ML18040B018
| ML18040B018 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 08/19/1983 |
| From: | Curtis N PENNSYLVANIA POWER & LIGHT CO. |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0694, RTR-NUREG-0737, RTR-NUREG-694, RTR-NUREG-737 PLA-1785, NUDOCS 8308230292 | |
| Download: ML18040B018 (211) | |
Text
RESULATORYQ'IFORMATION DI'STRIBUTION 'SEM (RIOR) qe~
)'AOCES8 ION NBR 0 8308230292 DOC 0 DATE>> 83/08/19 NOTARIZED t .NO DOCKET,'
'FAGIC',:50 388 "Susqueh'anna 'Steam:Electr ic Stations 'Uni;t 2i Pennsyl va 05000388
'AUTH SHAME AUTHOR AFFILIATION
,CURTISiN AN< tPennsyl.vania.'Power 8, Light tCo<
~REC I'P ~ NAME RECIPIENT AF F IL'I ATION
'SCHNENCEREA ~ Licensing Branch '2 unreflecting
'SUBJECTS For.wards 1nev ito, SAR C apter. 18 r esponses,to 'TMI items 'Chapter will be 'included =in "next 'FSAR amend, DISTRIBUTION lCODE: B001'0 iCOPIES )RECEIVED:LTR, "ENCL
'SIZE',
Licensing "Submittal: PSAR/FSAR Amdts .8 >>Related "Co'rnespondence'
.'ITLE:
NOTES!icy NMSS/FCAF/PM'L'PDR 2cys, Lpmi DQ sk ECIP IENT iCOP IES RECIPIENT ~COP IES ID 'CODE/NAME LTTR'ENCL'D "CODE/NAME LiTTR,ENCL NRR/DL/ADL 1 "0 NRR L82 BC ,1 0 NRR L82 LA 1 0 <PERCHiRB 01 1 1~
INTERNAL: ELD/HOSE 1 IE FILE 1 IE/DEPER/EPB >36" 13 IE/DEPER/IRB 35 1 IE/DEQA/GAB '21- 1 NRR/DE/AEAB 1
-NRR/DE/CEB 11 NRR/DE/EHEB 1
.NRR/DE/EQB 12 2 NRR/DE/GB 28 '2 NRR/DE/MEB 18 NRR/DE/MTEB 17 1 NRR/DE/SAB '24 1 NRR/DE/SGEB 25 1 NRR/DHFS/HFEBOO NRR/DHFS/LQB 1 NRR/DHFS/PSRB 1 32'RR/DL/SSPB 1
NRR/DS I/AEB 26 1 NRR/DSI/ASB 1 NRR/DSI/CP8 10 1 NRR/DS I/CSB 09 1 NRR/DSI/ILOSB 16 1 NRR/DSI/METB 12 1 NRR/DS I/PSB 19 1 AB "22 1 NRR/DSI/RSB 1 REG FILE 04 1 1
'23'RGN1 i /MIB 1 0 EXT'ERNAL ~ ACRS 41 BNL(AMDTS ONLY)
.DMB/DSS (AMDTS) FEMA REP DIY 39 1
>TOTAL NUMBER OF COPIES REQUI'RED: L'TTR 57 ENCL
Pennsylvania Power 8 Light Company Two North Ninth Street ~ Allentown, PA 18101 ~ 215/ 770.5151 Norman W. Curtis Vice President-Engineering 6 Construction-Nuclear 21 5/770-7501 PUG ] 9 1983 Director of Nuclear Reactor Regulation Attention: Mr. A. Schwencer, Chief Licensing Branch No. 2 Division of Licensing U.S. Nuclear Regulatory Cormission Washington, D.C. 20555 SUSQUEHANNA STEM EUXTRLC STATION Z4I RESPONSE FOR UNIT 2 ER 100508 FILE 841-1 PZA-1785 Docket No. 50-388
Dear Mr. Schwencer:
Attached are revised pages to the Susquehanna SES FSAR Chapter 18. These pages reflect the responses to the KMI Items for Unit 2. Also included are administrative changes to update the TMI Items for Unit 1. Where no specific mention of a particular unit is made, the response is applicable to both units. It should also be noted that unless otherwise specified the equixment and modifications will be installed and operational prior to fuel load on Unit 2.
The following is an explanation of the changes:
18.0 This section has been revised to reflect the fact thag the responses in Chapter 18 are applicable to both units.. was It also revised to dlarify equipment identification numbers for Unit 2.
18.1.1.3 This section has been revised to incorporate the Nuclear Training Instruction for the qualification of the STA's.
was also revised to state that the STA's have been trained It and are on shift.
18.1.4.3 This section has been revised to incorporate the reference to NTI-QK-3005. It was revised to state that the initial startup crews have been trained.
18.1.7.3 This section has been revised to state that the Nuclear Safety Assessment Group (NSAG) is functional.
18.1.8.3 This section was revised to state that the eaergency procedures have been developed using the BWR Emergency Procedure Guidelines.
Boor 8308230292. 830819 '., I.;,'., I /it
'PDR ADOCK 05000388
.. "'PDR 4-/NI7LP &57,
Page 2 SSES PLA-1785 ER 100508 File 841-1 Mr. A. Schwencer 18.1.12.3 This section has been revised to state that Administrative Procedure AD-QA-406 has been developed.
18.1.13.3 This section has been revised to state that Administrative Procedure AD-~-306 has been developed.
18.1.16.3 This section has been revised to state that the mxlifications that were r~ed to be implemented prior to fuel load on Unit 1 will be inplemented on Unit 2 prior to fuel load. It was also revised to reference our response to Generic Letter 82-33.
18.1.17.3 This section has been revised to reference our response to Generic Letter 82-33 which states our position on SPDS.
18.1.19.3 This section contained typographical errors which were corrected.
18.1.21.3.1 This section has been revised to delete the first paragraph of this section. This paragraph was deleted since contained a caranitment to install a sampling system and it it has been installed.
18.1.21.3.2 In this section, the reference to Figure 18.1-11 was revised to reference Figure 9.3-9a.
18.1.21.3.2.1 This section was revised to state that all sample points have been installed.
18.1.21.3.2.2 In this section, the reference to Figure 18.1-11 was revised to reference Figure 9.3-9a.
18.1.21.3.2.3 This section has been revised to state that the piping station has been installed. Also the reference to Figure 18.1-12 was revised to reference Figure 1.2-20.
18.1.21.3.2.4 In this section, the reference to Figure 18.1-13 has been revised to reference Figure 1.2-4.
18.1.21.3.2.4.3- This section has been revised to state that the sample station enclosure has been vented.
18.1.21.3.3 This section has been revised to state that the sample facilities have been provided.
18.1.21.3.3.1 This section has been revised to state that the on-site facilities have been provided.
Page 3 SSES PLA-1785 ER 100508 File 841-1 Mr. A. Schwencer 18.1.21.3.3.2 This section has been revised to state that the EOF facilities have been provided.
18.1.21.3.3.3 This section has been revised to state that arrangements have been made for off-site analyses.
18.1.21.3.4.2.3- In this section, the reference to Figure 18.1-14 has been revised to reference Figure 18.1-11. Also the reference to Figure 18.1.21-4 has been revised to reference Figure 18.1-11.
18.1.21.3.4.5 This section was added to state that the procedure for core damage estimation will be completed prior to the startup following the first refueling outage of Unit l.
18.1.22.3 ,This section has been revised to state that the "Mitigating Core Damage" course has been developed and is available.
18.1.24.3 This section has been revised to state that the acoustic monitoring system has been installed.
18.1.29.3 This section has been revised to state that the valves in the Liquid Radwaste, Reactor Water Sample and Reactor Building Chilled Water systems have been edified so that they do not automatically reopen on logic reset. Also this section has been revised to state that the purge and vent valves for Unit 2 will be fully qualified to Branch Technical Position CSB6-4 prior to fuel load and for Unit 1 prior to startup following the first refueling outage. This section has also been revised to state that the high radiation monitors have been installed and are operational.
18.1.30.3.1 This section has been revised to state that the noble gas monitors have been installed.
18.1.30.3.2 This section has been revised to clarify from what sources the radiation exposures were calculated. In a recent reevaluation, PPGL has determined that access to these monitors is not possible if airborne contamination is taken into account. The previous access study was based on the shielding study (Item II.B.2) which did not include airborne contamination. As a result of this reevaluation, the filters will be moved to a=location which is accessible. This relocation will be campleted prior to the start-up following the first refueling outage of Unit 1.
18.1.30.3.3 This section has been revised to state that the containnent t
high-range radiation nnnitor has been installed.
Page 4 SSES PZA-1785 ER 100508 File 841-1 Mr. A. Schwencer This section has been revised to state that the containment pressure monitor has been installed.
This section has been revised to state that the containment water level rmnitor has been installed.
This section has been revised to state that the containment hydrogen monitor has been installed.
This section has been revised to update this response to the present BWR Owners Group position.
This section has been revised to state that the time delay relay has been installed.
This section has been revised to state that the ADS logic modification selected has been approved and that additional information is needed.
This section has been revised to state that the autcmatic switchover of the RCIC suction frcm the condensate storage tank (CST) to the- suppression pool on low CST level has been installed.
This section has been revised to state that the cooling water to the recirculation pump seals have been modified to receive emergency power for Unit 1. This nedification will be made to Unit 2 in order to keep the tw'o units the same even though this modification is not ray@.red to be made on Unit 2. The Unit 2 mxLification will not be made prior to fuel load.
This section has been revised to state that all reactor water level indications use the same reference point.
This section has been revised to eliminate the reference to Appendix I of the Bnergency Plan since Appendix I has been eliminated fram the Emergency Plan.
This section has been revised to state that the response to the requirements are contained in the Em rgency Plan.
Item 4 in this section has been revised to state that the Technical Specifications include the 5 gpn criteria.
This section has been revised to delete the last line of this section.
This section has been revised to reference Section 2.2 for the evaluation of potential hazards frcm nearby facilities.
Also this section adds a reference to Table 18.1-17 which
Page 5 SSES PLA-1785 ER 100508 File 841-1 Mr. A. Schwencer
\
references information required for the NRC control roam habitability evaluation.
Table 18.1-10 This table has been revised to correct typographical errors.
Table 18.1-17 This table was added to provide references to applicable sections of the FSAR, Brergency Plan and Technical Specifications for use in the control roam habitability evaluation.
Table 18.1-18 This table was added to provide a listing of possible ICC detection devices.
Figure 18.1-11 This figure titled "PASS PAID" has been deleted.
Figure 18.1-12 This figure titled "Location of PASS El 719'-1" " has been deleted.
Figure 18.1 This figure titled "Location of PASS El 729'" has been deleted.
Figure 18.1-14 This figure titled "Specific Conductance of pH of Aceous Solutions at 25 C" has been renumbered as Figure 18.1-11.
Figure 18.1-15 This figure titled "Typical HPCI/RCIC Steamline Break Detection Logic" has been renumbered as Figure 18.1-12.
Figure 18.1-16 This figure titled "Typical Reactor Water Level Display" has been renumbered as Figure 18.1-13.
New Figure 18.1-14 titled "Downcmer Water Level History" has been added.
New Figure 18.1-15 titled "Water Level As An Indicator of Core Overheating" has been added.
New Figure 18.1-16 titled "Cladding Tenperature Sensitivity to Core Uncovery Time" has been added.
18.2.2.3 This section has been revised to state that the Vice President-Nuclear Operations reviews and approves the Shift Supexvisor's responsibilities.
18.2.9.3 This section has been revised to state that the Senior Vice President-Nuclear has issued a statement of policy establishing the primary responsibility of the Shift Supervisor for safe operation.
18.2.12.3 - This section has been revised to state that General Electric Conpany has reviewed all systems for, Unit 1.
s~p tests associated with NSSS
Page 6 SSES PIA-1785 ER 100508 File 841-1 Mr. A. Schwencer 18.2.15.3 This section has been revised to state that a safety analysis on station blackout has been submitted and that Generic Letter 83-24 has been issued.
18.2.33.3 This section has been revised to state that the Technical Specifications include these reporting requirerrents.
18.2.38.3 This section has been revised to delete the reference to Appendix I of the Emergency Plan since this appendix does not exist.
Table 18.2-1 This table has been revised to delete the canmitment for writing a station blackout test procedure.
These changes to Chapter 18 will be included in the next revision to the FSAR.
If you have any questions, please call.
Very truly yours, N. W. Curtis Vice President-Engineering 6 Construction-Nuclear cc: R. L. Perch-NRC G. G. Rhoads-NRC
SS"=S-FSAi".
Th' chanter contains a response for each T'?I-related
<~q rom~nt, hc chapter j s dividc~d j nto sect j ons wh jch con~a in u'he espon'es to all =eauiremen".s fo" applicants fo= operatina l jcen ses. The ".ablo of contents ident fies which section pro vides the resoonses or a q'~n doc>>ment.
Each sec'.ion ad.'resses all the equ'ements in i'..s correspondina document. A esponse is only qivrn =o .h- most recent 'n the series of "equi ements which con+ains an exp'ana-ory text. For example, if. an explanatory text of requirement I.A.l.l appears on both NUTMEG 0737 and NUPEG 0694, a response 's provided .o NUTMEG 0737 since it. supersedes all previous requirements. If requirement I.A.1.2 appears in both NURFGs 0737 and 0694, but the only explanatory tex;. is ir. NtJBFG 0694, the response is provided to NURFG 0694 utilizinq the implementa.ion dates of !JUBEG 0737.
