural backfill design, (2) compaction criteria, (3) construction specifications and procedures, (4) inspection and '
testing procedures, (5) compaction quality control results, (6) special investi-gations of placed backfill, (7) potential remedial actions, (8) inspection of ,
onsite laboratory, and (9) evaluation of test fill. In addition, special studies undertaken by the committee included (1) supplemental testing of soils for maxi- i aus and minimum densities, (2) review of settlement measurements,.(3) anal-ysis of liquefaction potential based on standard penetration test (SPT) data, ,
(4) analysis of liquefaction resistance cf structrual backfill underlying mat foundations,(5 maximum density by wet method, (6) statistical analysis, (7) comparison o)f SPT results with backfill placenent and density test results, (8) locations and sequence of Category I backfill placement and density test distribution, (9) review of Eartnwork Inspection Reports, and (10) review of evaluations and audit findings. Each of these 19 items is discussed in detail in the committee's final report (Hendro"pt, Seed, and Wilson, 1981). In summary, X the committee's overall conclusion was that "the condition of the fill as it now exists is entirely adequate for the design requirements of this project."
The staff considers the field 4cceptance Criteria, placement and compaction procedures, and testing procedures and results reasonable and acceptable. ,
On the basis of the test results in the FSAR and the very detailed review and
, analyses performed by the expert committee, the staff concludes that the Cate-gory I structural backfill for the power plant buildings is acceptable for the design conditions.
2.5.4.4.3 Essential Cooling Water Intake and Discharge Structures Excavation Excavation for the essential cooling water intake and discharce structures began in May 1978 and was completed in June 1978. Olsen-cut excavation, similar to that described for the main plant, extended to elevatten 3.0 m:1 (layer B) for the jntake structure and to elevation 11.0 ms1 (layer As) for the discharge structure. The excavations were extended beyond the perimeter of the structure ,
i to provide uniform foundation support on the structural backfill. Geologic !
inapping and foundation verification, as described for the main plant, were per-forsed (f5AR Appendices 2.5.A and 2.5.8). Dewatering was not needed and the l slopes were stable during construction. Foundation verification zone,s are shown l en FSAR Figure 2.5.4-1178, and test results are presented in FSAR Table 2.5.A.6-1. ;
l i l Backfill i
i The intake structure (79 feet by 136 feet in plan) was founded at elevation 6 feet asl on 3 feet of compacted ,tructural backfill. The trapezoidal-shaped discharge structure (maximum plan dimensions of 54 feet by 53 feet) was founded at elevation 15 feet asl on 4 feet of coepacted structural backfill, Meterial
, types, placement and compaction procedures and requirements, and field testing precedures were basically the same as those described for the main plar*. area.
l The staff considers the excavation procedures, foundation verification procedures and results, and backfill procedures and test results reasonable and acceptable.
!
- ATTACHMENT L
. ST HL AE /GW/
, PAGE 4 OF68 2.5.4.4.4 Essential Cooling Water Piping System 1
l Encavation The essential cooling water (ECW) piping consists of 30-inch-diameter pipe with the inverts varying between 17.5 feet as1 and 10 feet esl in layer As as shown in profile en FSAR Figures 2.5. A.3-1 through 2.5. A.3.3. A silty sene of clay i
intertedded with silt lenses was foun** ing ng the design phase, And when en-dur'trench, countered in the bettes of the ECW p'. it was escavated and replaced l
by structural backfill. Sewatering was not needed and the escavatten slopes were stable. Foundation criteria and verification procedures were established '
j and complied with during construction. Foundation verification senes are shown
)
en F5AR Figure 2.5. A.4-4, and test results are presented in F5AR Table 2.5. A.6-1.
i
! Sackf111 i
N ECW pipes were embedded in structural backfill and were supported either on caspacted structural backfill or en a lean concrete bedding placed individually In narrow strips under the pipeline. Where necessary, the concrete bedding was provided in lieu of campacted backfill to ensure competent contact between the pipe and the backfill underneath. The Category I esterial in which the ECW l
pipes were embedded outside the structural excavatten limits was compacted to a minioue 715 relative density, although design studies indicated that esterial at only 455 relative density would have a facter of sa'."ety of 1.3 against lique-faction. Category I structural backfill sgpertint) the pipes was placed to the horisontal limits of the trench escavation and to a depth of at least 18 inces below the pipes and a height of at least 6 feet #4 eve the pipes.
The staff considers'encavation and backfill for the ECW piping reasonable and l acceptable.
j ,
i 2.5.4.4.5 Essential Coeling pond sec(t.4.s
( '
8'I esc _evetten rea of 46.5 acres
- y. The unter surface of the essential cooling pond (ECP) feet. The nat- ele coversat ural ground surface outside the ECP is appremiestely at elevation feet. Foun-detten verification senes are shown en FSAR Fleure 2.5.4-1175, and test results are presented in FSAR Table 2.5. A.6-1.
The ECF soil investigation during con-atruction aise included both geologic mapping and the escavatten and legging of
.3-9). The escavation ex- ,
K Se test pita in the pond bottan (F5AR Figure 2.5 posed only sells o equipment. Sewatering The escavatten was done with conventional earth-sev trwction.
was not needed, and the slopes were stable during
. Backfill .
b The geologic espring and legging of test pits in the pond bottaa indicated that toward the eastern one-third of the pond area, setts of the As layer became
- chiefly clayey silt (ft.) with silty sand lenses (see FSAR Figure 2.5.g.5.3 10).
The silty sells were encavated te a 2-feet depth below pond battaa, and the area was beckfilled with campacted clay. No other backfill was required for ,
the pond.
South Temas $tt 2-57 ,
L
A l i ALH MLN I .R.,
ST HL AE Nc7(/
PAGE /D OF58
. ML l location, description, installation,andmonitoringpro( res for these instru- .
eents are described in F5AR Appendia 2.5.C. Plots of cti heave and settlement are shown en F54R Figures 2.5.C-9 through 2.5.C-10A from the start of co'nstruc- l tion until Results are summarized._ta Tab' e NJ.
(3)RJDKevci.A4 To teDJlTBd .) :
Differential tween buu m ngs, as determined from the instrumentation :
system, are shown on FSAR Figures 2.5.C-11 through 2.5.C-12A and are summarizea ,
in Table 2.10. The values in Table 2.10 are maafous values that were measured
' from the start of monitoring. As shown on the plots, however, the differential 1 movement between buildings as of the date of the last data (1983) is generally l considerably less than the previously experienced easimum values.
The seasured differential movement profiles within buildiros are shown on FSAR Figures 2.5.C-13A through 2.5.C-14A and are summarized in "able 2.11. ;
. As indicated in Tables 2.10 and 2.11 measured differential movements between '
the buildings of Units 1 and 2 have been relatively small and are within the F design limits established. Differential movements within the buildings are :
, within design limits.
r
( Reevaluation I
Various construction-phase reanalyses were performed between 1977 and 1983 and are discussed in FSAR Appendix 2.5.C. These reanalyses co,nsidered the actual /
construction load schedule, the effect of the main cooling reservoir embankment load and water load to be applied, and the results of the field monitoring pro- -
gram. Projections of future settlement were made using these analyses and in-
' strument data and were checked using a consolidation model based on a compression / :
tension spring analogy. Results of the analyses are presented in Tables 2.12 and 2.13 and are shown on FSAR Figures 2.5.C-9 ,through 2.5.C-10A and 2.5.4-93. i The design criteria for differential movement affecting piping systems are ap-In April 1983, !
plicable after pipe connections between buildings have been made. !
the applicant estimated that no pipe connections would be made before the end of 1984 (day 3400). In the event maximum design movement is approached during plant lifetime, ample time would be available to take corrective action such as ,
adjustment of piping and equipment supports. The applicant indicates that the NRC staff will be advised should differential settlement values approach the. [
design criteria lielts shown in Tables 2.12 and 2.13.
The staff considers the design soil parameters, assumptions, analyses procedures, e and results reasonable and acceptable. Results of the monitoring program for i l
measuring actual movements indicate the structures are performing within estab-lished 1<eits. Actual movements are beginning to stablifre, and, generally, !
small amounts of additional movements are anticipated throughout the operational life of the structures. The buildings should be capable of functioning properly l without any adverse effects free settlement er differential movement.
2.5.4.5.5 Subgrade Reaction l The coefficient of subgrade reaction, K,, or foundation modulus, was used for the structurs1 design of eat foundations as described in FSAR 5ection 2.5.4.10.4. ,
Daformation produced by contact stresses instead of displacements caused by fot.th Ter.as M R 2-60 l
_ ~ _
ATTACHMENT A-ST HL AE-/(oW '
- PAGE // OF5f rwMimd The Category I water level. NormalAoperating level is elevation h5.5 f h portion of the embanYment was protected with a concrer.e racing as shown in plan on Figure 2.22 and in the section on Figure 2.23.
The field investigation for the slopes of the ECP, obtained during the design phase, consisted of 43 boring , 4r slope stability studies, the scils in the upper 20 feet were of prima.; *nterest. These soils are predominantly overconsolidated clays with random slickensides and partly healed desiccation cracks. In the upper 5 feet, the clays are medium stiff to very stiff and of high plasticity. Below 5 feet, the clays are stiff to very stiff and of medium to high plasticity. J 2.5.5.3 Stability Analyses Slope stability analyses were made for static, earthquake, and rapid drawdown conditions for the perimeter dike, the center dike, the excavated pond slope, and the combined dike and pond slope. For both static and pseudostatic analyses, the slopes were analyzed by the modified Bishop circular-arc method. Checks of wedge and other noncircular surfaces were also made with the Morgan' stern-Price method, and the lowest factor of safety was found. Soil strength parameters used in the analyses are summarized in FSAR Table 2.5.5-1. The strength param-eters were determined from consolidated-undrained (CU) triaxial tests from un- The disturbed in situ samples and undisturbed samples from the ECP embankment.
computed factors of safety are compared with the acceptable factors of safety for the conditions analyzed and tabulated in FSAR. Table 2.5.5-2. In all cases the computed factors of safety equaled or exceeded acceptable values.
An analysis was also made of the pore water pressures that might develop in cohesionless soil layers at the ECP during an SSE,and of the effect'of the pore pressures on the postearthquake stability of the dikes and slopes. The analysis indicated that excess pere pressures would be very small and any tendency for strain in the cohesionless layers following an SSE would be accompanied by a tendency for dilation that would immediately dissipate the excess pore pressure.
In addition, a static slope stability analysis was made for the assumption of an increase in pore water pressure equal to 0.25 in cohesionless soft layers.
Results showed that potential sliding surfaces were in clay layers above a depth of 5 feet and did not pass through the cohesionless layers. Therefore, the minimum factors of safety for the dikes and slopes were unaffected by the assumed increase in pore pressure.
2.5.5.4 tocclusions The staff considers the design parameters and methods of analyses reasonable and acceptable. The ECP slopes will remain stable under both static and seismic loading conditions. :.fouefaction potential at the ECP site is discussed in Section 2.5.6 of this SF.R.
2.5.6 habankments and Dans 2.5.6.1 General Description 2.5.6.1.1 Main Cooling Reservoir The main cooling reservoir (HCR), shown on Figure 2.24, will be used to dis-sipqte the excess heat from the circulating cooling water. Principal features South Texas SER 2-64
ATTACHMENT ST HL.AE /6 PAGE /B.0F @
l The results of the liquefaction potential analysis based on the use of standard penetration tests are presented in FSAR Table 2.5.4-35 and indicate a minimum factor of safety of 1.4.
The variation of induced shear stress with depth within the ECP, obtained from response analysis using an artificial accelerogram, is shown on FSAR Fig-ure 2.5.4-124. The cyclic strength characteristics of cohesionless soils were determined from 28 cyclic triaxial tests. The shear stresses to cause initial liquefaction and 10% strain are also plotted on FSAR Figure 2.5.4-124. The mini-num factor of safety for initial liquefaction was 1.6, and for 10% strain it was 2.2.
The analysis performed was for free-field conditions that exist within and adja-cent to the ECP. The overall effect of the dikes and slopes will be to increase to some degree the factors of safety against liquefaction as compared with free-field values.
Conclus tofis ,
The staff concludes that the ECP embankment and dike are low earthen structures
. founded on competent foundations as indicated in the design investigations and later verified during construction. Studies were performed to evaluate slope stability under static and dynamic loading conditions, liquefaction potential, and seepage ccnditions. Adequate margins of safety were obtained for stability and potential liquefaction, and adequate measures were included for control of seepage for the heads that are expected. Slope and embankment protection, in the form of reinforced concrete or soil-coment, is considered adequate. Instru-mentation consisting of piezometers and benchmarks are available for monitoring conditions throughout the life of the facilities. Design, construction, and planned operational aspects of the ECP embankment and dike are also cinsidered adequate. With proper inspection, monitoring, and maintenance, the embankment and dikes should perform satisfactorily without developing any adverse conditions that would affect the safety or safe operation of the plant facilities. If problems develop, a system of deep wells is available as a secondary makeup system for the ECP.
i 2.5.6.4.2 Main Cooling Reservoir Source and Characteristics of Borrow Material
! The embankment and dikes were constructed of onsite clay soils from strata la, I
lb, and 3 and of onsite sand from stratus 2 (sand sources A and B). Locations of these borrow sources are shown on FSAR Figure 2.5.6-14. Sand snurce A, gen-i erally at a depth of 8 to 25 feet below ground surface, contained between 10%
1 and 25% (average 11% to 12%) material that passed through the no. 200 sieve.
I Sand source B, 8 to 35 feet below ground surface, contained 10% to 40% material
( that passed through the no. 200 sieve (average 14% to 15%) 4 cm/see toPermeability of I
thesandforthesandcoreandblanketvariedfrom8:2xQ0 aboratory tests, 4.1 x 10 8 cm/sec. Borrow characteristics, determined fr j are presented by McClelland Engineers, Inc. (1975) and shown on Figure 2.26.
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ATTACHMENT
- ST.HL.AE No7
. PAGE/S OF g heave of the toe or piping. Where this potential condition exists, a horizon-tal sand blanket was provided, as shown on Figure 2,27, to relieve pore water pressures. u 2.5.7 Reevaluation of C.ompleted Main Cooling Reservoir N
To satisfy the provisions of RG 1.59,1"Desisin Basis Floods for Nuclear Power Plants," Revision 0, in effect at the' time the South Texas construction permit was docketed, an instantaneous nonsech .sistic breach of the embankment was pos- g3 ;
tulated to determine flood levels resusting from a brea:h of the MCR. This v (O i postulatedbreach,discussedinFSARSection2.4,consistedoftheinstantaneousj j Flood pro- l removal,orbreach,ofa2000-foot-longsectionofthgembankment.
tection designs of the m th, w.s u-" " ~ ^ safety-related structures d '
and components, including the ECP,* based on the most critical effects of i this breach. Issuance of Revision 2 of RG 1.59 changed Appendix A of the guide to ANSI N170 for guidelines on pcstulation of events producing the design-basis flood. ANSI N170 suggested evaluation of the effects of scour and erosion caused 17 a postulated breach. The applicant chose to make an evaluation of.the integ-rity of the segment of the MCR embankment facing the safety-related plant struc-tures rather than an assessment of scour effects on the plant.