These responses are aoplicable to both Unit 1 and Unit 2, howeve , +he equipment identificat'on numbe s must be corrected by eplacinq a Unit desiqnator with a JJnit 2 designator. For 1
example, valve HV-15713 in Unit corresponds wi".h HV-25713 in 1
Unit 2, control panel 1C601 corresponds wi' panel 2C601 in Unit 2.
- 18. 0-1
-8308230292
- 18. 1 RESPONSE TO PF.')UIRHNHNTS IN NUR:"G 0737 18,1.1 SFI f> TgCHNICAL N DVISOR g:. A. 1. 1$
- 18. 1.1.1 Statement of. Pegui=emert
":ach licensee shall provide an on-shif",. '. echnical advisor to the shift supervisor. <'e shift technical advisor (<T~) may serve more than one unit at a multiuniz si".o.
the adviso function for the various uncs.
if quali fied:o perform The STA shall have a bachelo-'s deqree or equivalent in a scien+ific or enaineerinq discipline and have eceived spec'fic traininq in the response and analys's of the plant for tran. ients and accidents. The STA shall also receive ~raining in plant desian and layout, includinq the capabilities of instrumentation and cont:ols in the con.rol room. The licensee shall assign normal duties to the STAs tha per..ain to the enqinee ing aspects of assur'nq safe ope"ations of the plant, includinq +he review and evaluation of. operatinq experience.
The need for the STA po it'n may be elim'ated when the qualifications of the shift superv'sors and senior ope=a:ors have been upgraded and the man-machine interface in the con" rol room has been acceptably upqraded. Howeve=, until those long-term improvemen+s are attained, the need f or an O'ZA proqram will continue.
The staff has not yet established the do+ailed element-. of the academic and traininq requirements of the S=A beyond the guidance q'ven in the Vassallo let+er on November 9, 1979. Nor has rhe staff mac'e a decision on the level of uoqradinq -equi"ed for licensed operatinq personnel and the man-machine interface in .,he control oom that would be acceptable for eliminating,.he need of an STA. Until these requirements fo" eliminat'ng the STA pos'tion have been established, =he staff continues to =equire that, in addition to the staffing reau'remend specified in Subsec.ion 10.1.3, an STA be available fo" duty cn =ach operate ng shift when a plant is beinq operated in Nodes 1-3 for a BVR. A.
o her time,, an STA 's not =equired to be on duty.
Since the November 9, 1979 letter was issued, several efforts have been made o establish, for the longer +erm, the minimum level of experience, education, and training fo" STA.. These efforts include work on the rev sion to ANS-3.1, work by the Institute of Nuclear Power Operations (INPO), and in ernal staff efforts.
- 18. 1-1
SSES-PSALM INPO has made a vailab'e a document ent'ied>>Nuclear Powe" Plant Sh'" Technical Advisor--Recommerda ions for Position Description. Ou alifications, Fducation and Traininq.>> Sect'. ons 5 and 6 of the IN PO document describe the education, trai ning, ar.d experience reau i ements for STAs. ;h. N".-.C s-'.aff finds tha: the de .rriptior.s as se+ forth in Seczions 5 and 6 of fcevis'r. 0 to th=. IHPO docume nt a"e an acceptable aoproach for the selection and +raining of person'nel to s a f :h~ STA positions. (Note:
This should. not be interpret t~d to mean ~hat this is an NHC "equirement at this time. The 'nier.t is to "efer to +he IHPO document as acc eptable for interim qu'dance fo" a utili+y in planninq its ST A program ove" +he long term (i.e., beyond the January 1, 1981 reauirement o have STAs in place in accordance with the qualif ication requirements specified 'n the staf 's November 9, 197 9 lett.er) .
Appl can+s for operating licen=es shall prcvide a description of their STA trainina and requalification program in their application, or amendments the"eto, on a schedule consistent wi..h the NRC licensina review schedule.
ADplicants for operatinq 1'censes shall provide a description of
+heir iona-term STA program, includinq aualifica-'on, selec" ion criteria, traininq, ar.d possible phaseout. The description shall be provided in the aoplicat'ion, or amendments the"eto, on a schedule consistent with the NBC licensinq review schedule. The descr'ption =-hall include a comparison of the long-term p" oqzam wit.h the above mentior.ed INPO document.
- 18. 1. 1. 2 Interpretation The appl.can= i s +o develop a tzaininq proqram in compliance with
",.he Novemb er 9, 1979 le.ter and submit. a descrip+icn to the NRC.
The applica nt i s to provide STA cove"aqe for all operatinq shif ts. Candid ate will complete a traininq p"ogram and pass a cert.ifica ion e xamination prior to assumption of duties. The appl:cant is o develop a long-term p ogram -;o maintain or Dhaseout STAs.
- 18. 1.1. 3 S+a+ement of Response The. program tor the . election and trairinq of STA's is detailed in rTI-QA-3030, >>Initial Traininq and Certif'cation of Shift Technical Advisors".
STA coveraqe is provided on opera..inq shift"=. in accordance with Subsection 6.2. 2 of the Technical Specifications. STA' perform
- 18. 1-2
SSES-FSAH
..he duties'nd have the responsib'ities ou-lir.ed in plan.
orocedure AD-OA-40. "Cor.duct of Technical Support.."
STAs meet: the auali f ica+ ion =equiremen s of th~ Vassallo le ..te of Yovemhe r 9, 1979 All STA train'nq is completed and STAs a=e ready fo" shift ass iqr.men-.. The STA proaram described above will be maintained lonq- erm un;.il such time as phaseout is permit-.ed ir. accorda nce wi"h NRC ins ructions.
- 18. 1. 2 SHI>~ . UPERVISOR BESPOiVSIBII,ITIES gI A.l. 2g No requirement stated in NUREG 0737. I~ef er to Suhsectior. 18.2. 2 which contains the response to the requirement s ared in NUREG 0694.
18.1.3 SHIFT HAJJVIHr, rI. A. l. 3l 18, 1,3,1 Statement of Reguiremen=
Applicants for operatinq licenses shall include in thei administrative procedures (required hy licer.se conditio ns) provisions qoverninq required shif t s+a ff inq and moveme n of key individuals about the plant. These provisions are roqu ired to assure ti a qualified plant personnel to man the operat iona 1 shifts are readily available in the even,. of an abnorma 1 or emerqency situation. Interim requi"ements for sh'ft s- affinq are qiven in Table 18.1-1.
These administ at'e procedure: shall also set forth a policy, the objective of which i" to prevent situations whe -e fatigue could reduce tho ability of operatinq personnel to keep +he reactor in a safe condition. The. controls established should assu e that, to the ex ent practicable, personnel are not assiqned to hift. du+ies while in a fa-.'.iques condi ion that could siqnificantly reduce their mental alertness or their decision makinq ability. The controls =hall apply +o the plant s aff who perform safety-related functions (e.q., senior reactor ooerators, reactor opera ors, auxiliary operators, heal-h physicists, and key maintenance personnel).
IE Circular No. 80-02, "nuclear Powe Plant Sta f f Mork Hour da+ed Fehr>>ary 1, 1980 discusses the concern of overtime wo=k fo" members of the plan+ sta ff who perf orm sa f ety-r elated functions.
>he quidance co ntained in he IE Circular No. 8 0-02 was am en ded by the July 31, 1980 letter.. In turn, the over t me guidance of the July 31, 19 80 letter was rev'sed in Section I. A. 1. 3 of. WtJB,";6-0737. The HRC has issued a pol icy st a-.,ement wh ich f urt h er
- 18. 1-3
revises the overtime quidarce as s+ated in H(JR "G-0737. This auidance is as follows:
Enough nlant opera.ina personne 1 should be employed to maintain adecuate shif~, covoraqe withou, rout'e heavy use of overtime.
The object'e is to have operat inq ' personnel work a no" mal 8-hou" day, 40-hour week while the pla nt operatina. !however, in the event that unforeseen problems reau'e su bs" ..n+ ial amounts of overtime to be u ed, or. dur'ng extended period s of shutdown fo" refuelina, m" d'or maintenance o" majo plant modificat'ons, on a
.em porary basis, the followina cuidel ines si all he followed:
(a) An individual should not be permit ed o work mor.. than 16 hours straiqht (excluding sh.'= turnover time) .
(b) An individual should not be permitted to work more +han 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24-hour period, no more than 24 hou"s in anv 48-hour period, no mor~ than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any sever. day period (all excluding shif" urnover time) .
(c) A break of at least eiqht hours should be allowed bew+een work periods (including shift turnover time) .
fd) Except during extended shutdown periods, the use of over ime should be considered on an individual basis and not for the entire s aff on shi
'.?ecoqnizinq that very unusual circumstances may arise requirinq devia+ion from the above auidelines, such deviation shall be autho ized by the plant manager or his depu y, or higher levels of manage,ment. Ti'e paramount consideration in such au+horization shall be that siqnificant reductiors ir. the effectiveness of operatina personnel would be h'qhly urlikely. Authorized deviations to the wo"king hour quidelines shall be documen.ed and available for NRC review.
In addition, procedures are encouraqed that would allow licensed operators at the cortrols to be pe iodically relieved and assiqned to other duties away from the control hoard during their tours of d>>ty.
Operatinq license applicants shall comple e these administrative procedures before fuel loadinq.
18.1.3.2 Interpretation None required.
- 18. 1-4
SSFS-FSAB
- 18. l. 3, 3 Staiemer+ of Resoorse he fac'li+ y s-a f f inq "equi remen ts are ores~n".ed in Suh ~ct.ion 6 2-2 of +h e Technical Specification=. These requi.emen "s are cons ster t w'+h those given in .able 18.1-1.
The plant oolicy on operat'ors personnel working hou=s is d iscussed in administra+ive procedu=e AD-QA-300, "Conduct of Ooe"ations."
18.1.4 IN.",EDIATF. UPGBADTl'.G OF P.F.AC.OB OP BATOB AND ."F.t~IOB BF..ACTOB OP~PAVOB TBAZ~IVC, AND gUAiIFICATIOVS gr.A.2.li
- 18. 1.4,1 Statement of. Beaui emen-Aoplicants~ for enior operator 1'censes shall have 4 years of esponsible power plant. experience. Bespons'ble power plant exoerience should be that obtained as a cortrol oom operator (fossil or nuclear) or as a power plan+ s aff e..qineer involved in the day-to-dav activities of the facilitv, commencina with the inal year of construction. A maximum of. 2 years power plant experience may be ful illed hy academic or related technical trainina, on a one-for-one time basis. Two years shall be nuclear power plant experience. A leas" 6 months of the nuclear power plan, experience shall be at the plant for which he seeks a license. Ef fective date: Application= =eceived on or af ter Na y 1, 1980.
Aoplican.s for senior operator licenses sha11 h ave held an opera.or's license for 1 year. Fffective Dave:: Applica iors received a f ter December 1, 1980. The NEC has r; ot imposed the 1-year expe"ience requirement on cold apolican s fo" SPO licenses.
Cold aoolicar ts are +o work or. a facili? y not v et in operation; their spraining programs are des'qned to supply +he equivalent of the expe ience not available to them.
Senior operator~: Applicants =hall have 3 months of shift training as an ex ra man on shif Con. ol room operator~: Applicant" shall have 3 months training on shift as an extra person in the control "oom. Kffective date:
Applications received after Auqust 1, 1980.
~Precritical applicants will be requi ed to meet unique aualifica ions de..igned to accommodate the fact thar. their facility has not yet been in opera ion.
- 18. 1-5
SSES-FSAP
<zaininq p"oazam.. shall b~ mod fied, as necessary, +o Drovide:
- 1) raininq in heat t=ansfe, fluid Flow and
+.hermodynamics.
- 2) Tzaininq in the use ot installecl plan+ sys:ems to control or mitiqate an accident 'r. which ",) e core is severely damaqed.
- 3) Inc eased s on reacto= and plan'. transien+s.
fective date: present proqrams have be~n modified in emphas'.
re. ponse ...o Bulle ins and Orciezs. Revised proqrams should be submit. ed for OLB review by Auqust 1, 1980.
Content of the licensed operator zequali 'ca.ion p"oqrams shall be modified to include instruction, in heat transfer, fluid flow, thermodynamics, and mit'qation of accidents involvina a deqraded core. E fective date: .'lay 1, 1980.
The critezia for zequiz'nq a licensed individual to pa"ticipate in accel crated reaualification shall be modified to be consistent with the new passinq qrade foz issuance of a license; 80% overall and 70% each cateqory. Effective da+e: Concurrent with the next facilitv administered anrual zequalifica~ion.examination after the issue date of. this requirement.
Proq=ams should be mod'ed to require,he con rol mani p>>lat ions listed in Enclosure 4 of HUBEG 0737, -'tern I h.2.1. Nor mal cor trol manipulatiors, such as plan+ reac+or staztup-, must be performed. Control manio>>lations durinq abnormal or em e-qency operations must be walked ~hrouqh with, and evaluated b yi member of the tzaininq staf f a.. a m'nimum. hn appropri ate
.imulator may be used to satisfy the reauiremen+s for c ontrol manipula ions. Effective da~e: Proqrams modified by il uqust 1, 1980. Renewal application received af ter November 1, 1980 mus" reflect compliance with the proqram.
Cert ificat ions completed pursua nt to Sections 55. 10 (a) (6) and
- 55. 33a (4) and (5) of 10 CFR Part 55 shall be siqred by "he hiqhest level of corporate manaqement for plan+ operation (for example, Vice President for Operation.".) . Effective date:
Appl.icatiors received on or after Hay 1, 19RO.
18.1.4.2 interpretation None rea>>i ed.
SSFS-FSAR Statemen of Response A program is es.ablished +o assure that all reactor operato" and senio" reactor opi rator licer.se cand'dates ( bey or.d ",.he i ni" ia 1 compliment required to star up Jlni+ 1 r'. 2) have tho. prescribed exper'nce, prepared and aua lifications, and t "a' certified in accordance inq. C andidat:s will be with bT I-QA-3005, >> L'ensed Operator Traininq Program-Implementation<<. Admln3.s ra 1 ve procedure AD-QA-30rr, <<Ope .ator Selection Tra ininq and Qualifications,>> details ".he process by whic h the qualif ications of candidates for operat'ons posi+ions vill be evaluated in the fu+ ure.
The initial startup crews have completed extensive <<raininq devised in part to recognize the non-operational sta+us of the units. This proqram includes "eal time .raininq on he Susquehanna SFS simulator which duplica.es the actual unit and thu, in many respects eauates to the expe" ience requirements.
Subsection 13.1.3 describes the qualifications commitments for the existinq plant staf f.
- 18. l. 5 ADi+IV<ISTRAT'IO":J OF TBAIiJIiVG PROGRAMS JI A 2. 3}
- 18. 1,. 5. 1 Stateme rt. of Reqiiirem~n ..
Pendir.q accreditation of traininq institutions, licensees and applican.s for. operat'nq .licenses will assure that trai ning center and facility instructors who teach systems, in<<eqrated responses, transient, and simulator courses demonstrate senior reac.or opera or (SRO) qualifications and be enrolled in appropriate requalification proqrams.
Trainir.q center and facility 'nstruc-,.ors who teach syste~
intearat d respon es, transien'nd simulator courses shall demons.ra.e their competence to NRC by succes'ful completion of senio operator examination. Effec+ive date: Applications shorrld be submitted no later than August 1, 1980 for individuals who do not already hold a senior operator licer.se.
t Instructors "-hall be enrolled in app" opriate requalification proqram. to assure:hey are coqnizant of current operatinq history, problems, ar.d chanaes to procedures and admini trative limi.ations. Effective date: Programs should be initiated Hay
- l. 1980.
Auaust 1, Proarams 1980.
should be submitted +o OLD fcr review hy
- 18. 1-7
- 18. 1,5. 2 Ir ~er oretat ion The "irstructors" referenced in his requir=ment are those individual.. who teach sys",.ems speci ic to R'<;?s, integra ted responses, transients, and simulator courses to 1'ensed ope ators or licen e candida es.
18,,1.5.:3 S-atem nt of Response Certif ication of instruc+ors ' described in Nuclear Department.
Instruc+'on NDI-QA-4.1.3. Th's procedure del'reates which instructors are required to pass an examination for certification of senio" reac.or operators (SRO) . All ins" ructors who teach materials identified in Subso..c+ion 18.1.5.2 are certified as SROs.
- 18. 1.6 8 EVISE SCOPE AN D CRIT ERIA FOR LICENSING EX A."1I N ATIONS gI. A. 3. 1$
18,1. fi. 1 Statement of Reau'.. ement A !.ew cateqory shall be added +o the operator written examina+ion entitled, "Principles of Heat e r Transfer and Fluid Mechanics."
A new cateaory shall he added to the senio- operator w" itten examination entitled, "theory of "luids and Thermodynamics Time limits "hall be imposed for complet ion of the wr. iten examina+ ions:
- 1. Ops rator: 9 hou s.
- 2. Senior Operator: 7 hou s.
The passinq grade fo the written exam'nation shall be 80%
overall and 70% in each cateaory.
All applicants for senior operator licenses shall he reaui ed +o be administered an operating;.'st as well as the written examination. Zffec+ive date: Examinations administered on or a fte- ."Jay 1, 1980.
Applicants will grant permission .o NRC to inform their facility management reqardinq .he esults of the examinations for purposes of enrollmen. in requalifica ion programs. Applications received on or a+ter tray 1, 1980.
- 18. 1-8
0 SSPS-FSAR Simulator examinat'ons will he include~ as par. of the licensina examinations.
18.1.6.2 t r r t IInterprets:ion None reauired.
- 18. l. <).3 Statement of Respons~
The reac, or operator and senior reactor operator trainirq p=oqram has been upqraded to include the subject material described 'n this requirement.'I Refer .o Subsec ion 18.1.4.3 for the response to require ment ~ A ~ 2. 3, >> Immediate fJpgr adinq o f Peac or Operator and Senior Reactor Operator Traininq and Qualif icat ions. >>
Candidates will be prepared. and cer-..ified in accordance with Nuclear Depar ment In true+ion NDI-QA-4. 2. 1. Tho. Susquehanna SKS simulato. is available for +he simulator portion of exam".
Apolication packaqes include a release which permits the YRC to
'n form PPGL management of exam- results.
18,1 7 FVALlrATION OF ORGANIrATION AND ~ANAGFAENT gI.D.1.2$
18,1,7,1 t t em<
Sta e ntt of Reauii r eement Fach applicant for an operatinq license shall establish an onsit~
ind< pendent safety enaineerina qroup (ISFGj to perform independent reviews of plant operations.
The principal function of the TSHG is,o examine plant operating characte istic.", NRC issuances, Licersinq Information Service advisories, and other appropriate sources of plan" des'n and operatina experience informa tior..hat may in<! ica te areas f o" improving plant safety. The ISING is to perform independent review and audits of plant activities includirq maintenance, modifica.ions, ope"ational problems, and ope..ational analysis, ard aid in the establishment of programmatic req>>irements for plant ac-.ivities. <There useful imp ovement": c~n he achieved, it is expected that this group will develop and present detailed recommendations to corporate manaqement for such things as rev'sed procedures or equipment modifica iors.
Another function of +he TSEG is to maintain surveillance of plant operations and maintenance activities o provide independen+.
verification that these ac..ivities are performed correctly and that human errors a"e reduced as far as p"acticab'e. The ISFG
- 18. 1-9
SSFS-FSAR will then be in a position to advise util ty manaqement on the overall auality and safety of opera.ions. Th~ ISEG need not p--"or@ de:ailed audits of plant ope.at'cns and .=.hall not be responsible for siqn-off functior s such .ha- i" becomes involved in ~he operat'ra o"qanizat'on.
The new ISEC shall not replace +he plant op rations rev 1&w committee (PORC) and ".he u+ility's indenend<~nt review nd audit qroup as'specified by current staff quid@lines (Standar d Review Plan, Reaulatory Guide 1.33, S'andard Technical ca tions) .
it is an additional ir<deper.dent aroup of a Specif'ather.
dedicated, full-+ime enaineers, located onsite, bu min'ive mu<m 0 f epor+inq offsite to a corpora..e offic'al who holds a h iqh- leve 1, technically orien+ed position that is not in the manaqe men+ cha in fo powe= production. The ISEG will increase the avail able technical expertise located onsite ar.d will provide con tinuinq, systematic, and independent assessment of plan act'vit ies.
Tnteqratinq the shift techn'ical advisors {STAs) into th e ISEG i some way would be desirable in that it could enhar.ce th e qroup's contact with and knowledqe of day-to-day nlant operatio ns and provide additional expertise. However, the STA on shif t is necessarily a membe" of the operatinq staff and cannot be independen+. of it.
is expected that the ISEG may interface with ~he quali+y assurance {QA) orqanization, but preferably should not be an inteqral part of +he QA orqanization.
The functions of +he ISFG require daily con:act with the operatinq personnel and continued acce s to plan-: facilities and records. The ISEG review functions can, therefo"e, best be carried out by a qroup physically located on-ite. However, for utilities with multiple sites, it may be possible to nerform portions of the independent safety assessment function in a centralized location for all the utili,.y's plants. In sucn cases, an onsite qroup still is required, bu. it may be sliqhtly smaller than would bc the case ' it. were performinq thc entire independent safety assessment f unct ion. Such case vill be reviewed on a case-by-case basis.
Th' reauirement shall be i mplemer.ted p=ior :o issuar.ce of an one atina license.
Refer to Subsection 18. 2.6 for the response to additional "equi=ements contained in NHRFG 0694.
18.1.7.2 Interpretation None eaui"ed.
SS F. S- F S A P.
- 18. l.7. 3 Sta+emen, o f Pesgonse The functions of the ISING a"e performed hy the H>>clear Safe.y Assessment Group (HSAG) . PB~~L' commitmen t to the HSAG i, addressed 'n a letter +rom 'l. ".. Cur..is to B. J. Younqblood on December 8, 1980 (PLA-585) arid are fur. her addressed in V>>clear Departmen+ Tnstruction HDI-1. 1. 2.
- 18. 1.8 . J!ORT-THRN ACCIDFNT AHD PROCFD!JP..". PRVI.,N gI.C. ]}
18.1.8.$ Statement of Requirement.
Reanalysis of small break LOCAs, tran."'nts, accidents, and inadequate core coolinq and prepara.ion of q>>ideline." fo" development of emerqency procedures should be completed and siibmi+ ed to the HBC for rev'w by January 1, 1981. The HBC staff will review the analyses and guidelines and determine thei" acceptabili+y by J>>ly 1, 1981, and will issue quidance +o licensee. on preparirq emerqency procedures from the g>>idelines.
Followinq !1RC approval of the quidelines, licensees and
- applicants fo" operatinq licenses issued prior to January 1, 1982. should revise and implement their emerqency procedures at the first refuelinq o>>taqe after January 1, 1982. Applican+s fo.
operatinq licenses issued after January ), 1982 should implement the procedures prior to operation. Thi- schedule supersedes the implementation schedule included in H!JBFG-0578, Becommendat'on 2.1.9 for item T.C.1(a) 3, Reanalysis of Transients and Accidents.
For those licensee - and/or owners qroup that will have difficul y in attaininq the January 1, 1981 due date for submittal of quidelines, a comprehensive proqram plan, proposed schedule, and a detailed jus;.if ication fo= all delays and problems shall be submi+ted in lieu of, he a>>idelines.
18.1.8. 2 In.eroretation The BRB Owners'roup quidelines may be utilized to develop emerqency procedures for accidents and t "ansients.
- 18. 1.8.3 Statement. of Response In the Clari'fication of the H!JBHG-0737 equi" ement >>for reanalysis of transients and accidents and inadeauateore coolinq and preoaration of q>> delines fo- development of emerqency procedures," NUBFG-0737 states:
SSHS-FSAB Owners aroup or ven<lo r submi.tais may be =e. e"enced as app op to support this reanalysis. Zf owners'roup or vendor submittals have already been fcrwarded to the staff for re view, a brief de sc=iption of the submiitals and justif ication of their adeauacy to support quideline d evelo pment is all tha 3.s reau3.re<. ~
PPGT. has participated, ar<<) will cor.t inue tc 'participate, in the B<~ H Owners'roup program to dev clop Emergency procedure Gu'delines for Gene=al Elec =ic Boil ir q hat ~r Reactors.
Foll owinq are. a brief description o". the subm'ttals to date, and a justif 'cation of thei' adequac y o support guideline devel o pmen t.
Description of Submittals (1) HED0-24708A, "Additional Informatior. Required for NRC Staff Generic A~port on Boilinq Mater Reactors,"
<>e vision 1, December 1<380.
Description and analysis of small break loss-of-coolant events, consider'r.g a ranqe of break sizes, location, ar.d conditions, including equipment failures and opera+o" errors; descriptior. and justifica .ion of analysis methods.
fb) Description and analysis of loss of feedwater events, includinq cases involvinq stuck-open relief valves, and includinq equipment failures and ope"ator e rors; description and justification of analysis methods.
(c) Description and ar.alysi of each FSAB Chapte= 15 event resul ina in a reactor sys.em tran.ient; demonstrat.ion of applicability ibad'es; of analyses +o each even,; demonstration of applicability of Emergency Procedure Guidel ines to ea ch event.
Description of natu al and forced circulatior.
roolina; factors inf luencinq natural circulation, including no neon dens r eest ah lis hme n t of forced circulation unde" t=ansient and accident conditions.
(e) Descript'on and analysis of loss-of-coolant events, loss of feedwater even"s, and stuck-open relief valve events, includinq severe multiple eauipment failures and operator er"ors which, if not mitiqated, could result in condit'ons of inadeauate core coolinq.
SS"S-CESAR (f) Desc" iption of in<iications availa41e +o the BHR opera or for he detection c+ adequate core coo linq (q) Description and jus ifica~icn of analysis methods
~
fo= extremely deqrade:1 cases.
(2) NEDO 24934, "BrtB Emergency P" ocedure Guidelines BBB 1-6," Pevision 1, January 198].