In support of evaluating the integrity of the MCR embankment, the applicant performed field investigations, laboratory tests, and analyses to show that the postulated breach of the dike in the vicinity of plant structures will not occur under all anticipated operational conditions. The applicant has also proposed an inspection and monitoring program fcr continued surveillance of the structure.
The staff has reviewed the results of these efforts. The details are presented in Appendix J.
The staff has concluded that there is reasonable assurance that the MCR dike in the vicinity of plant structure: is capable of containing the reservoir under all anticipstod operational cenditions. The postulated, instantaneous breach of 2000 feet of the section embankment is not considered necessary. Measures l to protect the plant from scour and erosion from such a nonsechanistic breach I are also unnecessary. 8e:ause initial filling of the reservoir to elevation
! 49 feet asl and the period immediately following filling cre most critical to l the' stability and safety of the MCR dike, the staff's 11nal judgment of the MCR l sust await the results of the applicant's future work. Tht staff will report the results of its evaluation in a supplement to this SER. This is a confirma-tory ites.
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e
- ATTACHMENT ST HL AE /0'7 PAGE /& OF 6K Table 2.7 Description of structures Structure General description Reactor containment Prestressed, post-tensioned, reinforced concrete building structure with 166-foot-diameter mat foundation.
Founded at elevation -32 feet ms1 on undisturbed sand and compacted structural backfill.
Fuel handling building Approximately 88 feet x 190 feet in plan. Rein-forced concrete structure. Deeper section (elevation -36 feet) on natural soil and shallower
- portions (-1 to 14 feet msl on structural backfill).
Mechanical and electrical 250 feet x 320 feet in plan, founded at elevation auxiliary building 4 feet ms1 on structural backfill.
Diesel generator building 82 feet x 107 feet in plan, founded at elevation 20 feet ms1 on structural backfill.
Auxiliary feedwater k-foot-diametertankonmatatelevation19 feet X storage tank msl on structural backfill.
Essential cooling water 79 feet x 136 feet in plan, founded at elevation (ECW) intake 6 feet ms1 on structural backfill.
ECW discharge 54 feet x 53 feet in plan, founded at elevation gfeetmslonstructuralbackfill. p
+
Table 2.8 Soil design parameters Total Buoyant Total stress Effective stress unit unit weight weight c 0 c 9 Location (lb/ft3) (lb/ft3) (lb/ft2) (degree) (lb/ft2) (degree)
Structural 134 72 - -
0 43 backfill Layer B 125 63 1430 9 1370 12 Layer D 125 63 1040 16 940 23 Layer E 122 63 - -
0 41 Layer F 125 63 450 18 360 29 South Texas SER 2-110
ATTACHMENT ol.
- PAGE l'7 OF M
- The staff concludes that the plant design is acceptable and meets SRP Sec-tion 3.3.2 and GDC 2 with respect to the capability of the structures to withstand design tornado wind loading and tornado missiles so that their design ref?ects (1) appropriate consideration for the most severe tornado recorded for the site with an appropriate margin (2) appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena (3) the importance of the' safety function to be performed The applicant has met these requirements by using ANSI A58.1 and ASCE paper 3269 to transform the wind velocity generated by the tornado into an effective pres-sure on structures and for selecting pressure coefficients corresponding to the structures' geometry and physical configuration.
The applicant has designed the plant structures with sufficient margin to pre-vent structural damage during the most severe tornado loadings that have been determined appropriate for the site so that the requirements of item 1 above are met. In addition, the design of seismic Category I structures, as required by item 2 above, has included in an acceptable manner load combinations that occur as a result of the most severe tornado wind load and the loads resulting from normal and accident conditions.
t The procedures used to determine the, loadings on structures induced by the '
l design-basis tornado specified for the plant are acceptable because these pro-I I cedures have been used in the design of conventional structures and proven to f provide a conservative basis that, together with other engineering design con-
' siderations, ensures that the structures will withstand such environmental forces. ,
The use of these procedures provides reasonable assurance that, in the event of a design-basis tornado, the structural integrity of the plant structures that have to be designed for tornadoes will not be impaired and, in consequence, safety-related systems and components located within these structures will be adequately protected and may be expected to perform necessary safety functions as required. This satisfies the requirement of item 3 above. .
3.4 Water Level (Flood) Design i
3.4.1 Flood Protection To ensure conformance with GDC 2, the staff's review of the overall flood pro-taction design included those systems and components whose. failure because of flooding could prevent safe shutdown of the plant or result in uncontrolled release of significgnt radioa'ctivity.
l The nominal finished grade level of the plant site is at elevation 28.0 feet mean sea level (msl). The maximum design flood level, including wind wave run-q due to cooling reservoir embankment breach, Colorado River das failures, and
. feet asl on the south face of the fuel l probable maximum precipitation, is on the north face of the essential cooling l handling building and 3Ddi feet as 0 Qed 3-5 South Texas SER I
ATTACHMENT .2__.
ST-HL- AE-No W PAGE /g 0F 45f earthquake (OBE) is 0.05g. The vertical peak ac elerations are taken as two- T thirds cf the peak horizontal accelerations. Both the horizontal and vertical /
design response spectra comply with RG 1.60, " Design Response Spectra for Seis-mic Design for Nuclear Power Plants."
The specific percentage of critical damping values used in the seismic analysis of seismic Category I structures, systems, and components complies with those specified in RG 1.61, " Damping Values for Seismic Design of Nuclear Power Plants."
The only exceptions are values to be used in seismic analysis of some of the cable tray systems.
The applicant informed the staff that the seismic analysis and design of able
)( trays and supports, as an integrated structural system, wi+t-incorporate amp-ing values ranging from 7% to 15% (as opposed to the maximum of 7% specified in RG 1.61 for bolted steel structures) depending on the zero period acceleration at the locations within structures. The applicant cited a test program report,
" Cable Tray and Conduit Raceway Seismic Test Program," test report, release 4, dated December 15, 1978, as a basis for this position. The tests were conducted by ANCO Engineers, Inc., with Bechtel Power Corporation.
At an audit meeting during the week of January 7, 1985, the staff requested that,the applicant provide information to demonstrate that the cable tray and support system is like the cable tray / support system used in the Bechtel/ANCO testing program. In a letter dated May 16, 1985, the applicant compared the following parameters for the two systems: (1) seismic input motion, (2) nat-ural frequencies of cable tray and supports, (3) material and connections for supports, (4) type of tray, (5) method of securing cables, (6) cable tray load- 3 ing, (7) configuration of supports, and (8) types of fire proofing. The staff has reviewed these comparisons between the tested parameters and the South Texas cable tray and support system parameters and finds that the South Texas cable tray system is well represented by the test program. Therefore, the use of 7%
to 15% damping is acceptable. (The staff has adopted similar positions for cable tray systems of other plants when a similarity between the tested system and the plant system has been demonstrated),
l The synthetic time history used for seismic design of Category I plant struc-tures, systems, and components is adjusted'in amplitude and frequency content to obtain response spectra that envelope the response spectra specified for the site.
The seismic Category I structures at South Texas are founded on the support media consisting of alternating layers of stiff to hard clays and dense silts and sand, which extend to depths of several thousand feet. The embedment depths of various Category I structures range *from approximately 4 feet to 59 feet.
On the basis of its review, the staff concludes that the seismic design para-meters used in the plant structure design are acceptable and meet SRP Sec-tion 3.7.1, GDC 2, and Appendix A to 10 CFR 100.
3.7.2 Seismic System Analysis This review is combined with that in Section 3.7.3 below.
I ATTACHMENT A.
- ST HL AE-IfAV 3.7.3 Seismic Subsystem An'alysis PAGE /9 OF 58
) The scope of review of the seismic system and subsystem analyses for the plant included the seismic analysis methods for all Category I structures, systems, and components. It included review of procedures for modeling, seismic soil-
! structure interaction, development of floor response spectra, inclusion of tor-sional effects, evaluation of Category I structure overturning, and determina-l The review has included design criteria and proce-l tion of composite damping.
dures for the evaluation of interaction of non-seismic Category I structures with seismic Category I structures and the effects of parameter variations on l floor response spectra. The review has"also included criteria and seismic analysis procedures for reactor internals and Category I buried structures outside the containment.
l The system and subsystem analyses were performed by the applicant on an elastic and linear basis. Modal time history multidegree-of-freedom methods form the bases for the analysis of all major Category I structures, systems, and com-ponents. When the modal response spectrum method was used, governing response parameters were combined in conformance with RG 1.92, " Combining Modal Responses and Spatial Components in Seismic Response Analyses." The square root of the sum of the squares of the maximum codirectional responses was used in accounting for three components of the earthquake motion. (The component factor method has also been used for this purpose.)
Floor spectra inputs used for design and test verification of structures, sys-tems, and components were generated from the time history method, taking into account variation of parameters by peak widening. A vertical seismic system
,} dynamic analysis is employed for all structures, systems, and components where analyses show significant structural amplification in the vertical direction.
Torsional effects and stability against overturning are considered.
and dynamc.
The inertial effects [f an earthqueke on buried systems and tunnels have been adequately accountedIfor in the analysis. The principles used to account for the effects of statichresistance of the surrounding soil on buried system defor-mations were based on the theory of structures on elastic foundations,g and they are acceptable.
~
aM us. e-prop <uja.W . m*& d
- The applicant has used two-step finite element analysis to account for the soil- ,
structure interaction (SSI) effects for the major Category I structures. In .f m .
the first step, translational and rotational responses (acceleratiorj) at the N8 M s soil / foundation interface are obtained through the use of a two-dimensional pla,ne-strain finite element model. In the second step, the interface accelerationJ +m M*
are applied to fixed-base, detailed, three-dimensional structural models to obtain the floor spectra and responses for the structural design. The appli-s cant performed additional SSI analysis using elastic half-space methodology and a single-step finite element analysis to achieve the following objectives:
(1) To comply with the staff position that requires that the SSI analysis in-clude both elastic half-space (EHS) and finite element approaches for all seismic Category I structures founded on soil (SRP Section 3.7.1, NUREG-
. 0800). Further, these seismic Category I structures should be designed to responses obtained by any of the following methods:
ATTACHMENT c4
. ST HL AE-/bW PAGE BOOF Sg 1
The applicant has met GDC 50 by designing the containment internal structures to accommodate, with sufficient margin, the design leakage rate and calculated I pressure and temperature conditions resulting from accident conditions and by l ensuring that the design conditions are not exceeded during the full course of l the accident condition. In meeting these design requirements, the applicant has used the recommendations of regulatory guides and industry standards indi- i cated below. The applicant has also performed appropriate analyses that demon-strate that the ultimate capacity of the structures will not be exceeded and that establish the minimum margin of safety for the design.
The criteria used in the design, analysis, and construction of the containment internal structures to account for anticipated loadings and postulated conditions that may be imposed during their service lifetime are in conformance with estab-lished criteria, codes, standards, and specifications acceptable to the staff. -
These include meeting the intent of RGs 1.57, " Design Limits and Loading Combina-tions for Metal Primary Reactor Containment System Components," and 1.142 and the following industry standards: ACI-349; ASME Code Section III, Division 2,
" Code for Concrete Reactor Vessels and Containments"; ASME Code Section~ III, Subsections NE and NF; AISC " Specifications for the Design, Fabrication, and Erection of Structural Steel for Buildings"; and ANSI N45.2.5.
The use of criteria as defined or described above provides reasonable assurance that, in the event of earthquakes and various p'ostulated accidents occurring within the containment, the interior structures will withstand the specified design conditions without impairment of structural integrity or the performance of required safety functions.
i 3.8.4 Other Seismic Category I Structures The other seismic Category I structures at the South Texas Project are (1) mechanical and electrical auxiliaries building (2) diesel generator building
. (3) fuel handling building l (4) essential cooling water intake structure
! (5) essential cooling water discharge structure (6) Class 1E underground electrical raceway system (7) auxiliary feedwater storage tank The seismic Category I structures other than the containment and its interior structures are structural steel and/or concrete. The structural components consist of slabs, walls, beams, and columns. The major code used in the design of concrete seismic Category I structures was ACI 318-71. For steel seismic Category I structures, the AISC " Specification for the Design, Fabrication and Erection of Structural Steel for Buildings" was used.
The concrete and steel Category I structures were designed to resist various combinations of deed and live loads; environmental loads including winds, tor-l nadoes, the OBE, and the SSE; and loads generated by postulated ruptures of l high-energy pipes such as reaction and jet impingement forces, compartment pres- ,
' sures, and the impact of whipping pipes. oQ sc% c.4 The current staff position requires that the (ce: ret: int: mrij structurghshould be designed in accordance with ACI-349, " Requirements for Nuclear Safety-Related South Texas SER 3-24
. ATTACHMENT ST HL AE-/l,7 PAGEA) OFSg Structures," as amplified by RG 1.142. By letter dated May 16, 1985, the appli-cant compared the South Texas design criteria (ACI 318-71) and ACI 349-76 and RG 1.142 to assess the impact of the current staff position on structural design (other than containments).
These comparisons indicated that the South Texas design, in general, has met the intent of the current staff requirements. In particular, the controllThg load combinations for the design are the same as those currently required by the staff. (See SER Section 3.8.3 for a dis'assion of the applicant's criteria for welded anchor studs, grouted rock-bolts, and anchor bolts for certain applications.)
On the basis of the above findings, the staff concludes that the design of the
.y containment internal structures is acceptable and meets 10 CFR 50.55a and g GDC 1, 2, 4, 5, and 50.
The applicant has met 10 CFR 50.55a and GOC 1 with respect to ensuring that the g etirut St =D structures are designed, fabricated, erected, constructed,
[;g o tested, and inspected to quality standards commensurate with their safety func-tions to be performed by meeting the regulatory guides and industry standards g
indicated below. ,
The applicant has met GDC 2 by designing the safety-related structures other than the containment to withstand the most severe earthquake that has been established for the site with sufficient margin and the combinations of the l effects of normal and accident conditions with the effects of environmental loadings such as earthquakes and other natural phenomena.
l The applicant has met GDC 4 by ensuring.that the design of the safety-related structures is capable of withstanding the dynamic effects associated with mis-l siles, pipe whipping, and discharging fluids.
The applicant has met GDC 5 by demonstrating that structures, systems, and com-ponents are not shared between units or that, if shared, they have demonstrated that sharing will not impair their ability to perform their intended safety functions.