Guidelines for BVR amer qency Pr ocedu e."- based on iden+ifica ion and esponse to plant symp:oms; incl<rdinq a range of equipment fa'lures and operator errors; ncludinq severe multiple e<ruipment failu es
~
and operator errors wh'ch, if not mitiqated, would result in condit'ns of ir.adequate core cooling; includinq conditions when core ccolinq s atus is unce tain or unknown.
Adeqrracg of Submittals The submittals desc"ihed n paraqraph A have heen discussed and reviewed extensively amonq the B'~P Owr.ers'roup, the General Electric Company, and the NRC staff. The NBC staff has found (WUREG-0737, page Z.C.1-3) that "the analysis and quidelines submitted by +he General Electric Company (GE)
Owners'roup...comply with the requirements (of "..he NUREG-0737 clarifica+ion) ." In Reference 18.1-1, the D'ector of the Div sion of Licensinq states, "we find ".he Emergency Procedure Guidelines acceptable or trial implementation (on s x plants with appl'a-..ions f operating licenses or pendinq PPGT, believes that in view of these findings, nc f urther detailed justification of the analyse or q>>idelines is necessary at this time. Referer.ce 1 further states, <<during
+he course of implementation we may iden.ify areas that require modification or further analysis and just . =ca+ion."
The er.closure .o Reference 18.1-1 identifies several such areas. PPGL will work wi..h the B~.'R Owner.".'roup in esponding o such requests.
By ou commitment +o wo k with the Owners'roup on such requests, on schedules mutually aqreed to by the A)BC and the Owners'rouo, and by reference tc the BwB Owners' oup analyses and gu'delines already submitted, cur response to the h'OREG-0737 r~quiremen+ "for reanalyses of transients and.
accident, and irradequate core coo]inq and preparation of quidelines for developmer.t of emergency p" ocedu "es" by january 1, 1981, is complete.
SSES-FSAR Emerqency proceduros have been developed ba=ed on those qu ide lines.
18,1. 9 SHI~T RELI F AND TURNOVFF PRACFDU"'S gE. C. 2g No requi"ement stateR in NUPEG 0737. Refer to Subsection 18.2.8 which contains <<he response to the requirement in NUHEG 0694.
18.1.10 SHIFT SUPFR'JESOR RESPONSIBILITY QI.C.3$
No reauirement stated in NUTMEG 0737. Befe to Subsection 18.2.9 which contains the response to the requirement in NUREG 0694.
18,1,11 CO~~POL anpm ACCESS gI.C.4)
No requirement stated in NUTMEG 0737. Refer +o Subsection 18.2. 10 which contains the response +o the requirement in NUPEG 0694.
18, 1,12 FEEDBACK OF OPERATI NG EXPERIENCE JI.C. 5)
- 18. 1.12. 1 S t t atement of r Requirement Applicants for an opera+inq license shall prepare procedu" es in assure that information pertinent to plant sa fety oriqinatinq inside o" outside tne utility organization i= continually supplied o operators and other personnel and is incorporated into train'nq and retra'ninq p"oarams. These p ocedu es shall:
(1) Clearly identify orqani"ational responsibil'es for review of operatinq experience, -'.
he f eedback of pertinent information to operators and other personnel, and the incorporation of such information into train'ng and retraininq programs; (2) Identify "..he administrat'e and technical review steps necessary in translatinq .ecommendations hy the operatinq experience asse ..ment group into plant actions (e. q.', chanqes to procedures, operatinq orders);
(3) Identify the recip'ents of various ca:.ego" ies of information from operatinq experience (i.e.,
supervisory personnel, shif" technical advisors,
SSES-FSAF operators, maintenance nersonnel, health physics technician") o" otherwi-e p ovide m ans through which such information can he read'ly rela+ed to the job fiinctions of the recipien Provide means to a "sure that affecte<1 personnel hecomri awa "e of and understand iriformaticn o,. suf icient imnortance ~hat should not wait fo= emphasis -.hrouqh routine tra ininq and retraining programs; (5) As u..e that plant personnel 0o noi routinelv eceive extraneous and unimportant informa 'on on operatinq experience in such volume that iz would obscure priority info mation or othe= wise <retract from ove all job performance and prof icier.cy; (6) Provide sui able chocks to assure that conf lie inq or contradictory information is not conveyed to operators and o+her personnel until resolut on is reached; and, (7) Provide periodic internal audit o assii"e that +he feedback proqram functions effectively at all levels.
This requiremen+ shall he implemented prior to issuance of an operatinq license.
18.1.12.2 Interpretation r
Hone required.
.1,Q,3,,12. 3 Statemen t of Response PPFfL las developed a comprehensive proqram for feedback of operatinq experience. Components of he proqram are as follow-.
Operat'nq exper'ence f .om other u,.ilities and other industry sources is initially reviewed and disposi+ioned by the Industry Fvents Feview Proqram (IEHP) . The IFRP is designed to assure plant persorinel do not routinely receive extraneous and unimportant information, that information is no. contraR'c+ory or conflict'nq, that information 's esolved p io= to dissemination and that important information is rapidly routed to the app onriat~ personnel. A desc=iption of th organization, tesponsibi i ities and procedures of the IERP can he found in Nuclear Department Ins truct ion NDI-QA-6. 2. 2.
The Shift Technical Advise= (STA) as pa of the Operations Assessment Function will he the focal point for dis .emination of
SS'S-FSAB operatinq experience informa= on 'to happ" opriate plan- personnel.
This will include; Feedback of. pertirent. information to operators and othez s:ation pe"sonnel and t:ansmi.+al of. info" ma ion to the Nuclear Traininq Group for inco"poration in-o app"opriate rain'q p=oqram o Initiatinq, when requi .:d, plant procedure changes and/or plart modi fica..ion =equests.
o Discussinq with shift p~rsonne1 ope" a+ing expe ienre in formation of suf f icien t impor .a r.ce t ha i'. camo be def ez red to the =et..a'.inq p"oq" am.
o Editinq information provided to plant personnel to minimize excessive oz con flic=i nq in for mat ion and d is+ r i bu .. i ng information to appropriate furctional units.
Adm. nistrative Procedure AD-QA-406 further defines this func ion and the interfaces amonq the STAs and the Nuclear Safety Assessment Group. Nuclear rraininq, Opera+ions and the Industry Even+s Beview Proqram.
General information from +he nuclear indus.ry anu information of qeneral interest from inside the company will be disseminated +o appropriate personnel. The details of +his program are described in Nuclear Department Instruction NDX-OA-6.2.1.
The NQA oraanization will selec+ively audie, portions of the feedback 0" oqram to assure i+. functions effec.ively a all levels.
- 18. 1 ~ 13 VEBIFY COBBECT PEPFOBi1ANCE 0. OPEPATING ACT V3:TIES (I.C. 6)
- 18. 1. 13. 1 S ta temen+ of Beau iremont Licensees'rocedures shall be reviewed and revised, as necessary, to assure that. an effective system of verifying th*
correct performance of operating activi+ies is provided as a means of reducinq human errors and improvinq the quality of normal operations. This will reduce the frequency of occurrence of situations +hat could esult in or contribute to accidents.
Such a verification system may include automat.ic system status monitorinq, human verification of operations and maintenance activities independent of the people performirg the activity (see NtJBEG-0585, Becommendation 5), or both.
SSES-FSAB Imp lerner.ta+ ion of aut oma tic s +a tus mor itorirq '.. reauir ed will reduce the extent of human ve maintenance act'ities rif 'ca.ior. of operations a rd i
but wi 11 not <<1' na-,.e - he n eed for such verificatior. in all instances The procedu=es adopted by the licensee= may consist of two phases--one befo e and one a fte" ins+allation of automatic sta tus monitorinq equirment, recruired, ir. accordance with item I. D. 3.
if Procedures must be reviewed and revised p" io= to f uel load.
1.8-1. 13. 2 In ternretat ion Hone reauired.
18.1.13.3 S .agement of Response Administrat've procedure AD-QA-306, "System Status ard Equipment Control,'i provides the means to verify co rect performar.ce of sur veillance and maintenance activities. S.atus veri icatior.
utilizes con".col noon indica,.ion . passant lv available, operability testinq where appropriate, or indeper.dent verification by a second qualified person. The procedure defines circumstances when independent human verification is required.
The procedure also 'ncorporates the requirements of i em XI.K.1.10 (see Subsection 18.2.26) for the removal from ar.d restoration to service of safety related systems and components durirq normal operations and maintenance activities.
3v8. 1. 14 NSSS VENDOR REVIFrtT OF PBOCEDUB ES JI C 7}
No requirement,. stated in HUBEG 0737. Refer to Suh ection 18.2. 12 which cor.tains the response to the "equirement in NUBEG 0694.
18.1.15 PILOT NONITORING OF SELECTED ENFBGc:NCY PROCEDURES FOP NEAR TERN OPERATING LICENSES JI.C.RQ No requirement stated in NUREG 0737. Refer to Subsection 1P.2. 13 which cortains the resporse to the requireaent in NUBFG 0694.
- 18. 1-17
18.g.16 CONTROL BOO~ nESZGN B~yrEV gr. D.lg
].8. 1. 16. 1 St a< ement e of ReOu'ement All licensee"= and applicants for. ope;a-.ing licenses wil1 be required to conduct a Qe+ailed cont ol-rccm design r. view to identify and correct desiqn deficiencie . This detailed control-room design review is expec ed to take about: a year. rheref ore, the Of fice o f Nuclear Reactor Regulation ('.ERR) " 'quires t ha, hose apn] icants for operatina licenses who are unable ~o complete this review prior to issuance o" a 1'cense make n~eliminary assessments of their control corns to identify iqnificant human factors and 'strumentation problems and establish a schedule (to be approved by NBC) for co recting deficiencies. These applicants will be requ'red to complete the more detailed control room reviews on the same schedule as licensees with operating plants.
Applicants will find NUREG/CB-1580, "Human itEnqineerinq of value to =efer to the draft document Guide to Cortrol Room
"=.valuation, << in per forminq the preliminary assessmen".. NRB will evaluate the applicants preliminary assessments includinq the performance y NRB of onsite review/audit. The MRR ons'e
?
review/audit will be on a schedule consistent with licensing needs.
This requirement shall be. met prior to fuel load.
18.1.16.2 interpretation Aoplicants for operatinq license are reauired tc perform a prelimina "y control room desiqn assessment which should be based on NUREG/CP-1580. Th's assessment vill be reviewed by the NRC, who will subsequently recommend changes fo.. co rec.ing deficiences. Applicants mus+ submit for NBC approval a schedule for correctinq +hese deficiencies.
Applicants will be reauired to perform a de+ailed control rocm desiqn assessment followinq NURFG 0700 issuance. Th's i" not requ'red to be completed prio" to issuance of an assessmen+
opera+ing license.
- 18. 1.16. 3 Sta tement of Response A detailed control room review to identify significant human factors problems was conduc ed by PPF.L with assistance f rom
SSES "FS AB expe ienc d human factors pe"sonnel f rom Gen. al Physics Co poration. This review was based on the crite=ia qiven in draft iVURKG/CB-1500 ~
D>>r.'a the week of Oc+obe 27, 1900, the NBC pe r foe med an onsi<<.e review of the Susquehanna cnr.trol oom. The re s>>1 t s o this review were formally t ansmitted to PPGL or. Jan uary 31, 1981. A meetinq was held on Februa y 3, 1901 in Be".hesd a to disc>>ss and clarify the NHC findir.gs. On Februa y 27, 1901 pp6 L submi"ted a fo mal response to all NBC findings (refer +o P LA-6 40) . Th'
,response included a schedule for implementing he f i.nd 1ngs addre.,sed in the HRC report.
All of these findings were addressed prie" +o U n' 1 fueL load or have been incorporated into the scope o th<<pl armed NUBEG 0700 review. All modifications that we required to be imolemented prior to Unit 1 fuel load vill also be implemen ted in Uni+ 2 prior to fuel load (if applicable). Also see P LA-1621 da ted 4/15/83 fo" PPGL's resoonse to Generic Let te" 0 2 3 3 ~
18.1.17 PLANT SAFETY PARAMETER DISPLAY SYSTEM gT. A.2}
18.1.17.1 Sta+ ement of Beauiremen+
Each applicant and licensee shall install a safety parame+er disolay system (SPDS) that will display to operatinq pe=sonnel a minimum set, of parameters which def'ne +he safety status of the plant. This can be at+ained =hro>>gh continuous indication of direct and deri.ved variables as necessa y to assess plant safety s ta lls The operational date for he SPDS is October 1, 1.982.
1.8, 1 ~ 17..1. 1 Function The purpose of the safety parameter <<1isplay system (SPDS) is to ass'st control room personnel in evaluating the safe.y status of the plant.. The SPDS is to provide a continuous indication of plant paramete- s or derived variables reoresen-ative of the safety stat>>" o the plant. The primary function of the SPDS is to aid the oper ator in he raoid. detection of abno"mal ope atinq condi.ions. Th e functional criteria fo" the SPDS presented in this section a= e applicable for use only ' the con trol room..
It is recoanized that, upon the detec+ion of an abnormal plant +o status. it may he desirable to provide additional information analyze and diagnose the cause of ..he abnormality, execute
SSF.'S FSAB corrective actions, and mon'tor plant, response as secondary SPDS functions.
As an operator aid, the SPD S serves o concentrate a mir.imum set of plant parameters f rom wh ich the plant sa fe >y status can hc assessed.; he qrouping of parameters is based or. the fur.ction o enhancinq "he operator 's ca pabilitv .o asses- plant atus in a timely manner withou surve yinq the entire control room.
However, the assessment bas ed or. SPDS '.. likely -,.o be followed by confirmatory surveys of man y non-SPDS contzcl rocm indicators.
Human-fac.ors engineerinq shall be incorpczntecl in the various aspects cj the SPDS desiqn:o enhance r.he fur!ct ional effectiveness of control room personnel. The d e. iqn of the primary or princ. pal display forma. shall be as simple as possible, consis+ent with t he reauired f unctior. and shall include pattern and coding techniques to assist the operator's memory recall for the detect'n and recoqnition cf unsa f e operatinq condi ions. The human-fac" ozed conce ntrazion of these signals shall aid the operator in functionally ccmparin q signals in +he assessment of safety status.
All data for display shall be validated where practicable on a realtime basis as part of the display +o cent zol room personnel.
Fo- example, redundant sensor data may be compared, the zarqe of a parameter may be compared to predetermined limits, or other quantita ive methods may be userl o compare values. Mhen an unsuccessf ul validation of. data occur., the SPDS shall conta in means of ident' yinq the impacted par amet ez {s) . Oper at inq procedures and operator traininq in the use of he SPDS "hall contain 'nformation and provide guidance fo: +he resolution of unsuccessful data validation. The objective is to ensure that the SPDS presents the most current and accurate status of +he plant possible and is not compromised by unidentified faulty processirq or f ailed sensors.
The SPDS shall be 'n operation du" ing no" mal and abr.ormal.
opezatinq condi+ions. The SPDS shall be capable of display'ng pertinen+ info"mation durinq steady-state and tzansien.
conditions. The SPDS shall he capable of pzeser.ting the magnitudes and the trends of parameters o" derived va=iables as necessary to allow rapid assessment of the currer.t plant sta+us by control room personnel.
The parameter trendinq display shall contain recent and current maqnitudes of the parameter as a functior. of time. The derivation and presentation of parameter trendinq during upset conditions is a task that may be automa+ed, thus fzeeinq +he operator to in+erpret the trends rathe- than qenezat e them.
Display of +ime derivatives of the Oarametezs in lieu of trends to both optimize operator-process commun'ca..ion and con ezve space may be acceptable.
- 18. 1- 20
SSHS-FS AP,
~
he SPDS mav be a source of. information to o her svs-ems, ard the func+'onal criteria of these systems shall s.,ae the regu red~
interfaces w'th the SPDS. Any irrterface b ct~een the SPDS and safety system shall be isolated 'n accorda nce wi'h the sa fe.y system criteria +o preserve channel ind~pe nR-:.nce and ensu re the intear'-y cf the sa ety system in he case of S?DS malfun ct ion.
Desian provision= shall be included in the interfaces bet ween ",.he SPDS and nonsaf ety systems to ensure the n egrity of the S? DS upon fai'rrre of nonsafet,y eau'pmer.-,.
qualification program shall be established to demons. rate SPDS conformance to rhe func+ional criteria of this document.
18;1..17.1.2 Location The SPDS shall be located in the control rocm with additional display.= provided in the TSC and -:.he FOF. The SPDS may be SPDS physically separated from the normal control boards; howeve, shall be readilv accessible and visible to the shift supervisor, it control room senior reactor operator, shift technical advisor, and at least one reactor operator from the normal operati.nq area.
If the SPDS is part of the con+rol boa" d, i. shall be easily recoanizable and readable.
18.1.17.l. 3 Si ze The SPDS shall be of such size as to be ccmpatible with the existinq space in the control a ea. The SPDS display shall be readable "rom the emergency operating s+ation of the control room senio reactor operator. I,. shall not interfere with normal movement or with full visual access to othe control room operating systems and displays.
- 18. 1. 17.1. 4 Sta ff ina The SPDS shall be of such design tha. no opera+ ing personnel ir.
ad<hi.ion to the normal control room opera tinq staf f are equirerl fo" its operation.
18.1.17.1.5 Display Considerations The display shall be responsive to trans'en-.. and accident sequences and shall be sufficient to ind'cate the status of the plan . For each mode of plar.t operation, a s'nqle primary display format desiqned acco.dir.q to acceptable human-factors
SSES-FSAH pri nci ples (a 1imi ted number of nazame-ers o=:le iv.d variables and ante the Dre j t rends jn an orqan'zed display tha can be "ead'ly pd by an operator) shall he displayed, from which plant safe y status car. be inferred. It is "ecoqn'. zed tha-.. i= may be desirable to ha ve the capability +o recall additional da=a on secondary formats o. d 1s pla ys ~
The primary display may he inddual plant Dazameters o= may be composed of a number of parameters or dorive l va "iahles givina an overall system status. The basis for the selection of -.he minimum so+ of parameters in the prima ry display shall he documented as pa" t of the de. ign.
The importan .. @lant functions relat.- d to the pr'mary display while +he plant is generating powez shall ir.elude, but not be limited .o:
0 Reactivity control o React.or core cooling and hea., removal from pz mar y =ys~ em z.
0 Reactor coolant system inteqrity 0 Radioactivity control 0 Containment inteqrity The SPDS may consist of several display fozma+s as appropr ate +o monitor and presen+ the va"'ous pazamete"s o" derived variables.
For each plant operatinq mode, these format may ei.her he au.omatically displayed or man>>ally selected by -'e opezato- to keep control room operating personnel informed. Flexibility to allow foz in e action by the operator is desi "able in the di play designs. Also, where feasible, the SPDS should 'nclude some audible notif'cation to a3ert pe sonnel of an unsafe operating condition.
The SPDS need not be limited to +he pz~vio>>sly s;ated unct'ons.
It may ir.elude othe func+ions that aid operatinq personnel in evaluatinq plan status. It is desirahl~ that the SPDS he sufficiently flexible to allow for future ircorporation of advanced diaqnos+ic concepts and evalua+ion techniques and systems.
18.1.17.1.E Design C iberia The total SPDS neerl not be Class 1E or meet the single-failure criterion. The sensors and signal conditioners (such as preampli.;iers, isola+ion devices, e.c.) shall be desiqned and crualified to meet Class 1K standa ds for those SPDS parameters that are also used by safety systems. Furthermcre, sensors and siqnal conditione"s fo" those parameters of the SPDS ider.tical to the parameters specif'ed within Regulatory Guide 1.97 .hall be designed and aualified to the criteria ta+ed in Regulatory Guide
- 18. 1-22
1.97. For SPDS application, i. i also acceota ble to have Class 1E qualified devices from the sensor:o a post- accident-acce.".sible location, such as outside containmen t, and then non-1E devices from contair.ment to the d'splay (or pro c assoc) on the presumpti on tha". +hese componer.ts can be repa' ed'c replaced in an accident env'onmenz. The processing anR di splay devices of
.he SPDS shall be of proven hiqh qualitv and re 1 iabili . y.
he funcion of the SPDS is -,.o aid '.he operator in the interore ation of transients and acc'Rents. Th is funct ion sha11 he prov'ded during and followinq >13 events exo ected to occur du inq the. li e of; he olan+, including ea thai> akes. To ach'eve
.his funct ior., the display svstem shall. not onl y take adequate account of human fac..ors--the man-machir.e in-..ec face--bu+ sha ll also be suf iciently durable to function during ar.d after earthquakes. Because of current echnoloqy, it may not be possible ",.o satisfy these cciter ia withir: one S PDS s Vst em From an ope"a+ional viewpoint, it. is pcefe=red that only one display ..ystem be usod for evaluatinq .he safetv status of the plant. One disolay svstem simplifie the man-machine interface and thus minimizes operato" er"ors. However, in recogn'ion of he restraints imposi d hy current technoloqy, an alternative is to desiqn the overall SPDS functior. wirh a primary and backup displav sy tern: (1) the primary SPDS display woulD have hiah performance and flexibility and be human fac.oreR bu+ need not be seismicallv qualified; and (2) the backup display system would be operable du inq and followinq earthquakes, such as the ..ormal control room displays needed to comply with Regulatory Guide 1.97. The display system (or sys ems) prov'ded for the SPDS function shall be capable of function ina duc'nq and following all desiqn basis events for the olant.
Tn all cases, both the prima=y SPDS display anil the backup SPDS seismicallv aualified portion of the display shall be sufficiently human fac.ored in its design eo allow the control room operations staff to perform +he safety status design to allow the control zoom opeca+ions staff to perform the safety sta+.us assessment task in a timely manner. Dependence on poorly human-engineered Class 1E seismically qualif'ed inst "uments that are scattered over the control board, ra~her +han concentrated for rapid safety status assessmen, is not acceptabl: for this function. An acceptable approach would be to corcen=rate the seismicallv qualified display into one seqmen. of the control board.
The dynamic loading limitations of he SPDS desiqn shall be def ined and incorporated into the traininq p" oqram. The control room operations staff shall be provideR with sufficient information and criteria to allow for performance of an operabilitv evaluation of SPDS is an earthquake should occur.
- 18. 1-23
SS S-ESAR The SPDS as used in the control zoom shall be desiqned to ar; operational unava'labil'ty goal of 0.01, as def'ned in Section 1.5 of NUREG 0696. The cold shutdown unavailabi) ity coal fo th SPDS durinq he cold shutdown and efuel'nq modes fo- .he reacto shall be 0. 2, as def 'ned in Section 1.5 of NVREG 0696.
Technical specifications shall be es.ahlish ed '.c be con sist~ nt with the unavailability design goal of the SPDS and wi".h the compersa"ory measures provided durinq pe ric ds when the SPDS is inoperable. Operat'on of the plant w'th th SPDS ou o f ser vice is allowed provided that the cont ol boa d su f,zicient ly human factored to allow the operat'ons sta. f to p erfozm .he safety statu assessment task in a ,.imely manr~r. Dependence on poorly human-egqineered ins"ruments tha~ are scatt ered over he control board rathe than concentrated fo- rapid sa f ety =tatus assessment is rot acceptable for this function.
- 18. 1.17.2 Interpol etation None required.
- 18. 1.17.:3 Statement of Response Deta'ls on the SPDS are presented in the Emergency Plan and our "esponse to Gener' Let te" 82-33 (PLA-1621, dated 0/15/83) .
- 18. 1.18 TRAINING DURING LO!J-POKER TESTING QI. G.l}
No requiremert stated in NUREG 0737. Refer to Subsection 10.2.15 which contains ..he response to tte equirement in NVBE' 0694.
$ 8.).19 REACTOP. COOLANT SYSTEM VENTS /II.P.ll 18.1.19.1 Statement of Requirement Each applicant and licensee shall install reactor coolant sys em (RCS)'nd reactor pressure vessel (R!V) head high Doint vents remotely operated from the control "oom. Althouqh the purpose of the system is to vent noncondensible gases from the RCS which may inhibit co=e cool'nq durinq na+ural circulation, the vents must not lead to an unacceptable increase in +he probability of a loss-of-coolant accident (LOCA) or a challenqe to con+ainment inteqri+y. Since these vents form a pa t, of the reacto coolant pressure houndary, the design o" the events shall ccnform to the
- 18. 1-2Q
SSES-CESAR requirements of Appendix A to 10 CPB Par~ 50, "General Design Criteria." The vent system shall be desiqned wi h s>>f" icient redundancy that assures a low probab'lity of inadvertent or i=reve"sible act>>ation.
Each licensee shall provide the followinq information concerninq the desiar. and operation of -,.he hiqh poinr. vent "ys~em:
Subm ".. a descript '.. o.f the desiqn, 1ccation, size, and power supply for the ven+ system alonq wish results of analyse fo" los.. -of-coolart accidents iniria ed by a break in t?) e vent pipe. The results cf the analyses should demonstra. comoliance with the acceptance criteria of 10 CF R 50.06.
(2) Submit procedures and supportinq analy is fo" opera+or use of the vents tha+ also include,.he information available to he operator for initiatirq or .erminatinq vent, usaqe.
Documentation shall be submitted by July 1, 1981. Nodifications shall be comple,.ed by July, 1082.
18.1..19.2 Interoreta+ion None required.
.18.1.19.3 Statement Af Response The present desiqn of reac+or coolan~ and =eactor vessel ven systems meet these requirements.
The RPV is equipped with various means to vent the reactor durinq all modes of operation. All the valves involved are safety qrade, pcwered by essential busses ard a"e capable of remote manual operation from +he control oom.
The larqest portion of non-condersables ar~ vented throuah sixteen (16) safety relief valves (PSV 141F013A-S) moun+ed on the main steam l.ines. These power operated relief valves satisf y the in,.ent of the NRC position. Information reqardinq the design, qualification, power source of the e va3.ves has been p" ovided in Sections 5'. 1, 5. 2. 2, 6. 2, 6. 3, 7. 3 and 15.
Tn addition to power operated relief valves, the RPV is equipped with various othe" means of hiqh point ven inq. These a=e:
- 18. 1-25
SSES-FSAR Normally closed R.PV head vent valve (HV141-F001 and F002), operable ~ rom control room which di charges to drywell eauinm~nt drair. +ank. (Su",.sec=ion 5.1 and Fiaure 5.1-3a).