The applicant has met 10 CFR 50, Appendix B, in this area because the quality assurance program provides adequate measures for implementing guidelines relat-ing to structural design audits.
The criteria used in the analysis, design, and construction of all the plant's seismic Category I structures to account for anticipated loadings and postu-l lated conditions that may be imposed on each structure during its service lifetime are in conformance with established criteria, codes, standards, and
! specifications acceptable to the staff. These include meeting the positions 'of l
RG 1.142, ACI-349, and the AISC " Specifications for the Design, Fabrication, and Erection of Structural Steel for Buildings."
l The use of these criteria as defined or described above provides reasonable assurance that, in the event of winds, tornadoes, earthquakes, and various post-ulated accidents occurring within the structures, the structures will withstand the specified design conditions without impairment of structural integrity or the performance of required safety functions.
- ATTACHMENT .Z ST HL.AE /(gt) ,
PAGE A;tOF Sg )
I 3.8.5 Foundations Foundations of seismic Category I structures are described in FSAR Section 3.8.5. l l
Primarily, these foundations are reinforced concrete of the mat type. The major code used in the design of these concrete mat foundations was ACI 318-71. These concrete foundations have been designed to resist various combinations of dead l loads; live loads; environmental loads resulting from wind, tornado, and seismic effects; and the loads postulated by ruptures of high-energy pipes. .
Thecurrentstaffpositionrequiresthat[he(cen:reteint.cnalstrur+'rebe j designed in accordance with ACI-349, as amplified by RG 1.142. By a letter 4 dated May 16, 1985, the applicant compared the South Texas design criteria I (ACI 318-71) and ACI 349-76 and RG 1.142 to assess the effect of the current L[
staff position on the South Texas structural design (other than containments).
These comparisons indicate that the South Texas design, in general, has met the intent of the current staff requirement. In particular, the controlling load p
- combinations for the design are the same as those currently required by the staff. e, (See SER Section 3.8.3 for a discussion of the applicant's criteria for welded y anchor studs, grouted rock-bolts, and anchor bolts for certain applications.) v On the basis of the above findings, the staff concludes that the design of the d i
containment internal structures is acceptable and meets 10 CFR 50.55a and 3 GDC 1, 2, 4, 5, and 50. J X
The applicant has met 10 CFR 50. Sa and GDC 1 with respect to ensuring that the sn3.ei m-- i n--* strecir:: are designed, fabricated, erected, constructed, -
tested, and inspected to quality standards commensurate with their safety func-tions to be performed by meeting the guidelines of regulatory guides and industry
., standards indicated below.
The applicant has met GDC 2 by designing the seismic Category I foundation to withstand the most severe earthquake that has been established for the site with sufficient margin and the combinations of the effects of normal and accident conditions with the effects of environmental loadings such as earthquakes and other natural phenomena.
The applicant has met GDC 4 by ensuring that the design of seismic Category I foundations is capable of withstanding the dynamic effects associated with missiles, pipe whipping, and discharging fluids.
l l The applicant has met GDC 5 by demonstrating that structures, systems, and compo-nents either are not shared between units or that, if shared, they have demon-strated that sharing will not impair their ability to perform their intended i
safety functions.
1 The use of these criteria as defined or described above provides reasonable l assurance that, in the event of winds, tornadoes, earthquakes, and various pos-l tulated events, seismic Category I foundations will withstand the spec.lfied
! design conditions without impairment of structural integrity and stability or the performance of required safety functions.
1
ATTACHMENT A ST HLSE. /(f]((
PAGE BSOF Sg of the reactor vessel. South Texas has a heated junction thermocouple (W TC) ~)
system supplied by Combustion Engineering, Inc. This system measures the water level inventory in the reactor vessel above the upper core alignment plate even when a steam / water two phase sixt re exists in the reactor vessel. This is accomplished by the use of two lintical probe assemblies, each containing eight N TC sensors with i xfivid al splash shields, which are axially distri-buted inside a separator tube. The NTC sensors located inside this separator tube measure the collapsed water level (water inventory) in the reactor vessel above the upper core alignment plate. An NTC sensor consists of two physically i separated thermocouple junctions, one of which is electrically heated.
The probes are of the split probe design, with two sensors in the upper head region and six sensors in the upper plenum region. This design allows unambig-uous indication of water level in either region regardless of its instantaneous relative pressure. The sensors are located from the top of the vessel down to the top of the fuel alignment plate, giving the operator unambiguous indication of water level during system conditions associated with the approach to and re-covery from ICC. The location of the sensors is described in the applii: ant's letter dated October 24, 1985.
The RWLS is composed of two trains of NTC instruments. Each N TC instrument is assembled into a probe assembly. Each probe has eight electrically inde-pendent NTC sensors as discussed above. Each NTC train is powered from Class 1E power. The cables from the probes are routed to incontainment quali-fied junction boxes. The signals are then routed to the Class IF NTC proces-sors outside containment. The N TC processors perform the following functions:
(1) determine if liquid inventory exists at each NTC sensor position )
(2) provide control of heat power for proper N TC output signal level (3) provide status of each MTC assembly (4) provide a Class 1E redundant datalink with the QOPS, which then transmits I
the following data for display in the control room:
(a) temperature of each heated / unheated thermocouple (b) status of each NTC sensor: covered, uncovered, operating, or failed (c) liquid level inventory above the alignment plate (d) liquid level inventory in the upper head Block diagrams and more detailed descriptions of the RWLS can be found in the applicant's letter dated October 24, 1985.
4.4.6.2 Implementation Schedule The installation of the instrumentation for the detection of inadequate core cooling will be completed before fuel load. ,This schedule is acceptable to the staff.
4.4.6.3 Staff Evaluation The staff has reviewed the applicant's submittals which describe the ICC system proposed for South Texas. These submittals were reviewed for confomance to the requirements of NUREG-0737 Item II.F.2, " Instrumentation for Detection of Inadequate Core Cooling," Item II.F.2, Attachment 1, " Design and Qualification Crite-ia for Pressurized Water Reactor Incore Thermocouples," and Appendix B, South Texas SER 4-32
ATTACHMENT JL,
. ST.Hl. AE /674 PAGEp/ OF58 restraints. In a submittal dated April 4, 1985, the applicant responded to staff Q250.02N on the above topic: "During the PSI of STP Unit 1, HL&P will i document all Section XI examination requirements which are impractical to per-form, identify the limitations to examination of specific welds, and provide a technical justification for non-compliance with Section XI caused by lack of adequate access due to pipe whip restraints on or adjacent to welds subject to examination." In such cases the applicant stated that alternative examination methods (to those specified by Section XI) may be used in lieu of providing a technical justification for nonperformance of the Code-required examinaticns.
This information will be reviewed by the staff when it is submitted by the applidant.
The applicant has committed to identify all plant-specific areas where the Code requirements cannot be met and to provide a supporting technical justification 7forrequestingrelief. The staff review will be completed after the applicant
- (1) addresses the concepts used to select component supports subject to PSI O (2) establishes the methodology to examine the CSS piping welds i I (3) submits all relief reques with a supporting technical justification l The staff considers the review of the South Texas PSI program for systems and components within the reactor coolant pressure boundary a part of the open issue to be resolved subject to the applicant providing the above items. Evalu- i ation of the response will be reported in a supplement to this SER.
, The initial ISI program has not been submitted by the applicant. This program l l
will be evaluated after the applicable ASME Code edition and addenda can be determined on the basis of 10 CFR 50.55a(b) but before inservice inspection i, begins during the first refueling outage. !
5.2.4.4 Conclusions The conduct of periodic examinations and hydrostatic testing of RCPB pressure-retaining components in accordance with Section XI of the ASME Code and 10 CFR 50 will provide reasonable assurance that structural degradation or loss of leak-tight integrity occurring during service will be detected in time to permit corrective action beTore the safety functions of a component are compromised.-
Compliance with the preservice and inservice examinations required by the Code and 10 CFR 50 constitutes an acceptable basis for satisfying the inspection requirements of GDC 32.
j 5.2.5 Reactor Coolant Pressure Boundary Leakage Detection
! A limited amount of leakage is to be expected from components forming the RCPB.
i Means are provided for detecting and identifying this leakage in accordance
, with GDC 30. Leakage is classified into two types--identified and unidentified.
Components such as valve stem packing, pump shaft seals, and flanges are not completely leaktight. Because this leakage is expected, it is considered iden-tified leakage and is monitored, limited, and separated from other leakage (uni-dentified) by directing it to closed systems as identified in Position C.1 of RG 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems."
I In the containment building, identified leakage from valve stems, pump seals, reactor vessel flange, and pressurizer relief valves is kept within a closed I
)
ATTACHMENT i:A
- ST-HL AE-/FM PAGE MOF 58 6.3.4 Testing bility of the ECCS by sub-i The applicant has The applicant has committed to demonstrate the oper for the ECCS, except for committed to meet the intent of RG 1.79, "PreoperaCore h t demon-c:ntainment sump recirculation testing.A program has been established Secticn 6.3.1.
stratcs compliance with GDC 37.
as Preoperational Testing With the ECCS aligned for 6.3.4.1 i l ted and the breakers supply-Tssts will be conducted to verify Proper operation proper of SI signal startup.
generation andnormal p trans-l ding ing offsite power tripped. and delivery rates at pump runout cission, proper startup of the diesel i generators inc u ;
)
up, valve operating times, pump Proper operation starting t mes, of the accumulator isola with RG 1.79 except for I c:nditions would be established. d cnd tianalcheck test program valves would wouldbe beperfomed demonstrated. in accor anced in Section 6.3.1.
emergency sump recirculation testing as discusse Periodic Component Tests d all necessary support sys-6.3.4.2 R:utine periodic testing of the All ECCS ECCScomponents components cananbe testedh on line or tems at power will be performed. hi mini-l have power locked out. Pumps are operated individually supplied in this te l
i a complete cycle. Series check valvesAthat visual form a pressure inspection of pumpboundary are seals, valve flow test lines.lines to perform leak tests. will be made to detect leakage.
l with l sure instrumentation packings, flange connections, and relief va ves Accumulator status will be monitored Operation of accumulator by level and pres and check v isolation during plant operation. is cooled down and RHR is operat-may be verified by means of test lines. f the valves, pump to conduct an integrated test when the ding plantthe starting and loading ing.
This integrated test will demonstrate operability o circuit breakers, andThe applicant automatic has stated circuitry, incluthat the ECCS components intent of ASME Code Section XI.
of the dieseT generators.tnd systems are designed to meet the ur
- e Safety Injection Pump Testing 6.3.4.3 the applicant's response to staffperformed With regard to SI pump performance testing,ut flow together with Q211.31 (Amendment 26) indicated that a series of testsce tes ing subjecting the pump to a thermal d transient four start-stop cycles.at runo Mea-injection of particulate matter; a 100-hour en uran formed at minimum, design, and runout flow; f e andandtion, after hydrauli ,
surements taken during the test included vibra d no unusual or !
seal leakage. The pump experienced no performance d j each test.
excessive wear. te.
The staff concludes that the SI pump tests were adequa l l
r
( !
i I
ATTACHMENT .A.
ST-HL4E /64 l PAGE& OF Sg isolation, a 45-second isolation time (including detection delay), Pasquill )
type F atmospheric diffusion conditions, and a uniform 5 percentile windspeed of 1.5 m/sec. Detection and isolation for chlorine is not needed because a di-luted sodium hypochlorite solution will be used for water treatment processes at the South Texas facility, and only small amounts of chlorine will be stored on site.
The staff has evaluated the control room operator doses following a LOCA using the review procedures outlined in SRP Section 6.4. The major assumptions and parameters used in the analysis are presented in Table 6.1. The resultant estimated radiation doses of 9.1 rems to the thyroid and 3.1 rems to the whole body are within the guidelines of GDC 19.
The staff concludes that for protecting control room personnel under accident conditions, the control room design meets GDC 4 with respect to the design of structures, systems, and components to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents. ,
The staff concludes that the control room habitability system of the South Texas facility is acceptable and meets GDC 19, with respect to maintaining the control room in a safe and habitable condition under accident conditions by providing adequate protection against radiation and toxic gases. This conclusion is based on the staff review and evaluation performed in accordance with SRP Section 6.4 and RG 1.78.
In meeting the guidance and positions of the SRP and regulatory guides, the -
s
, applicant has demonstrated that the control room will adequately protect the )
i, '.', control room operators and will remain habitable in accordance with the post-
,'/ Three Mile Island requirements of NUREG-0737 Item III.D.3.4.
6.5 Engineered Safety Feature Filter Systems 76 ~
The staff has reviewed the applicant's design, design criteria, and design bases for the ESF filter systems according to SRP Section 6.5.1 (NUREG-0800). These acceptance criteria include the applicable GDC, ANSI N509-1996/ " Nuclear Power Plant Air Cleaning Units and Components," and ANSI N510-1994-tTesting of Nuclear 4 g-Air Cleaning Systems." Guidelines for implementation of the requirements of the acceptance criteria are provided in the ANSI standards, regulatory guides, and other documents identified in Section II of the SRP. Conformance to the acceptance criteria provides the bases for the staff's concluding that the ESF filter systems meet 10 CFR 50.
Each South Texas unit has two ESF filter systems: the main control room heating, ventilation, and air conditioning (HVAC) makeup and cleanup filtration unit and the fuel handling building (FHB) exhaust system.
6.5.1 ESF Atmosphere Cleanup Systems From the system description in the FSAR, the staff determined that the ESF at-mosphere cleanup systems are designed consistent with GDC 41, 42, 43, 61, and 64, as referenced in the SRP. In the evaluation of the FHB system design effi-ciencies for removal of elemental iodine and organic iodines, the staff assigned the system decontamination efficiencies of 95% for elemental and organic iodines South Texas SER 6-28 l
ATTACHMENT .2 h
- ST HL AE /F///
. PAGE E/0F5 g !
t 99% for the control room makeup air filtration and 95% for control room recir-culation filtration system. The staff also assigned a 99% removal efficiency for particulates for the HEPA filters in accordance with RG 1.52, " Design, Testing, and Maintenance Criteria for Postaccident Engineered-Safety-Feature Atmopshere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants." The staff determined that the provisions for l instrumentation, readout, and alarm were consistent with SRP Table 6.5.1-1. I Adequate indication and alarms are provided in the control r9ce to ensure proper monitoring of systems performance, per RG 1.52.
On the basis of its evaluation, the staff concludes that the control room make-up air and recirculation cleanup systems are adequately designed to control the concentration of radioactive materials and pressure within the control room in accordance with applicable regulations following a postulated design-basis accident.
6.5.1.2 Fuel Handling Building Exhaust System The function of the fuel handling building (FHB) exhaust system is to fil.ter airborne radioactive iodine and particulates that may leak from spent fuel or from a spent fuel drop accident, or from the emergency core cooling and contain-
! ment spray systems during a design-basis accident. The exhaust flow is dis-chargedthroughtheplantmainventstgek.
The FHB exhaust system maintains the FHB at a negative pressure. The exhaust system consists of two 100% capacity filtration systems and three 50% capacity filtration exhaust fan systems. Each exhaust system consists of three filtration units, and each filtration unit consists, in order, of an electric heater for T i,
humidity control, prefilter, HEPA filter, 2-inch charcoal filter, and a HEPA J filter. Downstream of all the filter units are three 50% capacity exhaust booster fans. During normal operations, two out of three main FHB exhaust fans discharge the flow without filtration. However, should an accident occur or high activity be detected in the FH8 exhaust, three FHB exhaust booster fans wouldstartandtheFHBexhaustwouldbedirectedthroughone[ofthetwofil-tration systems. All the charcoal filters are protected by a- water spray sys-tem in case of overheating or fire. -
'ord 6.5.2 Containment Spray as a Fission Product cleanup System The containment spray system (CSS) has the dual function of removing heat and fission products from a postaccident containment atmosphere. The system, with its chemical additive subsystem, is designed to maintain acceptable water chem-istry parameters in the containment spray during the injection phase following a design-basis accident (DBA) and to ensure that methods are available to raise or maintain the pH of the mixed solution in the containment sump during the recirculation of the containment spray. The chemical additives are provided to enhance the removal of certain chemical forms of radiofodine from the containment atmosphere.
l The CSS consists of three independent, identical trains, each consisting of a
! spray pump, spray additive tank, eductor, valves, piping, and instrumentation.
The spray solution is delivered via two spray risers to the spray nozzles, which are distributed on four concentric spray ring headers located in the upper part of the containment. The CSS is actuated by a containment high-3 perssure signal.
ATTACHMENT C
- ST.HL AE- &2g PAGE ROF 58 cn the basis of 10 CFR 50.55a(b) but before inservice inspection commences
( during the first refueling outage.
6.6.4 Conclusions Compliance with the preservice and inservice inspections required by the ASME Code and 10 CFR 50 constitutes an acceptable basis for satisfying the applicable requirements of GDC 36, 39, 42, and 45.
6.7 Main Steam Isolation Valve Leakage Control System This SRP section does not apply to the South Texas plant because it is a PWR plant.
Table 6.1 Assumptions used in the calculation of control room operator doses following a loss-of-coolant accident Parameter and unit of measure Quantity Activity released to environment See SER Section 15.6.5.1 Control room envelope volume, ft 3 280,000 Pressurization makeup air inflow l ( Flow rate, ft /8 min 2,000 l ESF filter efficiency, %
Elemental iodine 99 Organic iodine 99 Particulate iodine 99 Maximum infiltration rate, ft /8 min , 10 Recirculation air flow Q6%d6)
Flow rate, ft /8 min ESF filter efficiency, %
Elemental iodine 95 t
Organic iodine 95 l Particulate iodine 99 Atmospheric dispersion factors, sec/m8 0-8 hours 8.15 x 10 4 8-24 hours 4.92 x 10 4 1-4 days 1.56 x 10 4 4-30 days 8.72 x 10.s l
I
=_ . - -__ _ _
ATTACHMENT L,
. ST HL AE &W
, PAGEJ49 0F 5 f (d) reactor protection on steam generator water level compensation system ;
(SGWLCS) and hot-leg averaging for overpower W trip
& ,e % 4 6 O r TSince the QDPS design is a microprocessor-based technology, the staff requested 9 the applicant to submit the verification and validation (V&V) program for the o development of Class IE software. The staff's evaluation of the system archi-tecture and the V&V program for the QDPS will be addressed in a supplement ta this report.
Several meetings were held with the applicant and the MSSS and 80P designers to clarify the design and to discuss staff concerns. Detail drawings--including piping and instrumentation diagrams, logic diagrams, control wiring diagrams, electrical one-line diagrams, and electrical schematic diagrams--were audited during the review.
l 7.1.3 General Conclusion l The applicant has identified the instrumentation and control systems important ;
to safety and the SRP acceptance criteria that are applicable to those systems.
The applicant has also identified the guidelines--including the regulatory guides and the industry codes and standards--that are applicable to the systems as identified in FSAR Table 7.1-1.
On the basis of its review of FSAR Section 7.1, the staff concludes that the implementation of the identified acceptance criteria and guidelines satisfies GDC 1, " Quality Standards and Records," with respect to the design fabrication, erection, and testing to quality standards commensurate with the importance of
( the safety functions to be performed. The staff finds that the NSSS and the BOP l instrumentation and control systems important to safety, addressed in FSAR Sec-tion 7.1, satisfy GDC 1 and, therefore, are acceptable.
7.1.4 Specific Findings 7.1.4.1 Open Items The staff's conclusions noted herein are applicable to the instrumentation I
and control systems important to safety with the exception of the open items listed below. The staff will review these items and report their resolution in a supplement to this report. The applicable sections of this report that address these items are indicated in parentheses following each open item.
(1) isolators test report (7.3.2.5)
(2) postaccident monitoring instrumentation confonnance to RG 1.97, "Instrumen-tation for Light-Water-Cooled Nuclear Power Plants To Assess Plant and En-virons Conditions During and Following an Accident,", Revision 2 (7.5.2.4)
(3) the qualified display processing systen ,
(a) software program verification and validation (7.5.2.5)
(b) interface with Class 1E control system (7.5.2.6)
(c) interface with alternate shutdown capability (7.5.2.7)
(d) adequacy of the QOPS (7.5.2.9) 4 1
ATTACHMENT OL_,
ST.HL AE /Mt/
PAGE 200F 59
( The steam generator low-low water level trip protects the reactor from loss of heat sink. This trip is actuated on two out of four low-low water level sig-nals occurring in any steam generator.
A reactor trip on a turbine trip (anticipatory) is actuated from emergency f trip fluid pressure signals or by two out of four closed signals from the turbine steen stop valves. A turbine trip causes a reactor trip above approximately 50K power (P-9 interlock).
A safety injection signal initiates a reactor trip. This trip protects the core against a loss of reactor coolant or a steamline rupture.
An urgent alare is generated when any of the following conditions occur in the corresponding logic train R or 5:
(1) printed circuit board removed or not properly inserted (2) closing of correspondiag bypass breaker (3) operation of the logic train test switches ,
Three separated urgent alarms, one corresponding to each actuation train A, B, or C, are provided when any of the following conditions occur in the corres-ponding actuation train:
(1) actuation train in test (2) loss of ac power to actuation train (3) corresponding safeguards test cabinets in test (4) loss of ac power to the corresponding safeguards test cabinet Occurrence of an urgent alarm in any two out of three actuation trains simul-taneously results in a reactor trip.
The manual trip consists of two switches. Operation of either switch deener-gizes the reactor trip breaker undervoltage coils in each logic train. At the same time, the shunt coils in these breakers are energized, thereby providing a diverse means to ensure that the breakers are tripped. There are no interlocks that can block this trip.
The analog portion of the RTS consists of a portion of the process instrumenta-tion system (PIS) and the nuclear instrumentation system (NIS). The PIS in-cludes those devices that measure temperature, pressure, fluid flow, and level.
The PIS also includes the power supplies, signal conditioning, and bistables that initiate protective functions. The NIS includes the neutron flux monitor-ing instruments, power supplies, signal conditioning, and bistables that initi-ate protective functions.
! The digital portion of the RTS consists of the solid state protection system :
(SSPS). The SSPS takes binary inputs (voltage /no voltage) from the PIS and NIS channels corresponding to normal / trip conditions for plant parameters.
The SSPS uses these signals in the required logic combinations and generates trip signals (no voltage) to the undervoltage coils of the reactor trip cir-cuit breakers. The system also provides annunciator, status light, and com-puter input signals that indicate the conditions of the bistable input sig- 1 nals, partial- and full-trip functions, and the status of various blocking, I
(
l
ATTACHMENT A ST HL AE./6?t/
. PAGE 3 /OF SP 7.2.2.3 Reactor Coolant System Temperature Measurement l(
By letter dated August 2, 1985, the applicant indicated that the reactor coolant temperature measurement system will be modified by eliminating the bypass loop manifold to simplify maintenance procedure. The QDPS will perfom averaging of the reactor coolant hot leg temperature signal and then transmit the signal to s protection and control systems. By letter dated December 4, 1985, the appli-
, cant documented the reactor cool. ant temperature measurement sensor arrangement and calculational methodology in the FSAR. Three fast response, narrow-range resistance temperature detectors (RTDs) mounted in themowells located upstream of the pressurizer surge line connection are used to measure the hot leg tem-perature on each reactor coolant loop. These three inputs per loop are fed to the QDPS where a sensor quality check is performed. If an input is out of a valid range (which is calculated by the QDPS), that input is automatically declared invalid, and an input bias will be used to compensate for the loss of a valid input in a loop. If more than one invalid input is detected in the same loop, the averaging calculation is not performed, and the reactor trip function that requires T hat as an input signal is alarmed on the control, board, and the operator must place that channel in a tripped mode in accordance with the Technical Specifications.
One fast-response narrow-range RTD is located in each cold leg at the discharge of the reactor coolant pump. Temperature streaming in the cold leg is not a concern because of the mixing action of the reactor coolant pump. Therefore, the cold leg temperature input is not processed by the QDPS, and the Tcold signal is directly to the 7300 process protection system. The loop T,yg and AT variables are calculated in the 7300 protection system. In addition, there g " o are three well-mounted fast-response RTDs per loop for the excessive cooldown) protection logic that will remain in the protection system.girhese RTDs can be usedasbackupfor[T M :gggtig{- {new penetrat{o4s are necessary.
The wide-range RTDs located in the hot leg and cold leg piping of each loop are independent from narrow-range RTDs. The wide-range RTDs are qualified for post-accident monitoring. .
The staff has reviewed the modified design of the reactor coolant temperature measurement system including the sensor arrangement and calculation methodology.
The staff concludes that the new design hastbe".ter r**nante ti=alon temperature measurement and less maintenance problems and, g therefore,isacceptaglly
._- e e v.um_ +-;~.a u o sge 7.2.2.4 NUREG-0737 Item II.K.3.12--Confire Existence of Anticipatory Reactor a Trip Upon Turbine Trip The South Texas design includes an anticipatory reactor trip on a turbine trip _ _ . ,
above 50%. of rated thermal power (P-9 interlock). The staff finds that the de- -
sign is in compliance with the TMI Action Plan guidelines. !
7.2.3 Evaluation Conclusion
-i The staff has conducted an audit review of the RTS for conformance to appli-cable regulatory guides and industry codes and standards. In Section 7.1 of this SER, the staff concluded that the applicant had adequately identified the g
( :.
i
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ATTACHMENT ST.HL4E /G7 ((
PAGE AAOF Sf_
(4) feedwater line isolation
) (a) high-high steam generator water leyel (2/4 in any steam generator)
(b) safety injection (see item 2)
(c) low compensated Tcold (2/3 in any 1 op) interlock with P-15 (d) low primary loop flow or low T,yg in two out of four loops (e) high feedwater flow in any loop interlock with P-15 (f) manual (1/2) ,
(5) containment isolation phase A (a) manual (1/2)
(b) safety injection (see item 2)
(6) containment spray actuation (a) manual (either of two sets; both switches in a set must be actuated)
(b) high-3 containment pressure (2/4)
(7) containment ventilation isolation (a) manual (1/2)
(b) safety injection (see item 2)
(c) high radiation (1/2) ,
(8) auxiliary feedwater initiation
,)
(a) manual (each component)
(b) safety injection (see item 2)
(c) low-low steam generator level (2/4 in any steam generator)
(9) centrol room envelope isolation (a) safety injection (see item 2)
(b) high radiation (1/2)
(c) high toxic chemicals (1/2)
(d) smoke (1/2)
(e) loss of offsite power (f) manual 7.3.1.1 Standby Diesel Generators and ESF Load Sequencers Actuation There are three independent, physically separated, standby diesel generators supplying power to three associated load groups designated train A, train B, i and train C. Each standby diesel generator is automatically started and
- Mb g the ESF loads based on three modes of operation
(1) mode I: safety injection (SI) actuation (2) mode II: loss of offsite power (LOOP)
(3) mode III: SI actuation coincident with LOOP 1
ATTACHMENT .L
- ST-HL AE //o79 PAGE 33 0FS F (4) start /stop controls and transfer switches for essential cooling water .
intake structure ventilation fans located at motor control centers (5) open/close controls for various valves not requiring immediate or constant control located at motor control centers (6) disconnect switches for solenoid valves to fail open or close air-operated valves located at the auxiliary relay cabinets
- 7.4.2 Specific Findings 7.4.2.1 Auxiliary Feedwater Control The staff's review of the auxiliary feedwater system (AFWS) included the following:
(1) automatic initiation (2) capability of controlling flow to establish and maintain steam generator l
1evel (3) capability of controlling the steam generator pressure (4) capability of isolating a faulted steam generator resulting from a feed-water or steamline break 9
(5) capability for post-trip control from auxiliary shutdown panel .
The motor-driven pumps are automatically started by the load sequencers on loss of offsite power. However, the isolation valve will not be open until a two-out-of-four low-low water level signal from any steam generator or a safety injection (SI) signal is received. The auxiliary feedwater (AFW) turbine-driven I pump is supplied with steam from steam generator ID through the steam inlet i valve, the steam inlet bypass valve, and the turbine trip and throttle valves.
, These valves all receive open signals on an AFW initiation. The steam inlet l valve receives its open signal through'a time delay (approximately 15 seconds).
l This time delay allows steam flow through the steam inlet bypass valve to ac-celerate the turbine to a speed that allows the turbine governor to assume speed control before the steam inlet valve opens. Manual control of these valves can be performed from the control room or from the auxiliary shutdown panel. The QDPS controls the flow into the steam generators through the AFW regulating valves. The safety grade valve control function is performed by a microprocessor-based control system. Contact output signals for automatic con-trol and position indication of AFW regulator valves within upper and lower flow limit These signals maintain AFW flow within acceptable limits until manual control is assumed by the operator. Manual control capability is provided both in the control room and on the auxiliary shutdown panel.
see perdeA The steam generator power-operated relief valves (PORVs) and tae.r controls are designed as safety-related equipment. A pressure transmitter and pressure con-troller are provided for each of the steam generators to actuate the PORV and control the steam pressure at a predetermined setting. Manual control capa-bility is provided both in the control room and on the auxiliary shutdown panel i
ATTACHMENT
- ST.HL AE /Mt/(
PAGE M0F 52
( neutron flux, etc.) that can cause a reactor trip are either indicated or re-corded for each channel. Parameters associated with automatic actuation as well as those required to enable the operator to manually initiate engineered safety features systems are displayed. The indicators provided for the actuating para-meters display the same analog signals as those monitored by the engineered safety features actuation system. Any reactor trip will actuate an alarm and an annunciator. Such protective actions are indicated and identified down to the channel level.