- 2. Yormally open reacto head ven line 2 DBA-112 which discharqes to main steam line "A". (Subsection 5.1 and F'aure 5.1-3a) .
3 Hain steam driven HCTC'and fiPCT. system tu=h'nes, operable from the cont. ol room whic). exhaust to suppression pool. (Subsectiors 5.3 and 6.3 and Figures 5.4-9a and 6.3-la) .
Although the power operated =elief valves fully satisfy the intent of the HHC requirement these other means also provide protection aqa'st accumulation of non-condensables in the HPV.
The desian of the RCS and HPV vent sys-ems is in aa"cement with he qeneric capabilities proposed by the BMH Owrers'roup, with the exception of isolation condensers. Susquehanna SES 's not equipped with isolation condensers. The BVH Owners'roup position is summa."ized ir. NEDO-24782.
Operation .of the equipmen+ de:crihed above during abnormal operatinq conditions is controlled by the Emergency Oporatinq Procedures. Hhile .hese procedures do not specifically address ventina of non-condensable gases, they do addr ss p=oper utilization of equipment to recover from undesirable conditions presented by the presence of non-condensables or by other circumstances.
The HCS and RPV vent ystems are part of the oriqinal Susquehanna
! SES desiqn basis. A pipe break in eithe cf these systems would be the same as a small mair. steam line break. A complete ma'nsteam line break is within the desiqn basis (see Subsections 6.2.1.1.3.3.2 and 6.3.3) . Smaller s'ze breaks nave been shown to he of lesser severity (see Subsections 6.2.1.1.3.3.5 and 6.3.3.7.3) . Therefore, no new suoportinq analysis is necessary in response to HUHEG 0737. In addition, no new 10CFH50.46 conformance calculations or cortairment comhustible qas concentration calculations are necessary. Non-condensable qas releases due +o a vent line break would be no more severe than the releases associated with a mainsteam line break. Nainsteam line break analyses included cortinuous ventinq cf non-condensable qases with hiqh hydroqen concen+ ations. These analyses demonstrate conformarce to 10CFH50.46.
- 18. 1-26
SSF S-FS AP 18,1. 20, Plant Shield ina gI.I. B. 2g I l.P,.
I<<) l.
I~ ~ I II 20.1
~ ~
S ta temen I ~ I>> II ~ II ~ I ~ ~ I t of Beau
~~
~
r irement
~
~ ~ t Pith +he assumption of a pos. accider t release o" radioactivity equivalent to that described in Pequla-o y Guides 1. 3 a nd 1. 4 (i. e., the eauivalent of 50'7" of the core radioio<3in~, 100'j of the core noble qas 'nventory, and 1'K of the core solids are contained in the nrimary coolant), each licensee shall perform a adiation and shieldirq-desiqn review of the spaces around sys.ems that may, as a "esul. of an accident, contain h'qhly radioactive materials. The desiqn review should identify the locat'on of vital areas and equipment, such as the cont.rol room, radwaste control stations, emerqency power supplies, motor con+rol centers, and instrument areas, in which pe sonnel occupancy may be unduly limited or safety equipment may be urduly deqraded hy the radiation fields durinq postaccident operations of these systems'ach licensee shall provide fo" adeauate access to vital areas and protection of safety equipment by desiqn chanqe , increased permanent or temporary shieldinq, or postacciden procedural controls. The desiqn review shall de+ermine which +ypes of corrective actions are needed for vital areas thcruqhout the facilit
$ 8.1.$ 0.1.1 Documentation Required e i r fo"r Vi~al 0 r ea Access e
Fo vital area acces, opera inq license applicants need to provide a summary of the shieldinq desiqn review, a description of the review 'esul+s, and a description of the modifications made or to be made ..o implement the result of +he review. Also to be provided by the 1'censee:
(1) Source terms used includinq time af ter shutdcwn +hat was assumed for source terms in systems.
(2) Svsems a sumcd to contain hiqh levels of activity in a post-accident situation and jusitif ication for excludinq any of those qiven in the>>Clarification>> of NURFG 0737.
Ar .as assumed vital for pos+-accident operations includinq
~ 4 (3) justification for exclusior. of any cf those qiven in the "Clarification" of NUREG 0737.
(4) Projec;ed doses to individual= for necessary. occupancy time in vital areas and a dose rate map fcr pctentially occupied area s
SSFS-FSAR 18,$ . 20.1. 2 Documentation Re<ruired fo" Equipment gua lificat ion Item II.B.2 s.ates, "Provide the information requested by the Commission Memcranc<!m and 0 der on equipment qua1if'a-.ion (CLI-80-21) ." This memorandum, wi, h regard to equipment qua lificaticn, requests info."-
matior. on environmental qualification of safety related electrical equipment.
18.1.20.2 Ir.teroretat'on
- 18. 1. 20. 2. 1 Sou"c< Ter m~
he source term for recirculated depressurize<', coolan need not bo. a. sumed +o contain noble gases, theref ere the RH2 shu+down coolinq system which may initiate at low reactor pressu "e only will be assumed to contain ..olely haloqens and particulate . The HPCI and LPCI systems do not recirculate reactor coolant hut, rather, suppres ion pool water. They vill al o be esser.+ially void of noble gases.
Leakaqe f om systems outside of. con ainment need no be conside ed as potential sources. Also, con. ainment and equipment leakaqe (from systems outside cor.ta'nment) need not be considered as potent'al airborne sources wi.hin the reactor buildinq. It follows that airborne sources and any other uncor..tained sources in the ro.ac+or building do not need be consi<".e ed in this shieldinq review.
- 18. 1. 20. 2. 2 Post-Accid en t Sist ems
'7he s" andbv gas treatment system, or equi valent, is qiven a system which may contair. hiqh levels of radioact'vity after an accident. Airborne activity from leakage cf equipment outside containment has been clearly established as beinq outsi<! e the review requirements. Drywell leakage must then provide the ac ivity processed by the SGTS. This review will assume the drywell does indeed leak to the reactor building to provide a source within the SGTS. However, this airbo r e ource will not be evaluated any fur+ her ir. the re vie v.
- 18. l. 20. 2. 3 Equipment gualif ica+ ion Provide a description of the environmental qualification proqram ard result.. for safety rela+ed electrical equipment both inside
- 18. 1-28
SS HS ASAP and outs'de of. containment.. It is our unde standinq tl.a radiation aualification of non-electrical sa=ety related .
equipment need no* be reported.
)8.1.20.3 Statement of Resnons~
The required post-accident study is divi')ed in=o twc parts; one dealing w.'th a summary of the shieldinq design review plus vital area access, another dealing with eguipmen g>>alification. A summary of the shieldi.nq desiqn review, results, and methodology used to determine radiation doses is presented below. The results of the eauipment qualification proqram a' scheduled <<o be submitted separately, and in compliar.ce wi h commission memorandum and orde CLI-80-21.
The re .ults of +he shieldinq review of contained sources are that all vi"al a"eas are access'ble post-accider.t and no shielding modifica+ions are necessary ..o comply "o NUPHG 0737.
- 18. 1. 20. 3. 1 In trod>>c tion lf an acciden, is are released from postulated 'n which large amounts of activity the reactor core, then pathways exist which can transfe" this activity +o various areas of the reacto building.
These larqe radiation source terms present a hazard reqardinq poten",ially hiqh doses to personnel. In order to deal with this problem it has become nece sary to g>>antify these scurce terms, trace tho.ir presence and determine thei effects on he efficien.
performance of post-accido.r:t recovery operatior.s. To this end, the plan.. shieldinq of rrnits 1 and 2 has been reviewed for po "-
accident a d equacy.
This summa y presents the analytical bases by which -.he review was carried out. Systems required or postulated to process primary reactor coolant ou+side the con+a'nment during post-accident conditions were selected for evaluation. Large radiation sources beyond the oriqinal selected svstems. Radiation levels in adjacent plant areas due to contained sources in piping and equipment of tho se systems were then estimated to yield the desired information. Also included herein is a discussion of radiation exposure quidelines for plant personnel, identificai'n of areas vital to post-acciden+ operatior.s and availability of access to these areas.
As a byproduct of. this review, several radia.ion zone maps and associated curves have been produced. The maps will allow operations personnel to 'den+ify potent'al high exposure vital areas of the plant should an accident occur which ccntaminates 18 1-29
SSrS-CESAR he system considered in this 'tudy. The curves will allow them
.o estimate radiat'on levels ir. chese areas a( various times followinq an accident.t
- 18. 1.,20. 3. 2 Design Review Rases
- 18. 1. 20,~,2,1 Systems Sel o.c, ed for Shield'g Review A review was made to determ'e which sys".ems could be " equired to operate and/or be expecteQ to contain highly radioac+iv 6 materials followinq a postulated acc.'den+ whe=e substan tial co=e Qamaae has occurred. The documenta+ion qoverninq the a pproach to the shieldinq "eview is MURFG-0737.
A review of containment isolation provisions was conduc ted in acco dance with item II.E.4.2. This was done to assure iso lat i on of non-essential svstems penetratina the. cor.+ainment ho undar y.
Thus, systems other than those identified as having a s pecified function followinq an accider.t are assumed not to conta ' post-acciden+ ac+ivit.y and do not need to be considered in t he shieldinq review.
18.1.20.3,2.1.1 Core Sorry HPCI BCIC and RHR QLPCI mode}
The Core Spray, RHB (LPCI mode), HPCI (wa+e" side) ar Q RCIC (water siRe) systems would contain suppression pool wa" er beinq in jected into the reactor coolant system. Althouqh the iiPCI and RClC systems could also draw from the conder.sa+e s+oraqe tank, uppression pool water is assumed to be their only source of water for in1ection. The steam side of the fiPCI and BCIC systems would operate on reactor steam and would not be required when the reactor is depressurized. However, as a f irst estimate for equipment oualification, it 's assumed ";.hat -hese systems should also be available until one year po t-acc'dent.
- 18. l. 20. 3. 2. 1. 2 BHR /Shutdown Coolinq Node)
The RHR system recirculates reacto waste heat when it, operates in t.he shutdown coolinq mode. Operatior. in this mode requires that the reactor be in depressurized conditicn. Depressurization is expected o remove substantially all of the noble qases released into the eactor coolant whether it be by direct ventinq to the drywell or by quer.chinq reac.or steam in the suppression pool. Ano+her consideration i , followirq a pos ulated serious acciden+, the HPCI, RCIC, BHR (LPCI "lode) and/or Core Spray
SS FS-FSAR systems would in ject a substant ial amount cf water into the reactor coolant system. This shieldinq review will assume that the"e are no noble gases 'n t he reacto= water in th e RHP system from the shutdown coolinq mode. However, since . he exac t amoun t of d'lut'on of the "eac'or water is difficult =o de ermine, no dilution in addi+ior. to +he reactor coclar.t vclumc ls a ssumed o 18.1. 20. 3. 2. 1. 3 RHB QSupDression Fool Coolino:abode)
~
he BHR sy .tern in this mode c'rculates and removes heat from
."-oppress'on pool water to preven. pool boilinn. This a . ures availability of suppression pool water as a source for cooling the reactor and increases the efficiency cf a qiven coolinq ope ation with this source.
- 18. 1. 20. 3. 2. 1. 4 RflR QCor.tainment Spray Mode/
Unde" post-accident condi+ions, water pumped f"om the suppression pool throuqh the RHB heat exchanqer may be diverted to spray header system loops located hiqh in .he drywell and above the suppression pool. This mode of operation provides the ability o reduce cnntainmen+ pressure by condensinq atmospheric steam while cool inq the suppression pool wa ter. No credit ' taker. f or spra y removal of iodines.
18.1.20.3.2.1.5 CRD Hydr au.l.ic System The operation of the CRD system was reviewed o determine scram discharge headers will contain hiqhly adioactive water if ",he followina a postula ed acciden+. Prior .o a scram the CRD housinqs contain condensate water delivered by +he CBD pump
)<hen a sc" am occurs some of this ccrdensate water from .he CRD system is discharged to the scram discharge header. After the scram, some condensate and reactor water flows +o the cram discharqe heade which fills 'n a matter of a few seconds.
Since the vents and drains in .he scram discha" qe headers a e isolated by the scram, all discharge flow then stops. Since it is not reasonable to assume tha+ siqnifican core damage occurs in the first .few econds followinq a scram, the cram discharge header will initially contain only a m'x+ure of condensate and .
pre-accident reactor water followir.q th is po .tula+ed accident.
A f er the reactor scram, the sc am discharge and instrument volumes will contain'about 700 gallons cf pre-accident water, isolated by a sinqle dra'n valve leak tes+ed to 20 cc/hr. If the ini+ial scram closed the drain valve, then this leakage is
insignifican.. comoa ed to the s c=am discharge vo'ume an d insianificant as a post-accider. t co..cern. Ef:h-: drain valve fails to close, ooerator act'on is required to rese. <<h sc ram and close t he so ft-seat.ed scram dlscha gc va l ve ~ I E th is ac<<ion is not taken or .ails to close the valve, ther. post-acc ident sou"ces can en+er the liqu' ra dwa :e system bv leakinq past <<he CRD seals. The CRD withdraw wi+h the reacto= coolant.
line does not. directly corn murica+e Tr; liqht of the ar.<<icipated small leak rate". and the lack of single failure criteria consideration requi.emen+s, .he scram discharge drain valve was assumed t;o remain closed and any leakaqe was disreqarded.