Alams and annunciators are also used to alert the operator to deviations from normal conditions so that the operator may take appropriate corrective action to avoid a reactor trip. Actuation of any reactor trip channel will actuate an alarm. ,
7.5.1.1 QualfiedDisplayProcessingSystemArchitecture The qualified display processing system (QDPS) is an integrated data acquisition and display system to cover postaccident monitoring, safety parameter display, inadequate core cooling monitoring, emergency response capability, and some limited safety grade control functions. The QDPS perfoms (1) data acquisition and qualified display for postaccident monitoring, referred to as plant safety monitoring system (PSMS) l (2) safety grade control and position indication 'for steam generator PORV, auxiliary feedwater control valve, reactor v'essel head vent valves, l
and essential cooling water (ECW) valves for the chillers i
' (3) data acquisition and qualified display for auxiliary shutdown panel operation if there is a fire in the control room or in the relay room (4) steam generator level measurement compensation for the effect of temper-ature changes at the reference leg (5) averaging of the reactor coolant sys' ten hot leg resistance temperature detector signals to be used for protection and control functions McLuko The QDPS a .a nt; ef four redundant, chann'elized, Class 1E auxiliary process cabinets (APCS). These APCS receive Class 1E signals from the protection system processing cabinet or directly from Class 1E sensors. The APCS send data to redundant data base processing units (DPUs), which provide information to the operator via plasma display. A fifth non-Class IE APC provides data acquisition for non-Class IE signals that are needed to complete logical graphic displays.
The APCS perform engineering units conversioni limit checkt and isolation or buffering. The DPUs perform algorithms and formatting displays for the plasma display units. The operator can use the function keyboard at the display unit requesting information. There are a total of eight plasma display units. Six units are in the control room. They are grouped into two redundant sets, and each set consists of three units. The remaining two display units are on the '
auxiliary shutdown panel. Three demultiplexers provide outputs to drive analog meters, recorders, computer, 5 and annunciators. Two of the three demultiplexers are in the control room; the third is in the auxiliary shutdown area. The 1 1
(
ATTACHMENT L
- ST-HL AE IV/l/
PAGE 350F 52 The following postaccident monitoring recorders are driven by the QOPS
( demultiplexer:
(1) reactor coolant system pressure--wide range (2) steam generator level--wide range (3) containment pressure (4) containment water level--wide range (5) containment water level--narrow range (6) auxiliary feedwater storage tank level (7) auxiliary feedwater flow (8) core exit temperature (9) reactor coolant system subcooling (10) neutron flux--extended range (11) neutron flux--startup rate (12) reactor vessel water level (13) containment hydrogen concentration The staff's evaluation of confomance with RG 1.97, Revision 2, is addressed in Section 7.5.2.4 of this SER.
7.5.1.3 Bypass and Inoperable Status Indication System A system-level bypass and inoperable indicator is provided for each safety-related system. There is a separated lampbox for each ESF train. The component-level lampbox windows provide visual indication that a specific ESF c q.
has been bypassed or deliberately rendered inoperable during nomal plant operating modes. This indication also provides system-level annunciation to
( alert the operator that an ESF system or any of its support systems have been bypassed or deliberately rendered inoperable during normal , plant operating ,
modes. The following conditions (as applicable) are automatically detected for each monitored component of the ESF system:
(1) loss of control power (2) control handswitch in pull-to-lock position .
(3) circuit breaker not in operating position (4) control transferred from the control room to a remote panel (5) component not in its proper aligned position Deliberate manual actions that render ESF-actuated components and devices in-operable are automatically displayed on a component level. Some active com-ponents are not directly actuated by the ESF signal. Automatic display is pro-vided if such components are rendered inoperative more than once a year in such a way that the safety function is compromised.
The capability for initiating a manual bypass indication and an alare is provided via a system-level manual bypass switch to indicate the bypass /in-operable condition to the operator for those components or conditions that are not automatically monitored.
Manual bypass / inoperable indication may be set up or removed under administra-tive control. The automatic indication featur= of the ESF status monitoring system cannot be removed by operator action.
A1TACAMENT A
- ST&L AE- /(dt/
PAGE G OF N reed switch for position indication. Valve position is indicated and alarmed .
k in the control room and indicated on the auxiliary shutdown panel. The three pressurizer safety valves are spring loaded and opened by direct fluid pressure action. Each valve has an acoustic sensing device that actuates an alarm when the valve is not fully closed. The temperature for each safety valve and PORV is measured in the discharge lines and indicated in the control room. An in- ,
crease in a discharge line temperature is an indication of leakage or relief through the associated valve. High temperature will be alarmed in the control room. The valve position sensors are seismically and environmentally qualified.
The valve position indication is powered from the vital instrument bus, and backup methods for determining valve positions are available as an aid to operator diagnosis of an action. The staff finds that the design is in con ,
formance with the Action Plan guidelines and is, therefore, acceptable. l79 7.5.2.2 NUREG-0737 Item II.F.1--Accident Monitoring Instrumentaticn, Positions (4), (5), and (6)
Positions (4), (5), and (6) of this Action Plan ites require installation of the extended range containment pressure monitors, containment water level.
monitors, and containment hydrogen concentration monitors. Table 7.5-1 of the FSAR indicates that the information on these parameters is as follows: ,
(1) Containment Pressure (Extended Range)
(a) The instruments are environmentally and seismically qualified.
(b) The instrument range extended from 0 to 180 psig.
(c) Two channels are provided. '
(d) Two displays and one recorder are provided A the QDPS.
n
~ (2) Containment Water Level (Wide Range)
(a) The instruments are environmentally and sefsnically qualified.
(b) The instrument range is 0 to 6 feet, equivalent to 0 to 609,000 gallons of water.
(c) Three channels are provided.
(d) Two displays and one recorder are provided,af the QOPS.
A (3) Containment Hydrogen Monitors (a) The instruments are environmentally and seismically qualified. ;
(b) The instrument range extended from 0 to 10% concentration.
(c) Two channels are provided.
(d) Two displays and one recorder are provided,Aaf the QDPS.
In an instrumentation and control design review meeting on March 27, 1985, the applicant provided the instrument accuracy and functional requirements for these three items as follows:
(1) containment pressure: accuracy approximately 2% to 3%
function: determine potential for breach of containment
(
. - . , - , , , - . - - - - . , , - . - - - - - - - - - - , - - - _ , - - - - - _ , -------....r,n.
XTTACHMENT A, passo BRadcM cip. cow l PAGE ST.HL AE /M4 TIOFSK FRon1 A S E T O F~
of a 345-kV switchyard circuit breaker. Current and voltage input to both prim-ary and secondary protective relaying for each of the above circuits are respec-tively provided by separate sets of current transformers an 3voltage devices.
The control power for both primary and secondary protective relaying schemes is provided by two separate 125-V dc systems allocated exclusively to the switch-yard. Each 125-V de system consists of a battery, a battery charger, and 125-V dc distribution panel board. These two dc systems are connected by a normally 6p'@tiebreakerandareentirelyindependentoftheunitClassIEandnon-Class 1E battery systems. Each battery charger is connected to a panel board located in the South Texas 345-kV switchyard control house.
(HEGATWC 1EAmmfALS ARE NcAn1 ALLY -fleO)
The control and relaying cables for the switchyard breakers are routed in three parallel, independent cable trenches. The two outer trenches carry the primary relaying and control cables for all switchyard circuit breakers. The center trench carries the secondary (backup) relaying and control for all circuit Dreakers. Cables are routed from each circuit breaker to the respective trenches in such a fashion as to maintain separation between primary and secondary relay-ing and control circuits.
The offsite power from the utility grid and South Texas switchyard is supplied to the unit's onsite distribution system (ESF buses) through the respective unit auxiliary transformer, plant standby transformers nos.1 and 2, three internediate 13.8-kV standby buses, and three ESF 13.8-kV/4.16-kV transformers.
Each sttndby transformer has the capacity to supply all ESF loads of both units and two 13.8-kV auxiliary buses. These two standby transformers can be shared.
between the two units and, together with the unit auxiliary transformer, pro-vide the \wo preferred power sources to each unit.
In addition to the two standby transformers, there is a 138-kV/13.8-kV emergency
' transformer connected to a 138-kV transmission line that is independent of the l 345-kV trarsmission circuits and the South Texas switchyard. This circuit has the capability to be a source of offsite power to one of the three ESF buses in both units at one time. This circuit can be manually connected in case immediate and delayed sources and the onsite diesel generator sources are all out of service. The normal balance-of plant buses do not have access to the emergency transformer.
wo ce ne inG5 PAusnes t
cFFs, rE i The separation between the two standby transformer circuits (th; two offsite-power sources) and the emergency transformer circuit (additional offsite power source) to maintain their independence from each other is achieved by the following design features.
(1) The high-voltage circuits of each standby transformer are routed on separate steel structures and terminated on separate buses in the 345-kV switchyard.
The north bus is extended so the no.1 standby transformer leads do not cross over the south bus.
(2) The location of the steel structure for the 138-kV line to the emergency transformer and the 345-kV lines from the switchyard to standby transformers nos. 1 and 2 are so arranged that a complete failure of the structure ser-ving one of the transformers will not jeopardize the integrity of the structure or the associated high-voltage leads serving the other two transformers.
ATTACHMENT .L ST-HL AE- l(o9tj PAGE38 OF5x (3) The no. I and no. 2 standby transformers and the emergency transformer are physically separated from each other to prevent a single accident of one transformer (e.g., fire) from jeopardizing the operation of the other transformer.
(4) The 138-kV transmission line does not cross any high-voltage lead from the 345-kV switchyard to the plant.
Each of the three transforears is protected by its primary and backup relays.
The control power for the protective relaying system of standby transformers nos. 1 and 2 is provided from the respective unit's non-Class 1E 125-V de system; for that of the emergency transformer, it is provided by a branch circuit from the 125-V de system of 345-kV switchyard. The switchyard bus voltages, breaker status, and battery cor,ditions are indicated and/or alarmed at the control room and Houston Lighting and Power Company energy control center. Transfer from one offsite source of power to the other is essentially manually initiated from the control room.
A breaker-and-a-half arrangement and redundant relaying of the switchyard, trans-mission lines, and the transformers facilitates periodic testing of the offsite
. power system, its protective relaying components, and the 125-V de control power system during plant operation. The generator breaker is routinely tested during plant shutdown.
8.2.2 Grid Analysis The eight 345-kV transmission lines to the switchyard and the 138-kV line to ,
the emergency transformer are interconnected to five electric utilities composed of bulk power and distribution systems called South Texas Interconnected Systems (STIS). These utilities, together with several other distribution systems, form what is called the Electric Reliability Council of Texas (ERCOT). The appli-cant's steady-state (load-flow) and transient stability analysis using criteria in FSAR Table 8.2.3 and the results shown on FSAR Figures 8.2-6 through 8.2-12 demonstrate that outages of critical generator @aulting of critical busesfee l
over! ceded t=s;desien e-irsi_ts-will not endanger the supply of offsite po,wer to the ESF electrical systems Ax The maximum frequency decay was found to be 59.5 Hz and of such short duration that it will not cause any damage to the motors. These results also substantiated that at South Texas, the maximum cred-ible grid frequency decay rate is less than the maximum decay rate of 5 Hz/sec assumed by Westinghouse in the analysis of fuel damage caused by low flow due to grid frequency decay, without tripping of reactor coolant pump breakers.
-McK pa. g -rnev ,qasotq ist CVERLDADED TddsfSnt/SStad C/ACDITS Wit!C.d 4%LD 8.2.3 Conclusion Hwo SR THE A VMLA B /U TV O f TdB Cff$/ TG $O WEA S ON/ Y On the basis of the evaluation in Sections 8.2.1 and 8.2.2 of this SER, the staff has concluded that the design of the offsite power system for South Texas, Units 1 and 2, meets GDC 5, 17, and 18 and is, therefore, acceptable.
8.3 Onsite Power System 8.3.1 AC Power System The onsite ac power system is a Class 1E system and is designed to perform as a standby source of ac power in case the offsite (preferred source) power is not South Texas SER 8-4 l
ATTACHMENT A.
- ST-HL AE &7t/
PAGE M OF Sg )
l l
l available. The safety function of the onsite power system (assuming the off-site power system is not operative) is to provide sufficient capacity and capa-bility to ensure that the structures, systems, and components important to safety perform their safety functions as designed. The objective of the staff review was to determine if the onsite ac power system is designed to have the required !
redundancy, meets the single-failure criterion, can be tested, and is capable of independently supplying reliable power to all required safety loads. The acceptance criterion was to determine if the onsite ac power systems are designed in accordance with the criteria identified in Section 8.1 of this SER.
The standby ac power system is an independent onsite power system designed to :
automatically start on a safety injection and/or loss of offsite power signal and provide adequate power to the Class IE loads to ensure safe shutdown of the plant when the offsite source of power is not available. This system is designed to provide ac power to all Class 1E electrical loads and some selected non-Class 1E loads.
At South Texas, the onsite ac power system consists of three independent divis-ions of safety-related distribution systems. Two of the three divisions are necessary to mitigate the consequences of a design-basis accident. The design meets the single-failure criterion as failure of any one of the three divisions will not jeopardize the safety function of the onsite ac power system. Each division consists of a 4160-V bus supplie qm its independent 13.8-kV/4.16-kV ESF transformer and diesel generator and load centers and motor control ,
centers. The diesel generators (onsite aclsource) are independent of the ESF transformers supplying offsite power. L g e, .y The normal power supply to each of the three divisions is from its own ESF trans-former, which can manually be connected to any one of the four sources of offsite ac power as described in Section 8.2 of this SER. During normal operation one of the three ESF transformers is connected to the unit's auxiliary transformer while the remaining two are connected to the unit's standby transformer. On generator trip, the generator breaker opens and provides an immediate-access off-site power source to the one ESF transformer that is normally connected to the unit's auxiliary transformer. In case the generator breaker fails to open caus-ing the switchyard breaker to trip, the power to the unit auxiliary transformer will be lost and hence to the one ESF transformer connected to it. This will cause the respective diesel generator to start on loss of voltage and provide ac power to the affected division. This will not affect the other two divisions because they will continue to be supplied from the unit's standby transformer, which is not affected by the generator trip. The second offsite power source to the ESF buses can be manually restored as a delayed source of power. This source of power can be restored to all ESF buses from the other unit's standby trans-former or the unit's auxiliary transformer through the main stepup transformer.
The third delayed source of power to one of the three ESF trains in each unit can be restored from the emergency transformer as the last available source of offsite ac power.