- 18. l. 20. 3. 2.,1. 6 RNCU System For a major accident with resultirq core damaae, .hc RHCU system would automatically isolate on a low reactor coolant level siqr.al and would contain no hiqhly radioactive material ~ beyond the second isolation valve. S'ce the cleaning capacity fo this system is small. it would be imoractical tc use 't for T'AX type accident recovery and i<< 's excluded from +his shieldinq review.
18.1.20. 3. 2 1.7 L quid Radwaste System
~auioment drains and compartmer.+ floor d" ains se=vic'nq ECCS systems are isolated from the reactor.. hui ld nq sump. All oipinq that may contain hiqh ac<<ivit.y po t-accid en". water is also isolated from the reactor building sump a nd "adwaste systems.
CRD system isolation is discussed ir. Suhs ec,ion 18.1.20.3.2.1.5.
Since no sianifican amounts of pos -acci rien. activity can reach the liquid radwas+e system, i+ is exclude d from this sh'eld'nq review.
- 18. 1. 20. 3. 2. 1. 8 HSIV Leakage Con+ rol Sos+em Subseauent to a oostula+ed acc dent, system ope "a+ion may heqir.
upor. actua<<ion of the manual switche= in he control zoom. This system may only be activated upon a permissive reactor pressure iqnal (35 osiq) . The method used o depressu ize ihe reactor <<o
<<his level has a large ef fect on the amount of activity ootentially available for passaqc through +his system. For examole, the HPCT system can d. piete the reactor s+eam activity considerably with only a few minu+cs ooeration. whichever depressurization method is chosen, the NSXV-LCS system remains as one that mus+ be included in the shielding review.
- 18. 1-32
SSPS-FS AP.
- 18. l.. 20. 3. 2.1. 9 Sa m2ling Syst ems Samplinq sy,tems requ'red or desired for post-accident
'c1ude "he Containment A tmos pherc;"oni t orinc System, t he Plant Vent Samplinq System and he Post-hcciden.. Samplinq Sy t em ~ Fa ch of .hese sy."'ms/s+ations may contain post-accident sou rces and is 'nrludec in the shieldinq review.
18.1.20.3.2.1.10 Standby Gas rea-.ment. System The Eeacto Building Recircula+ion system is used after an accident. This disperses airborne activity throughout the reactor building and refuelinq floor. The SGTS sys-em collects airborne activity, concentratinq haloqens with' the charcoal filters while releas'nq noble gases outside the seconda y containment. The charcoal filter is considered to be a source of.
contained activity and is included 'n his shieldinq review. The a.sumptions used in determininq this contained source are:
- 1) Drywell leakaqe at l~< per dav.
- 2) SGTS process 'rate of 1 "eactor buildinq/refuelinq floor volume per day.
- 3) 99Ã cha rcoal f ilte" ef f iciency f or halcqens. 0% charcoal filte efficiency for noble gases.
18,l,20. 3. 2.1.11 Containment Atmosphere gD".vwellg vhe free volume of the primary containment is assumed to ini.ially contain larqe amounts of pos-.-accident activity, namelv 100Ã of. the core noble gases and 25~i of the co" e halooens. Shine th ouqh the drywell wall was examined to determine the effects on "eactor building radiation levels. Besults indicate he six foot thick drywell. shield wall reduces shine to radiaticn /one I levels. Shine through penetrations presen s no additional hazard because pipinq is directed to penetration rooms where area dose ates will be dominated hy internal pipinq.
18.1.20.3.2.1. 12 Sup2ressinn Pool /lie well/
The suppression pool is assumed to initially con+ain 50'. of the core haloqens and 1'5 of the core particulates post-accident.
Shine throuqh the wetwell wall wa examined to determine the effec.s on radia ion levels in the reactor building. I,, was
SS.. S-FS AB determined that the six foot thick wetwall shield wal} reduces wetwell shine to radiation Zone I levels in the reac:or build'nq.
lg, l. 20.,'3. 2,2 1 gad iioactii ve r e Release Source e r rtt ions s e F"a The followina release fractions were used as a basis for determininq the concen~-a..'ons for the shieldinq review:
Source A: Containment Atmo.;phere: 3.00,. noble aases, 25~ ha) ogens Source B: Reactor Licruids: 100% noble qases, 50- haloqens, 1% solid" Source C: Suporession Pool Liquid: 50n halcqens, 17~ solids Source D: Reactor Steam: 100K noble gases, 255 halogen.
The above release fractions were applied to the total cu"'es available or the par+icular chemical species (i.e., noble qas, halogen, or solid) for an equilibrium fission product inventory for Suscruehanna as listed in Table 18.1-2.
The Requlatory Guide 1.7 solids releaso frac>> ion of 1% was u sed for Cs and Rb on this review. Further evaluations of t he TY. T.
"adioactivity releases may conclude that hiqher release fractions are appropria,.e. However, until. the release mechanisms and release fractions have been quantified, the existinq re qula tory quidance will. be fol'lowed. No noble gases we e include d in the suppression pool liquid (Source C) because Requlatory G u'de 1.7 has also se+ this preceden+ in modelinq liquids in the pool (See References 18.1-4 and 18.1-10) . Furthermore, cursory a naly es have indicated +hat the haloqens dominate all shieldinq requirements and tha: contributions to the total dose r ates f om noble qases are neqliaible for the purposes of shieldin q design review.
~
8,$ ,20~3,2,3 Source Term guantif ication Subsection 18.1. 20. 3. 2. 2 above outlines he ass>>a pt ions used for release fractions for the shieldinq desiqn review. These release fractions are, howeve", only the first step in modelinq the source terms for ..he activity concentrat'ions in the sys ems under
=-eview. The importan+ modelinq parameters, decay time and dilution volume obviously also affect any shieldinq analy is.
The followinq sections outline +he rationale for the selection of values for these key parameters.
- 18. 1-34
SSES-FSAR l8. l. 20. 3. 2. 3. 1 Deca/ Tlm~
"or the first stage of the shieldiaq desiqn "eview process, minimal decay Time c.pdi was used with "..he above releases. )he primary reason for +his was ~o develop a sec of accident radia" ion zone maps normal'ed to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> decay.
- 18. 1. 20. 3. 2. 3. 2 Dilution Volume The volume used for d'ution is impo tant, af fcctinq "he calcula+ions of Rose rate 'n a linear fashion. 'Che followinq d'ution volumes were used with the release fractions and decay times listed above to arrive at the f inal scurce terms for the
.h'eldinq review:
Source A: Drywell and suppression pool f ree volumes.
Source B: Reac or coolant system normal liquid volume (ba ed on reactor coolant density at z,he operatinq temperature and pressure).
Source C: 'The volume of. the reactor coolant system plus the suppression pool volume.
Source D: The reactor steam volume.
18.1.20.3.2.4 SgstemgSource Summary Core Spray Sys'em: Source C Hiqh Pressure Coolant Injection System Liquid: Source C Steam: Source D (with credit fo" steam pecific activity reduction due to +u"hine ope"at'on) 0 Reactor Core Isolation Coolinq System Liquid: Source C Steam: Source D (with credit for steam specific activity reduction due to turbine operation) .
o Residual Heat Removal System LPC3'1ode: Source C
SSES-."SAR Shutdown Cooling Node: Source P. (with credit for noble gas release during vessel depressurization).
Supn ession Pool Coolinq and Con;a'menz Spray
.".odes: So urce C
.'lain Steam Isolation Valve-Leakage Con-.rol System Souzce D (wi,.h credit for s:eam specific hctivizy reduc ion due zo RCXC .u"bine operation) .
Sa mplina Svstems Con-..a inme nt ai sample: So ur ce A Reactor coolant sample: Source 9 Plant vent sample: 1% per day Orywell leakaqe followirq the filtration by the Standby Gas ',Treatment Sys em (see subsection 18.1.20. 3.2.1.10 for discussion of SGTS source assumptions)
Standby Gas Tzeatment Sy tern Charcoal f'1ter: 1X per day drywell leakaqe (See Subsectior. 18.1.20.3.2.1.10 fo" discussion of souxce assumptions) .
o Drvwell: Source A o i<etwell: Source C For each of +hese systems, piping a soc'ated with the appropriate operatinq mode was identified on piping and irstrumenza+ion drawings ard traced throuqhout the plant to their final des+ ination.
18,1.20.3,2,5 Dose Integration Factors for Pexsonnel Cummulat've radia ion exposure to personnel in v'tal areas (continuous occ>>pancy) is determined based upo.. a maximum one year exposure period. The inteqrated doses a e modified usinq Reference 18.1-8 occuoancy factors listed below.
- 18. 1-36
SSZS-:SAR Ti me gdagsg Qccu Dance Factors 0 io 1 1.0 1 to 4 0.6 over 4 0 '
":xposures for areas not continuously occupied (frequent and in.,reqvont occupancy) must be determined cas~ by case, that is, multiPly +he task duration by +he area dcse rate at the time of exposure ~
18,1.20.3.3 Shie] Rj ng Review Methodology 18,1,20. 3. g. 1 Radiation Dose Calcu lation Yodel The p=evious sections outlined the ra" ionale and assumptions for he selection of systems that would undergo a shielding design review as well as the formulation of the souzces fo" those sys,ems. The nex. step in +he review process was to use those souzces alonq with standard point ke nel shielding analytical techniques (Ref. 18.1-14 and 18.1-15) to estimate dose rates from those selected systems.
Scattezed "adia,.ion (e.q., shine over partial shield walls) was considered but was no+ significant since the net reduction in dose is seve al o"ders of maqnitude and no vital area is separated om a hiqh activity sou ce solelv by a partial wall.
Radiation levels for compartments containinq the system." under review were based on the maximum contact dose rate for any component ' the compartment. Radiation levels in areas no+.
contain'nq unshielded sources were based on maximum dose ra.es transmitted into areas throuqh walls of these adjacent compartments'. Checks were also made fo" any piping or equipment that could dizectly contribute to co ridcz dose rates, i.e.,
pip nq tha+ may be running directly in the coz" idor o=
equipmentr'pipinq in a compartment +hat could shine directly into corridors wi.h no attenuation th"ouqh compartment walls. There is no field routed small p'inq (i.e., pip ng less than 2>> in diameter) for FCCS systems.
Dose rates are cummulative and are summed over all systems in simultaneous operation 'n most cases. Th e exce pt'n .". stea m piping for the BCXC and HPCX systems. Bo th are high pressure systems and cannot be operated simultaneo usly with low pressure systems =uch as core spray. Thi . becomes a moot point, since those steam lines are routed in well shie lded ccmpartments, causinq no appreciable personnel doses.
- 18. 1-37
SSHS-FSAP 18.$ .20.3. 3. 2 Post-Accident Padiatior Zone >ans One of the p"inc'pal products of this r eve.e w is -.he .eries of accident radiat ion zone maps (Fiqur .s 18. 1-2 to 8) . The zone boundaries used 'n the maps are defin ed in Ta tie 18.1-3. The zor.e maps present .he calculated dose rates a t, one hou= after t he accident due to the sources described in Sub s ection 18. l. 20. 3. 2.4
-'n various areas of the plan-, si:e. The p:in cipal sources o radiation in each area are identified in Tabl e 18. 1-5.
The dose rates presented do not include cont"ibutions from normal operating sources wh'ch may be conta'ned in the plant a+ the t'me of .he accider.t since these contributior.s vill be minor outside of well defined and shielded areas. They al=o do not include do"e rate contributions due to potential airborne sources resultina from eauipment or drywell leakage.
The zone maps were used to de.ermine the accessibili+y of vital areas described in Subsection 18.1.20.3.3.4.
18,>,20.3.3.3 Personnel Radiation Exnosure Guidelines In order that doses .o occupied areas take on meaningful p"oportions, it is necessary to establish exposure qoals or auidelines. The aer.eral desian basis for +hese quidel'nes is 10CFB50, Appendix A, GDC 19. That mate 'al addresses control room habitabili y, includinq access and occupancy under worst case conditions. "...xposu es are not to exceed 5 rem whole body, or its equivalent to any part of the body, fo" the duration of any postulated accident. GDC 19 i= also used to govern desiqn bases for the maximum permissible do aae to personnel perfo" ming any task required post.-acciden+. These requirements translate "ouqhly in+o the objectives to be met in -h>> pos-.-acc'dent review as given below.
Had ~ at ion Fxposure Guideline ~
Occupancy Dose Rate Objectives Do e Objective Con+inuous 15 mB/hr 5 Rem for dura+ion Frequent 100 mB/hr 5 Rem for all activities Infreauent 500 mB/hr 5 Hem per activity Accessway 5 p/hr Included in above doses
- 18. 1-38
SSES-FSAR 18.l.20.3.3.4 Vital Area Identification and Access
- 18. 1. 20. 3. 3. 4. 1 Vital A-;. ea Clar i f icat ion Vital areas are those >~which will or nay r-quire occupancy to permit an operato to aid in the mitigation of or recovery from a n accident". Reference (38. 1-16) f urther def ines reco very from an accident as, "wh n he plan+ is in a inaction sa'f~- and stable condi.ior.." "This may ei" her be hot or cold shutdown, depending on the s'tua+ion." The 10 CPR 73.2 def of v'al area sha ll not. apply here.