All ESF motors with ratings greater than 300 hp are supplied from 4160-V buses, 150- to 300-hp motors are supplied from 480-V load centers and motors of less than 150 hp are supplied from 480-V motor control centers. All these motors are suitable for running at 110% of their nominal voltage rating. All 4.0-kV motors have a service factor ranging from 1.0 to 1.15, whereas generally all 460-V motors have a service factor of 1.15. All Class 1E motors are seismically 4
ATTACHMENT .2 ST HL.AE A7t/
PAGE 40OF 58 and environmentally qualified. Trouble alarms in the control room are provided for the ESF motors. Transformer impedances and standby diesel generator voltage regulator and exciter characteristics are selected to permit starting the largest motor on a particular bus while all other loads connected to the bus are already energized. This will not cause the voltage at the terminals of ESF motors to fall below 80% of their nominal voltage ratings. The applicant has committed to submit the onsite distribution system voltage analysis and the verification test results following the guidelines of BTP PSB-1 to confirm these design fea-tures. The staff will provide its evaluation of the analysis and test results in a supplement to this SER. This is a confirmatory item.
The onsite standby power systems of Units 1 and 2 are designed to operate inde-pendently of each other. Each standby diesel generator and load group in one unit is also physically separate and electrically independent from the other two standby diescl generators and their load groups. Currently, the total load on any one diesel generator is within its continuous full-load rating of 5500 kw The applicant has committed to limit any load increase in the future to be within its 2000-hour rating of 5935 kw. The diesel generator is, however, cap-able of operating at 110% of the continuous rating for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> out of any 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of operation with no reduction in maintenance schedule. All those non-Class IE loads that may be sequenced or manually switched to the diesel generator during a loss of offsite power are automatically isolated on receipt of a safety injection (SI) signal and may be manually loaded after resetting the SI signal.
The ac control power for vital instrumentation and controls in each unit is provided by six uninterruptible power supplies (UPSs). Each of the UPSs consists '
of a solid-state inverter / rectifier system, a de and an ac supply to @ inverter /
rectifier, and a manually operated, mechanically interlocked circuit breaker in each distribution panel'to permit energization of the bus either by the corres-ponding inverter / rectifier or by an alternate 120-V ac, single phase, regulated, "a -~ backup source. T@ of these UPSs are Class 1E. -T= cf thc:c fcr Cicss 1E "-
PUP _Ss-s]up3p power to instrumentation channels I and II@have their separate -semyyfg ower supplies from train A. The other' P5s els III
- i. # N ' " /ac and
' ' and de have their power supplies from trains B@and C; r@espectively.
IVCd Normal source of power to the UPS is 480-V ac power. Upon loss of power from the 480-V ac source, the UPS is automatically powered by the 125-V de system though an auctioneering circuit. There are no manual or automatic interconnections for i
switching between the redundant safety-related UPSs, or between the safety relatec and non-safety-related UPSs or other non-safety-related systems.
4 The staff identified concerns about the use of Ferro-Resonant transformers in Westinghouse inverters or the regulated source of the backup power supply.
Certain deficiencies were identified in these transformers by 1E Information l Notice.84-84. The applicant has committed to corrective action in a letter to NRC (ST-HL-AE-1215) dated March 29, 1985, for all such transformers consistent with Westinghouse Technical Bulletin NSD-TB-84-11. The staff found this correc-tive action acceptable.
The onsite safety-related power system is provided with protective devices to isolate faulted equipment and circuits, prevent damage to equipment, and minimize system disturbances. Each voltage level (4160 V, 480 V, and 208/120 V) is provided with protective relaying with primary and backup protection against various faults and overload conditions. The relay settings are coordinated for selective tripping (primary before backup) and quickly isolate the fault before South Texas SER 8-6
ATTACHMENT A.
. ST.HL AE /V/6 PAGE #IOF,57 Class IE electrical equipment is located in a structure or building that has a seismic Category I classification. These buildings or structures are so designed as to protect the class 1E electrical systems from such postulated occurrences as floods, hurricanes, and other natural events.
In general, major Class.1E electrical power distribution equipment located in the mechanical-ele ~ctcical auxiliary building is arranged so that each train of the three-train ESF system is located on a different floor elevation. Separate rooms or compartments are also provided within each elevation to enhance the physical and electrical independence of each train.
bG3 The standby diesel (generators are each located in a separate room of the diesel generator building # e1 r er* ~ 4:- The associated Class IE electrical equipment locat'ed within each standby diesel generator room is so located and protected within the room as to minimize the possibility of damage resulting from internally generated missiles, pipe ruptures, fire, etc. However, occur-rence of any of these events does not affect the ability of the remaining trains of the'ESF system to perfom their safety functions, since no two trains of
- Class IE equipment or cables are located in or routed through any of the other standby diesel generator rooms.
Non-Class 1E equipment located within seismic Category I structures or buildings j is arranged so that a loss of or damage to this equipment cannot prevent the '
Class IE equipment from performing its safety function. This is accomplished by isolation of such equipment from the Class IE equipment by scans of physical barriers, compartments,, or suitable physical separation.
( Electrical penetration assemblies are provided for cables entering the reactor containment building. Separate quadrants at three different elevations are
! selected for locating these penetrations. Three penetration areas are used for t separate ESF trains and the reactor protection system (RPS) channels. In areat where penetrations for bcth an ESF train and an RPS channel are located, the penetration assemblies are grouped separately. Centerline-to-centerline separ-ation between adjacent electrical penetrations y & -ithin a given train or channel is S& c. 1( T l 4. 'I. 'l " ~ f ** n i '
Control and instrumentation penetrations for W channels I and II. are at the same elevation. However, penetrations u Meir ed with these RPS channels are adequately separated to ensure their Nimit juring any possible event.
8.3.3.3.2 Physical Identification of ;,afety 4 alated Equipment All Class 1E equipment is provided with color-coded nameplates. Safety-related trays and conduits are identified by uniques numbers and colors to designate their train,< channel, or separation group at an interval not exceeding 15 feet.
Similarly, all safety-related cables are color coded in accordance with RG 1.75,
- Revision 2, " Physical Independence of Electric Systews."
8.3.3.3.3 Raceway and Cable Separation
! Cable trays within a given train or a separation group are separated on the
! basis of function and voltage class. Instrumentation cables and other low-level signal cables are rotted in separate raceways from power and control cables. Class 1E circuits of various separation groups are routed in separate i South Texas SER C-14 4
ATTACHMENT .2
)
- ST.HL AE /eWI PAGE C 0F SP On the basis of its review, the staff concludes that the light load handling system is in conformance with GDC 2, 61, and 62 as they relate to protection against natural phenomena and safe fuel handling including prevention of cri- )
ticality, and RG 1.13, Positions C.1, C.3, and C.6, and RG 1.29, Positions C.1 and C.2, with respect to overhead crane interlocks, prevention of unacceptable releases in fuel handling accidents, and maintaining plant safety in a seismic event, and is therefore acceptable. The light load handling system meets SRP Section 9.1.4.
9.1.5 Overhead Heavy Load Handling System The acceptance criteria for the overhead heavy load handling system include conforming to ANS 57.1 and 57.2. The guidelines in the " Review Procedures" portion of the SRP section and NUREG-0612 are used in lieu cf ANS 57.1 and 57.2.
The overhead heavy load handling system consists of components and equipment used to move loads weighing more than one fuel assembly and its associated han-dling device. The equipment includes the containment polar crane and the cask handling crane for safe handling of the reactor ve'ssel head, reactor internals, shielded plug segments, and the spent fuel cask. It also includes the fuel handling building overhead crane and the new fuel handling area crane.
The fuel handling portion of the overhead heavy load handling system is housed' within the seismic Category I, flood- and tornado protected containment and the fuel building. The containment polar crane, the cask handling crane, and other critical components of fuel handling systems are designed to seismic Category I ,
criteria so that they will not fail in a manner that results in unacceptable >
consequences such as fuel damage or damage to safety related equipment. How-ever, the cranes are not required to function following an SSE. Therefore, the design satisfies GDC 2 and RG 1.13, Positions C.1 and C.6, and RG 1.29, Posi-tion C.2. 4 pd @
The fuel handling ildingoverheadcraneisusedforhovingnewfuelassem-blies, cask head, and pool gates, and'other general ha ldling and is designed to withstand a single failure without dropping its load; 'hus, it meets E! REG-UE D The cask handlingacrane and the new fuel handling areadcrane are physically incapable of carrying heavy loads over the spent fuel pool because of their travel limits. No safety-related equipment is located along the path of travel of these cranes. The shipping cask is not lifted to an elevation above any structure high enough to cause damage that could result in unacceptable radio-logical release should the cask be dropped. A dropped cask cannot result in fuel damage in excess of that assumed in the design-basis fuel handling acci-dent, or damage safety-related equipment. For the polar crane, the applicant's reactor vessel head drop accident analysis (RESAR-41NSS) has demonstrated that the postulated head drop will not result in unacceptable consequences. Thus, GDC 4 and 61 and RG 1.13, Positions C.3 and C.5, are satisfied for the handling of heavy loads.
In addition, the applicant has responded to the staff's concern regarding the failure of heavy loads such as concrete shield plugs during handling. The
! applicant stated that such heavy loads including interfacing lift points, which do not satisfy the criteria for a single-failure proof load handling system, South Texas SER 9-8 l
l
ATTACHMENT 2 !
ST HL AE /Mz/ !
PAGE V30F5g motor generator set room, storage rooms, HVAC equipment rooms, radiation moni- l tor room, electrical penetration areas, miscellaneous electrical equipment J rooms, offices, and auxiliary shutdown panel area.
The EAB main area ventilation system is safety related and consists of three l SOE capacity redundant air handling units (AHUs), return air fans, and battery i coes exhaust fans with associated dampers, ductwork, and controls. Outside makeup air is supplied to the EAB main area AHUs from the control room ventila- !
These AHUs consist of profilters, high efficiency l g ~ tion common filters, Acoocoils, inletheating plenum. coils, and supply fans. The electric heating l coiin texcept the) battery room heating coils) are not safety related and are i automatically tripped on an ESF signal. The cooling coils are provided with '
chilled water from the safety-related essential chilled water system. The return air fans draw air from the required rooms via the return air ducts and deliver it to the corresponding main AHU. Return air is mixed with makeup air to form the total air flow through the main AHus. During smoke purge, the l return air fans exhaust the return air to the outside and 100% makeup air is l provided to the AHus. The battery room exhaust fans exhaust the battery rooms to the outdoors with an air change rate sufficient to maintain the hydrogen concentration level below 2% volume. A low flow alarm is provided in the con-l trol room to indicate the operation of the battery room exhaust fan.
The EAB ventilation system also includes the electrical penetration area venti-l 1ation subsystem. It consists of two AHUs with associated recirculating fans, one safety-related and the other non-safety-related, in each related electrical penetration area to recirculate room air and to provide cooling during emer-gency and normal conditions, respectively.. Cooling water to the safety-related cooling coil is provided by the essential chilled water system and to the non-safety related cooling coil by the technical support center chilled water sys-tem. During normal operations, the electrical penetration areas are supplied with ventilating air from the mechanical auxiliary building ventilation system.
The air is exhausted to the outside by two 100% capacity exhaust fans. The supply and exhaust systems are not safety related because they do not serve any safety function. ,
The EAB main area ventilation system is located in seismic Category I, flood-and tornado protected structures. All safety-related components are designed to seismic Category I, Quality Group C requirements. The intake and exhaust plenums are protected against tornado missiles. The failure of any non-safety-related equipment will not affect essential functions of the safety-related ;
equipment. Thus, GDC 2 and RG 1.29, Positions C.1 and C.2, are satisfied.
l All safety-related equipment is powered from its associated Class IE train and I is designed to meet the single-failure criterion. Pneumatic dampers are de-signed to fail in the safe position. Cooling to the safety-related air handl-ing units is provided from the associated essential chilled water system train. l The EAB ventilation system is designed to maintain environmental conditions in 1 the EAB main area compatible with the design limits of essential equipment located therein during normal, transient, and accident conditions. Thus, GDC 4
- is satisfied.
On the basis of the above review, the staff concludes that the EAB main area ventilation system is in conformance with GDC 2 and 4 as they relate to protec-tion against natural phenomena and assurance of a proper environment for the )
\
ATTACHMENT 1._
ST HL AE- /V/4 ,
PAGE 44 OF5e the exception of the supplementary cubicle coolers subsystem, all portions of the MAB ventilation system are not safety related.
The non-safety-related MAB main supply and exhaust system provides ventilation ,
at the proper temperature to several cubicles in the MAB. It consists of four i 25% capac+ty air handling units, each containing a prefi7ter, heating and eeH-M ing coil, three 50% capacity supply fans, and three 50% capacity exhaust fa7s l with associated isolation dampers, ductwork, and instrumentation. The MAB chilled water system provides chilled water to the rem ... ils of the supply g3 air system. Each supply fan is interlocked with its assoc ated exhaust fan to shut down when the exhaust fan trips to preclude positive pressure within the MAB. Air sampling station and radiation monitors are provided in the branch and main ducts that serve several cubicles that could become contaminated. The radiation monitors will alarm if the level exceeds their setpoint.
The safety-related supplementary cubicle cooler subsystem consists of 23 fan coil units. Nineteen units serve safety-related equipment during nomal or abnormal plant operation, and four serve non-safety-related equipment areas.
Cooling water to the fan coil units is provided by the safety-related essential cooling water system (for cubicles containing the component cooling water pumps), the cosponent cooling water system (for cubicles containing the charg-ing pumps), and the essential chilled water system (for cubicles containing the boric acid transfer pump, reactor water makeup pump, valve cubicles, radiation monitor rooms, and essential chiller areas). Power to the safety related fan coil units in each cubicle is provided from the associated Class 1E source.
The supplementary cooler subsystem is designed to seismic Category I, Quality Group C requirements. ,
The non-safety-related supplementary supply and exhaust subsystem provides ven-tilation to locker rooms, offices, and laboratories in the MAB. It consists of an air handling supply unit containing a filter, heating and cooling coils, two l 100% capacity supply fans, and two exhaust filter trains each containing a prefilter, HEPA filter, carbon filter, second HEPA filter, two 100% capacity exhaust fans, and associated isolation riampers, ductwork, and instrumentation.
l The MAB chilled water sy:,tes provides chilled water to tlie cooling coils. Air
~
i flow within the building is maintained from areas of lower radioactivity to i areas of higher potential radioactivity. Airborne particulate and noble gases are continually sampled and. analyzed by a. radiation monitoring system.
The design of the main and supplementary supply and exnaust systems as described above meets GDC 60 and RG 1.140, " Design, Testing and Maintenance Criteria for Normal Ventilation Exhaust System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants," Positions C.1 and C.2.
The non-safety-related MAB chilled water system is a closed-loop system that consists of four 33-1/3% capacity water chillers and recirculation pumps, one expansion tank, and associated piping and valves. The system provides cooling water to the MAB main and supplementary supply ventilation subsystems and fuel l Madling building ventilation system. The non-safety-related open-loop auxil-l iary cooling water system provides cooling to this system.
l
)
1
ATTACHMENT 1.
. ST-HL AE //,W/
PAGE 45 OF5F s
9.4.5 Engineered Safety Features Ventilation Systems (FSAR Sections 9.4.6, 9.4.7, and 9.4.8) )
The engineered safety features (ESF) ventilation systems include the diesel
/ generator building (DGB) ventilation system, essential cooling water intake structure (ECWIS) ventilation system, and main steam isolation valve (MSIV) building ventilation system. Ventilation for the auxiliary feedwater pump area is included in the MSIV building ventilation system. For discussion of venti-lation for the ESF electrical equipment rooms, emergency core cooling system (ECCS) pump rooms, and component cooling water and other pump rooms, see Sec-tions 9.4.1.2, 9.4.2, and 9.4.3 of this SER, respectively.
The DGB ventilation system is designed to maintain a suitable environment for equipment operation and controls and to minimize atmospheric dust levels in the DGB during normal, transient, or accident conditions. The (dy ventilation l system for each diesel area consists of a non-safety-related normal ventilation subsystem and a safety-related emergency ventilation subsystem. The normal ventilation subsystem maintains the ambient temperature of the area when the standby diesel generator is not operating and ventilates the fuel oil tank room. It contains intake and exhaust louvers, intake and return air dampers, inlet filter, supply fan, electric unit heaters, and fuel oil tank room exhaust fan. The DGB emergency ventilation subsystem is designed to operate when the diesel generator is operating to maintain the area temperature within design conditions. It consists of a large-capacity, safety-related supply fan powered from its associated redundant Class 1E bus. The intake and exhaust louvers and chambers are shared with the normal ventilation subsystem. The ventilation air inlet and exhausts are not shared with the combustion air intake and exhaust ;
system described in Section 9.5.8 of this SER.
The ECWIS ventilation system provides outside air to the essential cooling water pump rooms to maintain the area temperature within design limits during normal, transient, or accident conditions. The ventilation for each train of the essential cooling water (ECW) system is independent and consists of intake and exhaust louvers, intake and exhaust dampers, and two 50% capacity fans for each train. The safety-related ventilation fans are powered from their asso-ciated redundant Class IE buses. The fans start automatically on high room temperature or on receipt of the respective ECW pump start signal. The inlet and exhaust dampers fail open on loss of air or power. Unit heaters are provided to prevent freezing conditions within the building when the ECW pumps are not operating.
The MSIV building ventilation system consists of safety-related and non-safety-related ventilation subsystems. The safety-related ventilation subsystem pro-vides ambient air to the auxiliary feedwater (AFW) pump rooms and valve cubicles to maintain their temperature within design limits during normal and transient conditions. It consists of one 100% capacity safety-related supply fan for each of the four AFW pump rooms and valve cubicles. The supply fans are powered from their associated redundant Class IE buses. The ventilation fans are started automatically on high temperature or on receipt of the respective AFW pump start signal. The non-safety-related ventilation subsystem is designed to l provide ambient air to maintain concrete and area temperatures below design l temperature limits in the pipe restraint, valve cubicle, and penetration areas during normal operation, and it consists of two 100% capacity supply fans for South Texas SER 9-32
. ATTACHMENT ds
- ST-HL AE /M4 PAGE Wo OFSg The applicant has provided a fixed emergency communications system that is independent of the normal plant communications system at preselected stations.
A portable radio ccamunications system that uses hand-held portable radios has been provided for the fire brigade. A preoperational test will be conducted to ensure that the frequencies used for portable communications systems will not affect the actuation of protective relays.
Fixed repeaters will be required for portable radio communication. The i repeaters will be protected from exposure fires to maintain. communications.
On the basis of its review and the applicant's commitments, the staff concludes that emergency lighting will be provided in accordance with Section D.5 of Appendix A to BTP APCSD 9.5-1 and Section III.J of Appendix R to 10 CFR 50 and that the emergency communications provisions meet Section D.5 of Appendix A to BTP APCS 8 9.5-1. They are, therefore, acceptable.
. 9.5.1.5 Fire Detection and Suppression ~
1 -
i Fire Detection 4 The fire detection system complies with NFPA 720-1975, " Standard for the Instal-lation, Maintenance and Use of Proprietary Protection Signaling Systems for Guard, Fire Alarm, and Supervisory Service," for a Class A proprietary system.
Detectors are selected and located in accordance with NFPA 72E, " Standard on Automatic Fire Detectors." The fire detection system provides distinctive
' audible and visual alarms locally and in the control room. Primary and second-ary power supplies meet Section 2220 of NFPA 72D-1975. The staff concludes i that the fire detection system complies with'Section E.1 of Appendix A to BTP APCSB 9.5-1 and is, therefor 2, acceptabit. -
In Amendment 2 to the FHAR, the applicant stated that detection systems are pro-vided in areas that contain or present a fire exposure to safe shutdown systems and in all but a few safety-related equipment areas as required by Section III.F of Appendix R to 10 CFR 50. By letters dated October 31 and December 23, 1985, the applicant informed the staff that the areas without detectors fall into the following groups: (1) reactor containment; (2) fuel handling building--cask decontamination area, tank area; (3) mechanical auxiliary building--nonradio-l active pipe chase; (4) mechanical auxiliary building--volume control tank 'and valve rooms; (5) mechanical auxiliary building--HVAC system G==O, CAo*SO (6) postaccident sampling system control room; (7) tendon galleries; (8) per-sonnel airlocks; and (9) stairwells.
The staff's evaluation of the reactor containment fire detection capability is in Section 9.5.1.6 of this SER.
- The remaining areas and zones are devoid of combustible materials, or the com-
- bustible materials are in limited quantities (maximum fire severity of 5 min-utes) and are dispersed. Consequently, a fire of significant magnitude or duration would not be expected to occur. If a fire should occur, it would be detected by fire detectors in adjoining locations, by a partial detection sys-tea in the zone or area, or by plant operators who would summon the plant fire brigade. Until the fire brigade arrived, the perimeter construction of these locations would limit fire spread. The safety-related equipment in these loca-tions consists primarily of metal components, valves, and piping, which are not I South Texas SER 9-40
ATTACHMENT L.
ST-HL AE //,7(/
. PAGE 9/OF Sg overdesigned (subjected to low working stresses) for the application, thereby resulting in high operational reliability. The design of the fuel oil piping and components to the cited design philosophy and standards is considered equi-valent to a system designed to ASME Code,Section III, Class 3 requirements with regard to system functional operability and inservice reliability.
The fuel oil fill line--t us.k fill connection and connections to the yard stor-age tank--is designed as nonseismic Quality Group D. Because each fuel oil storage tank is being provided with its own seismic Category I, ASME Code, Sec-tion III, Class 3 emergency fill connection located on the roof of the diesel generator building, the design is acceptable, i The design of the emergency diesel fuel oil storage and transfer system con-forms to ANSI N195 with the following exceptions:
(1) Sections 6.1 and 8.0--Tanks and Instrumentation and Control j The design of the South Texas emergency diesel generator fuel oil storage
- and transfer system does not include a day tank and its associated instru-mentation. Each 7-day storage tank is located in a room above its respec-tive diesel generator and serves as the diesel generator day tank. It is located at a sufficient elevation to provide a net positive suction head to the engine fuel oil booster pumps. Because of the design, the instru-mentation required for the day tank is not required. The staff finds the design acceptable.
(2) Section 6.2--Pumps Because of the fuel oil storage and transfer system design described
(
above, a fuel oil t- f f pump to transfer fuel from the 7-day storage tank to the day tank is not required.
In addition, the diesel fuel oil storage system conforms with RG 1.137, " Fuel-011 Systems for Standby Diesel Generators," Positions C.2.a through C.2.g, with i
the following exceptions. The fuel oil quality tests will conform to the McGuire Technical Specification for fuel oil quality, as modified by the applicant's letter of September 19, 1985.
The emergency fuel oil fill line is protected from tornado missiles by virtue of the diesel generator roof design. The fuel oil storage tank vent line will shear off on impa:t of a tornado missile. The staff finds this acceptable.
The scope of review of the diesel engine fuel oil storage and transfer system included layout drawings, piping and instrumentation diagrams, and descriptive information in FSAR Section 9.5.4 for the system and auxiliary support sysens l essential to its operation.
I The basis for acceptance in the staff review was c(nformance af the design cri-teria and bases and design of the diesel engine fuel oil storage and transfer system to GDC 17 with respect to redundancy and physical independence, the
- cited regulatory guides, NUREG/CR-0660, and industry codes and standards.
On the basis of its review, the staff concludes that the emergency diesel engine fuel oil storage and transfer system meets GDC 2, 4, 5, and 17, NUREG/CR-0660, (-
i 9-58 l
ATTACHMENT A ST-HL-AE /foW
- PAGE #f 0F 58 i
,( the cited regulatory guides, SRP Section 9.5.4, and industry codes and stand-
' \ ards; thus, it can perform its design safety function and, therefore, is acceptable.
9.5.5 Emergency Diesel Engine Cooling Water System The design function of the emergency diesel engine cooling water system is to maintain the temperature of the diesel engine within a safe operating range under all load conditions and to maintain the engine coolant preheated during l standby conditions to improve starting reliability. The system is designed to meet GDC 2, 4, 5, 17, 44, 45, and 46. Conformance with GDC 2, 4, and 5 is dis-cussed in Section 9.S.4.1 of this SER.
The emergency diesel engine cooling water syses provides cooling to the engine jacket, lube oil cooler, governor oil cooler, fuel oil cooler, and air heaters /
intercoolers, and is composed of two subsystems: the closed-loop cooling water system and the open-loop cooling water system.
(1) The major components of the closed-loop cooling water system for each standby emertency diesel engine include an engine-driven jacket M water pump, ; acket water heat exchanger, an expansion tank (standpipe), an ac motor-driven circulation puep, combustion air heaters /intercoolers (one
- for each cylinder bank), an ac motor-driven jacket creMstandby pump, l an electric heater, and thermostatic three-way valves, as well as the re-quired instrumentation, controls, and alams, and the associated piping and valves to connect the equipment. When the diesel engines are operat-ing, the heat generated is rejected te the essen.tial cooling water system
( by means of the jacket ;="-f1 vater heat exchanger.
During operation of the standby diesel engine, the temperature of the l
diesel engine coolant is regulated automatically through the action of a j temperature-sensing three-way thermostatic valve. When the standby diesel engine is idle, the engine coolant is heated by an electric heater and con-tinuously circulated through the The temperature is controlled by a thermostat to keep the engin, engine.e wam and ready to accept loads within the prescribed time interval.
The diesel generator is capable of operating fully loaded without secondary cooling for approximately 2 minutes. The engine and expansion tank contain enough water to absorb the heat generated during this period. This inter-val is greater than the time needed to restore essential cooling water to the diesel engines if a loss of offsite power should occur.
(2) The open-loop cooling watcr system is a subsystem of the essential cooling water system. Each diesel generator has its own open-loop cooling water by separate trains of the essential cooling water system. This system provides cooling water for the diesel generator closed-loop cooling water system, the lube oil cooler, the fuel oil cooler, the governor oil cooler,
'and the combustion air intercoolers (one cooler per cylinder bank).
Alarms have been provided to enable the control room operator to monitor the diesel generator cooling while the unit is in the standby mode or in operation.
ATTACHMENT A-ST HL AE- &7//
- PAGE 44 OF 53 i
11 RADIDACTIVE WASTE MANAGEMENT The South Texas radioactive waste management system is designed to provide for the controlled handling and treatment of liquid, gaseous, and solid wastes.
The liquid waste system processes wastes from equipment and floor drains, sample westes, decontamination and laboratory wastes, and chemical wastes. The gaseous radioactive waste system provides holdup capacity to allow decay of short-lived noble gases from degassed primary coolant and treatment of ventilation exhausts through high efficiency particulate air (HEPA) filters and carbon adsorbers as necessary to reduce releases of radioactive materials to "as low as is reason-ably achievable" levels, in accordance with 10 CFR 20 and 10 CFR 50.34a. The solid waste management system collects and processes wet and dry waste generated by the plant and packages the material into .a solid product for shipment to a
- permanent disposal site. The radioactive waste management review includes pro-cess and effluent radiological monitoring and sampling systems.
The staff has reviewed the applicant's design criteria and design bases for the
. radioactive waste management systems in accordance with SRP Sections 11.1, 11.2, t 11.3, 11.4, and 11.5 (NUREG-0800). These acceptance criteria include the appli-cable GDC, 10 CFR 20.106, Appendix I to 10 CFR 50, and ANSI N13.1, " Guide to Sampling Airborne Radioactive Materials in Nuclear Facilities." Guidelines for implementation of the acceptance criteria are in the ANSI standards, regu-latory guides, and other documents identified in the SRP.' Conformance to the
) acceptance criteria provides the bases for the staff's concluding that the radioactive waste management systems meet the requirements of.10 CFR 20 and 50.
11.1 Liquid and Gaseous Effluent Source Terms The applicant calculated the estimated releases of radioactive materials in liquid and gaseous effluents using the PWR GALE Code described in NUREG-0017, Revision 0. The staff has reviewed ~these source terms and found them consistent with the guidelines of SRP Section 11.1. However, staff-calculated source terms using Revision 1 of NUREG-0017 were used in the staff's evaluation. The princi-pal parameters used in the staff's calculations are given in Table 11.1.
11.2 Liquid Radwaste Processing Systems 11.2.1 System Description
$.%.N The liquid waste processing system for Unit 1 is 4dentisa3 to that for Unit 2.
Each liquid radioactive waste processing system consists of process equipment and instrumentation necessary to collect, process, monitor, and recycle or dis-pose of radioactive liquid wastes. The liquid wastes from operation of Units 1 and 2 will be collected and processed in separate syst that consist of a l high purity system, a low purity system, a condensate ishing regenerant d system, and a laundry system. The high purity liquid ra aste system is de-signed to collect and process wastes from reactor coolant equipment drains, l containment normal sump, and other noncontaminated sources of reactor coolant.
The low purity liquid radwaste system is designed to collect and process floor draingwastefromthemechanicalauxiliarybuilding,aswellasotherpotential'y saA Q ~s $ ct M A &Flff South Texa SER 11-1
, . - - - - - - - .-._--.-...,..--.--.-.,_,w--- - - - - - _n-,-.e,---,, - . - , - - , - - . - - - - - - , . - - - - - - - , - , -
ATTACHMENT 1 ST HL AE- /M#
- PAGE 50OF 58 p55 contaminated waste. The condensate polishing reg erant and chemical radwaste subsystem receives high conductivity waste from the condensate domineralizer ;
regeneration subsystem, chemical laboratory drains, boron recycle evaporator condensate, and component cooling water sump pump. The laundry waste subsystem
, processes wastes from the laundry and shower drains and fuel handling building sump.
The liquid redwaste processing systems are designed to store processed water with the capability for eventual reuse in the plant. However, the design-basis valve alignment releases the processed water under controlled conditions to the environment. A radiation monitor in the plant discharge line will automatically terminateliquidwastedischag<i{*radiationmeasurementsgceedapredetermined level m l d H d m c f A **J
- h%% A ,
The high purity drain subsystem processes the low conductivity, high purity wastes. Waste is first collected in a waste holdup tank and is later processed through g mixed-bed domineralizer5(or an evaporator followed by a domineralizer) before'being stored in the outside waste monitor tanks or alternately, in the two waste evaporator condensate tanks. = 'l=: *-96 f.he processed water is either stored in the reactor water makeup storage tank Yor reuse or a por-tion can be discharged to the environment. The applicant estimates that the high purity drain subsystem will receive waste input flow of approximately 530 gpd, which is consistent with SRP guidelines. Adequate initial tank storage
,(10,000 gallons) is provided to accommodate flexible operation and the process flow rate (limited to 30 tym by evaporators) is sufficient to treat the daily inputs within an adequate time frame (less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per day).
The equipment and floor drain subsystem processes higher conductivity, low purity I wastes. Waste is initially collected in a 33,000 gallon tank, and later will I be processed through a' replaceable cartridge filter, an evaporator, a mixed-bed domineralizer, in series, before being collected in a waste management tank for sampling. The applicant estimates the equipment and floor drain subsystem waste input to be approximately 6900 tyd. The design process capacity of the floor drain subsystem is approximately 30 gpa (limited by the evaporator).
This flow rate is sufficient to process the daily inputs within an acceptable time (less than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per day). g g,g g g g,2 . 7, ,
The regenerant waste subsystem ill process resin regenerant from the condensate domineralizars and high conduct it waste from other radwaste area drains.
Regenerant waste (approximately 4 gpd) is first collected in the condensate polishing regeneration waste col ections tank. Then it is processed through a disposable filter, an evaporator package where water is separated out leaving a highly concentrated mixture of resin regenerant and solids in the evaporator bottoms. The concentrated waste in the evaporator is later sent to the solid waste processing system for solidification and disposal. The condensed water from the evaporator is then processed through a domineralizer, before it is temporarily stored in the wasta evaporator condensate tank. This water can then be recycled or discharged from the plant via the open cooling water system.
The laundry and hot shower waste processing systas first collects laundry waste and personnel decontamination drainage in a 10,000 ga11on hold tank. Then this waste is processed by filtration, temporarily stored in a waste monitor tank, and discharged from the plant via the open cooling water system.
ATTACHMENT .t_
. ST-HL AE-nom PAGE 510F Sg the quality group classifications used for system components, and the seismic 4 design applied to structures housing these systems and concludes that they meet the criteria as set forth in BTP ETSB 11-1 dated November 24, 1975, for radwaste systems and committed to by the applicant. The staff has reviewed the provi-sions incorporated in the applicant's design to control the release of radio-active materials in liquids as a result of inadvertent tank overflows and con-cludes that the measures proposed by the applicant are consistent with the cri-teria, as set forth in BTP ETS8 11-1. It is noted that RG 1.143, " Design -
i Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants" (Revision 1), replaced BTP ETSB 11-1. The staff concludes that the liquid waste system also meets the design guidance of RG 1.143.
11.3 Gaseous Radwaste Processing System 11.3.1 System Description i The gaseous radioactive waste processing and plant ventilation systems are l designed to collect, delay, process, monitor, and discharge potentially radio-active gaseous wastes that are generated during normal operation of the plant.
The systems consist of equipment and instrumentation necessary to reduce releases of radioactive gases and particulates to the environment. The principal sources of gaseous waste are the effluents from the gaseous waste processing system, condenser mechanical vacuum pumps, and ventilation ext.austs from the reactor building, auxiliary building, and fuel handling building.
The gaseous waste processing system (GWPS) is a charcoal delay system that i
! receives vent gases from the volume control tank purge, boron recycle holdup l
tank, boron recycle evaporator, liquid wasta evaporator, reactor coolant drain l tank, and pressurizer relief tank. In addition, for rapid degassing of the reactor coolant before refueling, a vacuum degassing compressed storage tank system is used for radioactive gas holdup. This gas is later released either through the charcoal delay system or through a bypass of the charcoal system.
The gas discharge from the charcoal delay system and gas decay tanks passes through a HEPA filter and is discharged to the environment thrcugh the mechani-cal auxiliary building vent.
The reactor building ventilation exhaust system discharges air (40,000 ft 8/ min) from the reactor containment to the atmosphere without filtration during refueling. Before refueling, the containment atmosphere is recirculated for 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> through the containment cleanup system for airborne radioactivity re-moval. This system consists, in order, of a redundant roughing filter, a HEPA filter, a charcoal filter, and a HEPA filter, followed by two circulation fans.
Gases released from liquid radwaste system equipment vents are directed through the plant vent header without filtration. Exhaust flow from the refueling floor of the FM is normally unfiltered unless activity is detected in the refueling floor erhaast duct If the latter is the case, then the refueling floor exhaust flow is diverted through redundant HEPA/ charcoal filtration units before it is released to the atmorphere. These filtration units are di ussed in Sec-tion 6.5.4.2 of this SER. f 2
The turbine building ventilation exhaust is unfiltered. The main condenser i mechanical vacuum pump discharge is unfiltered.
ATTACHMENT ,L.
> . ST.HL AE /MI/
PAGE 5AOF 58 f
indicate radioactive leakage betwean systems, monitor equipment performance, I and monitor and control radioactivity luels in plant discharges to the environs.
Table 11.3 provides the proposed locations of continuous monitors. Monitors on certain effluent release lines will automatically terminate discharge if radiation levels exceed a predetermined value. Systems that are not amenable to continuous monitoring, or for which detailed isotopic analyses are required, will be periodically sampled, and the samples will be analyzed in the plant laboratory.
All nomal airborne radioactive releases to the environs from each South Texas unit are from the plant main exhaust duct at the roof of the mechanical auxil-iary building. These gaseous releases are considered ground level releases for establishing the dispersion coefficient.
Redundant radiatinn_maniters ars W+Shd in the main exhauct. vent 3 They sam-ple effluent air down tream of the last point of mixing and provide continuous i
pg readout of noble gasD nT particulate activity. In addition, charcoal adsorbers
. Sampling flow is through areprovidedtosampletheairstreamforradiciodin{54ML4 a regulated isokinetic probe.
In addition to the main plant vent discharge point monitors, ventilation radia-
- tion monitors are also provided in the exhaust streams of the reactor contain-ment and fuel handling buildings. These monitors are used tg detect fuel han-dling accidents and tofi-tf fy cemi ef
- f t- rt!"f tAefore mixing and dilutioninthemainQ1antexhaustvent.ac. L &e W o cA sf g y . p M -
- The principal radioactive liquid effluent release point for each South Texas unit is the discharge of the circulating water system. Potentially radioactive inputs to the circulating water system will be continuously monitored for radio-activity before and during release. Because the circulating water is a high flow rate system, the radioactive inputs to the system will be highly diluted.
The liquid radwaste discharge process monitor providas alare signals for termi-nation of discharge of the liquid radwaste system in the event a predetermined Technical Specification limit is ,xceeded.
e Process liquid radiation monitors are also used in the closed-loop component cooling water system to detect any heat exchanger leakage of primary coolant.
11.5.2 Evaluation Findings The staff review included the locations and types of effluent and process moni-toring provided for each South Texas unit. On the basis of the plant design and the continuous monitoring and intermittent sampling locations, the staff has concluded that all normal and potential release pathways will be monitored.
The applicant's description indicates that the process and effluent monitoring system design meets RG 4.15, " Quality Assurance for Radiological Monitoring Progras.s." The staff also determined that the sampling and monitoring provi-
'sions are adequate for detecting radioactive material leakage to normally un-contaminated systems and for monitoring plant processes that could affect radio-activity releases. On these bases, the staff considers that the monitoring and sampling provisions meet GDC 60, 63, and 64 and RG 1.21, " Measuring, Evaluating and Reporting Radioactivity in Solid Waste and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants" South Texas SER 11-8
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c Source: I Provided at meeting between NRC and app cant, Bethesda, Maryland, April 2,1985
- d
ATTACHMENT A ST HL AE /04
- PAGE5V OFS8 Table 1 Stability analyses results--station 20+00 Computed factor Case of safety Steady seepage Upstream 1.82 Downstream 1.72 Rapid drawdown 1.50 (elevation 49 to .
39 feet)
Pseudostatic (seismic) 1.30 Table 2 Main coolins reservoir instrumentation--type, number, and frequency of monitoring Frequency of monitoring After filling
- During Type Number filling 0-1 year 1-%5 years After 5 years l M Monthly Bimonthly Quarterly Semiannually Benchmarks and settlement plates Piezometers Biweekly Monthly Bimonthly Quarterly Relief wells Biweekly Monthly Bimonthly Quarterly Inclinometers 12 i Biweekly Monthly Bimonthly Quarterly l
I k l
t l
l i
l f
12 Gulf C:: t growth "fcult" t rminalegy, o tcra indiccting that tha tyt,hrown cida cf a growth "fcult" 10 toward tha eccet. ATTACHMENT CLs,
- ST HL AE- 1674
- 41. Unconsolidated-Undrained (UU) PAGE S60F Ss l
A triaxial shear test (unconsolidated-undrained) in which a soil sample is stressed by applying an axial load while preventing pore water drainage. 35 e M %.E.I Definitions A glossary of significant geologic and geotechnical terms is included at the end of Section 2.5. From Sections 2.5.1 through 2.5.6, certain terms are asea ror defining specific areas of reference. These terms and their definitions ares t
- 1. Power station area. Immediate area of Unit 1 and Unit 2 plant site Category I structures and Turbine-Generatory Building; about 950 ft by 600 )
ft in size l l
- 2. ECP, ECP area The area northeast of the plant site for i the Essential Cooling Pond; about 46.5 acres
- 3. Site The area within the legal ownership boundary
- 4. Area, site area The area within 5 miles of the plant i site l 2.5.1-13 Amendment 35 i
......m
- 5. Vicinity, site vicinity The area within 15 miles of the plant site -
- 6. Region, site region The area within 200 miles of the plant site Whenever the acronym PSAR is used as part of a table or figure reference, it is referring to an existing figure or table that has remained unchanged and is in the STP PSAR (Ref. 2.5.1-83).
2.5.1.1 Reaiensi Geolony.
2.5.1.1.1 Physiography: The principal physiographic features of Texas are shown on PSAR Figure 2.5.1-1. More than one-third of the southeastern part of the state is within the Texas Gulf Plain, a broad arcuate lowland belt, convex to the northwest. The STP is located in Matagorda County on the Gulf Coast portion of the Texas Gulf Plain adjacent to the Gulf of Mexico (Ref. 2.5.1-8). The inner margin of the belt, trending east 150 miles from Del Rio on the Rio Grande and thence north-northeast some 300 miles to the Red River, is marked throughout much of its course by the Balcones escarpment and foothills of the comparatively higher plains and plateau to the northwest.
The single most distinct physiographic feature in the inland part of the
_ state is the Stake _d Plains, otherwise known as the High Plains. These
ATTACHMENT
, ST.HL AE-lM4 PAGE 560F 38
- The Light NE f Houston Lighting & Power P.O. Box 1700 Houston. Texas 77001 (713)228-9211 t December 16, 1985 ST-R.-AE-1532 File No.: G9.17 l l
Mr. Vincent S. Noonan, Project Director PWR Project Directorate #5 U. S. Nuclear Regulatory Commission Washington, DC 20555 South Texas Project Units 1 and 2 i Docket Nos. STN 50-498, STN 50-499 Response to DSER Open Item Reoardino Flood Protection
Dear Mr. Noonan:
The attachment enclosed provides STP's response to Draft Safety Evaluation Report (DSER) items.
The item number listed below corresponds to those assigned on STP's internal list of items for completion which includes open and confirmatory DSER items, STP FSAR open items and open MC questions. This list was given to
,your Mr. N. Prasad Kadambi on October 8, 1985 by our Mr. M. E. Powell.
The item which is attached to this letter is:
Attachment Item No.* Subject 1 D 2.4-5 Section 2.4.14 Flood Protection-A response to this item was previously submitted to your office on Detober 31, 1985 (ST-HL-AE-1474). This response revises the previous response.
- Legend
- D - DSER Open Item C - DSER Confirmatory item l F - FSAR Open Item Q - FSAR Question Response Item 4
i Ll/NRC/nn
Atttchment
- ST-HL-AE-1532 Filo No.: G9.17 1 Ny l of 1 l l
ATTACHMENT L
. ST HL AE IV/d ,
PAGE 570F 58 l DSER Open Item 2.4-5 Provide a procedure to require that waterproof doors and waterproof knockout panels be in a secured position under normal conditions. l Revised Response: ,.,
Reference #1 - ST-H.-AE-1103; Final Report Concerning Doors Used for Flood Protection; dated June 11, 1984
- 2 - ST-HL-AE-1418; Responses to DSER/FSAR Items FSAR Sections 2.4, 3.4 and 3.5; dated October 12, 1985
- 3 - ST-HL-AE-1293; Response to Main Cooling Reservoir Meeting Action Items; dated 7/1/85 The reference 1 letter as well as Question & Response 240.04N, outlines the project position regarding watertight doors: "The provision of single waterproof doors, which can be secured if conditions require, provides
, adequate protection against any credible flood". Also, the reference 2 letter provides the revision to FSAR Section 3.4 which documents this position. This position is further clarified for the following flood events:
- 1. MCR Breach - This flood event will not dictate the watertight doors to be normally closed, except in the unlikely event anomalies with
> the MCR embankment are discovered which require remedial action procedures to be implemented. Should this situation arise, as an
, additional precautionary measure adninistrative controls will be implemented to keep the watertight doors normally closed and i knockout panels on the DEAB in place. This will be further l
addressed in the MCR Operating Procedures as discussed in the reference 3 letter. Additionally, at the time such remedial action procedures are being implemented, only one Diesel Generator Building knockout panel may be removed at one time. ,
- 2. Other flood events - STP will have sufficient warning time prior to being inundated such that closure of the doors and replacement of knockout panels can be accomplished.
Ll/PRC/nn
ATTACHMENT 4 ST-HL- AE /4")</
o' **. PAGE 570F 5g Houston Lighting & Power Company a ST-HL-AE-1532 File No.: G9.17 Page 2 If you should have any questions on this teatter, please contact Mr. M. E. Powell at (713) 993-1328.
Very truly you(s.
S'l ,
M. R. Wisenburg Manager, Nuclear Lie sing RLE/yd
Attachment:
See above k
(
Ll/NRC/nn