For the purposes of this study, +he evaluat'on to determine, necessary vital areas considers all of those listed in Reference (18.1-3) . Upon examination several plant areas we e de+.ermined not to be vital. 1nstrumen+ panels were excluded because essential equipmen. con+rol and aliqnmer t has been es+ab'ished irr the con+rol room and requires no local actions. The radwaste control room is excluded because 1) no local actions are "equired to prevent spread of. pos.accident sources into the liquid radwaste system; 2) gaseous radwaste processing is not required, a.nd; 3) activity sources early in the post-acciden. "ransient are much too hiqh to be. effectively processed through .he liquid and eventually sol'd radwaste systems. Also exc'udcd a" e the po .t-LOCA hydrogen control system and the containment isclation reset control a"ea (which are operator actuated from the main cont ol room) . Lastly the emerqency power supply (i.e., diesel genera+o"s) was excluded since system initiation comes f"om +he contxol room and requires r.o local actions.
The resultinq list of areas considered vital for post-acciden operations at. Susquehanna appears in Table 18.1-4. Hote that security facilities are included as vital axeas with regards to maintaininq plant security.
- 18. 1. 20. 3. 3. 4. 2 Vital Area Access Those operator act'ns required pos".-acci den-; we=e reviewed to assur~ +hat first priority safety act'ons can be achieved in the postulated radiation fields. This review assures that access is available and reguized operatox actions can be achieved.
Ingress and eqress area dose rates to those vital areas identified in Table 18.1-4 were examined to ensu e compa.ibility wit.h the area being acce sad.
Plant effluent monitorinq s ations are loca.ed at five (5) plant ver.ts: two (2) for the Reactor Building, t wo (2) ~or the Tu=bine
- 18. 1-39
SSES-FSAR Building, and one (l) for the Standby Gas Treatment System. The reactor building mon'tors a=e automatically isolated pos"-LOCA and will contain no post-acc" dent activitv. The SGTS effluent
-.ample sta tion will contain po.,t-accident act.iviry ' sample ca tridqes: one (1) volume tric ar.d one (1) charcoal fil "er. The samples a" e locally shielded and p esent no acces- p=oblems in the area of the ta" ion. However, transporta+'on ard handlinq of, the filte= cartridges will:equire local shiel,".irq.
The Turbine Building Plant Vent Sample Staon (PVSS) may also contain post-acciden act ivity. Dose, i f any, w'l he of. a lower magnitude than that of the SGf'S ef flue."." fil.es because of en v ironmen ta 1 dispersion ar.d re-en t ry to the Vu rhine Du ild inq ven+ilation system. Xn the worse case, the Turbine Building PVSS doses will be much lower than those of the SGTS. In the best case, control room personnel may shut down the Turbine Puildinq HVAC system (which is non-safety related) . In this case, the Turbine Building PVSS may be void of po t-accident ac.ivity.
18.1.20.3.4 Results 18.1.20.3. 4.1 Radioactive Decay Rffects Pesults of the radiation level evaluation fo" the shieldinq desiqn rev'ew are presented 'n F'gures 18.1-1 to 8. Table 18.1-5 iden+ifie the sources contributinq to dose rates in each of the
~
plant areas shown on those figures. This ".able can be used in con junction with the decay curves (Fiqures 18.1-9 and 10) to estima+e radiat'on levels at, imes othe" than one hour. The procedure for times less than one day, 's tc multiply the radiation level (i.e., radiation zone limit) by:he decay factor qiven in Figure 18.1-9. For times qrea+e= than one day, it 's necessary to multiply by the decay factor in Figure 18.1-9 at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and by the decay factor in Figure 18.1-20 at the des ".erl decay time. This procedure 's conservative or areas in which the sources are shielded because it doe r.ot riqorously take in+o accoun +he sof+eninq of the, energy spectrum an consequent increase in attenuation for lonqer decay times. A deca y curve for source D, reacto" steam, is not included because the depletion effects due to steam usaqe by HPCI or PCIC removes much of this source shortly after the accident. In addition, HPCI and RCIC pipinq containinq source D is run in shielded cubicles and does not contribute significantly ou side those cuhicles.
- 18. 1-40
SSES-FSAP 18.1.20.3.4.2 Integra ed Personnel Fxnosures Personnel integrated exposures in cortinucusly cccupied areas we e calculated based on 100'ccupancv fo= zl.e first day, 60",o occupancy <rom day one throuqh four and 40K cccupancy for the duraiion (1 yea") . These calculations showed t h;.t pe sonnol exposures would be w'hin +he desiqr. objective of 5 R. m.
Expo,ure in Zones I, II -and III of the ccntrcl structure are 0.24, 1.6 and 3.1 Bem, ospectively. These doses dc no include
+he shieldina effects of interior walls, eau'pment, etc.,
~he:efore they:epreso.nt the maximum doso to control bu.'- ldinc oersonnel due to contained sources. Pe sonnel doses .o the North Gate House (ASCC) and Security Control Center from con.ained sources were found to he insiqnificant (i.e, ( A.l Bem) . These areas are a minimum of 300 feet from the reactor building whose walls are a minimum of 2.5 feet, of concrete.
Personnel doses at the Post-Accident Sample Station, Chemistry J.ahoratory, and Plant Vent Sample Station ar. calculated based on an es+imated task duration at specified .'es post-accident for a one person task fo ce (Befer to Table 18.1-4) 18.1.20.).4.3 Peactor Build'na Accessibility T?!e results "'show +hat the reactor bui'ding will Le generally inaccessible for several days af.er the accident due to contained radia.ion sources. High radiation levels can be expected a+
Elevation 645'-0" (Figure 18.1-3) eqardless of 'hich system( )
is (are) in operation. Badia tion level s at Elevation 719'-0" (Fiau "e 18. 1-5) and above are expected o aenerally be within Zone IV limi:s if the core spray and PHB containment spray systems have not been operated follnwina the accident. Thi" is because -',
hese are the only unshielded post,-acciden+. sys em sources at hese elevations. Other system sources are contained in shielded cubic les.
Exceptions to the e general Zone IV levels are areas in the vic'nitv of reactor coolant and conta'nment a+mosphere sampling lines which are routed to the reac"or building sample station at Elevation 779'-0". The dose rate 10 feet from the reactor coolant sampling lire one hour afte" the pcstulared acciden. may exceed 100 R/hr.
Besults for contained radiation sou ces show +hat t.he vital aro.a in the React. or Du.ildinq is accessible pos"-accident.
- 18. 1-41
SS r,S-FSAB
- 18. l. 20. 3. 4. 4 Control Building Accessibility oi Besul"s for contained radiation sources show tha.
the control tructure are accessible post-accident.
v'l areas in 1'R.l.21 POST-ACCID.".N'7 SAHPLING /II.R.3$
18.1,21.1 Statement of Reouiremert A desiqn and operational review of the reactor coolan+ and containmen atmosphere samplinq line systems shall be perfo" med to determine the capability of personnel tc promptly obtain (less
..han 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) a sample unde" accident conditions without incu "ing a radia+ion exposure to any individua 1 in excess of 3 and 18-3/4 rem to the whole body o" extremities, respec-.ively. Accident conditions should assume a Hequlatory Gu'de 1.3 or 1.4 release of fission products. If the review indicates that personnel could not promptly and safely obtain the samples, additional design features or shieldinq should be provided to meet the cr'teria.
A desiqn and operational review of +he radiological spectrum analysis facil'ties shall be performed to determine the capability to promptly auantify (in less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) certain radionuclides that are indicators of the deq ee of core damage.
Such radionuclides are noble 'ga es (which indicate cladding
.failure), iodines and cesiums (which ind'cate hiqh fuel tempera+ures), and nonvola+ile iso opo.,s (which indicate fuel meltinq). The ini.ial reacto" coolant spectrum should correspond
+o a Regulatory Guide l.3 or 1.4 release. The review should also consider the effects of direct radiation from piping and componen+s in the auxliary building and possible contaminat'on and direct radia+ion from airborne ef fluents. If the review indicates +hat the analyses required cannot be perfo"med in a prompt manner wi+h existinq equipment, then design modifications fo equipment p ocurement shall be undertaken to meet the criteria.
In addition to the radioloqical analyses, certain chemical analyses are necessary for monitorina reactc ccndi+-'ons.
Procedures shall be provided to perform boron and chloride chemical analyses assuminq a hiqhly radioactive initial sample (Reaulatory Guide 1.3 or 1.4 source term). Beth analyses'hall be capable oX beinq completed promptly {i.e., the bo on sample analy"is w thin an hou" and the chlorid- sample analysis wit.hin a shift) .
- 18. 1-42
The follow'q '+ems are clarifications of requirements identified in >.'UBHG-0578, l>URH6-0660, or t he Seotemhe 13 and october 30, l if t 979 c 1 ar ica".. ion le te r s.
(l) :he licensee sha ll have,he capab'lity to promptly ob.ain reac+or coolant samples and containment a "mo. phere sample .. The combined t'me allotted for sampli,.q and analysis hould b-.- 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> or less from
-. he time a decision is made:o take a sample.
(2) The licen. ee shall establish an cnsite radiological and chemical analys's capabili:y to provide, within the 3-hour +ime f"arne establi hed above, q>>antif'cat'on of the following:
(a) cer+a'n radionuclides in the reac.or coolant and containment atmosphere that may be indicators of
.he deqree of core damage (e.q., noble gases; iodines and cesiums, and nonvolatile isotopes);
(b) hydroqen levels in the containmen".. atmosphere; (c) dissolv d gases (e.q., hydrogen), chloride (time allo+ted for analysis subject to discussion below), and boron concentration of liquids.
(d) Alternatively, have inline monitoring capabilities to pe form all or part of the above analyses.
(3) Reactor coolant and containment atmosphere samoling durinq postacciden condi ions shall not require an isolated auxiliary sys.em (e.g., the letdown sys+em, reactor water cleanup system to be placed in opera tion in order to use the sampling system.
(4) Pressurized reac:or coolant samples a re not required '
the licensee can quantify the amcunt of dissolved gases with unpressurized reac+or coolan" samples. The measurement of either total dissolved aases o" hydrogen qas in reactor coolant samples is considered adequate.
treasuring the oxyqen concentration i~ recommended, but is not manda.ory.
(5) The t'me for a chloride analysis to be performed '
dependent upon two factors: (a) if wa er is seawater or brackish wa+.e" and (b) the plan+'s coolant if there is only a sinqle barrier between primary containment systems and ..he cooling water. Under both of the above conditions the licensee shall provide fo a chloride analysis within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of he sample being -.aken.
For all other cases, the licensee shall provide for +he
SSFS-FSAR analysis to be completed within 4 days. The chloride analysis does not have .o be <'ne onsite.
(6) The desiqn basis for plant equipment for reac or 4 coolan+ and con'ainment atmosphe e sampl'nq a n>etwell and Drywell Atmosphere Gas samples can be ob.ained f"om + wo sepa" ate a eas in bo'h the drywell and wet well. The sample lines tap in.o +he containment ai= monitor'nq sys,em samp1e l i nes ou-.. = ide o f prima ry contain ment and a f ter the second c cata i nme n t
'olation ralve The wo drywall sample +. aps are on t.he hiqh point l'e, samp1inq at elevation 790~ and the midpoint line, samp1inq at eleva. ion 750'.
b) Secondary Containment Atmosphere A sample line was installed to allow sam plinq of the secondary containment atmosphere.
c) Reac+or Coolant and Suppre,ssion Pool Liquid Samples.
~<<hen the reactor is pressurized reactor coolan+ samples will be ob+ained from a zap of f the jet pump pressure instrument system. The sample point is on a non-calibrated jet pump instrumen. line ou+side of prima y containment and after the excess flow check valve. This sample point location is preferred over the normal reactor sample po'nts on the reactor water clean up system 'nlet line and recirculation line since the reactor clean-up system is expected to remain isolated under accident condition., and i+ is possible that the recirculation line containinq the sample line may be secu"ed. The jet. pump ins+rument line has been determined to be the optimum sample point for acciden: condit iors since: l) +he pressure taps are well orotectecl from damaqe and debris, 2) if the reci=cula+ion pumps are secured,
+here is normally excellent circulation of the bulk of the coolant pas these taps {natural circulation), and 3) the taps are located sufficiently low ~o permit sampl'q at a reactor water level which is ever. below the lower coze suppor pla A sinqle ample line is also connected to both loop" in +he BHR system. The sample lines .ap of f the hiqh pre sure swi+ch irstrument lines comirq of" he common section of the BHB sys" em return line. This =;ample point provides a means of obtaininq a reactor coolant sample when the reactor is not pressurized and at least one of the PHB loops is opera+ed in the shu+down coolinq mode. Similarly, a suppression pool sample can he obtained from an BHB loop lined up in the supp ession pool coclinq mode.
- 18. 1-Q7
SSFS-FSAf3 18,1.23,,3.2.2 Isolation Valves and Sample Lin~s Containment isolation for +he drywell ar.:3 wetwell gas sample 1'nes is provided by the existinq cortainment a'r moni=orinq sample li..e isolation valves. he jet pump instrument sample line containmen+ isola ion is provided by an exis+ irq isola.ion valve and excess flow check valve upstream o the sample taD.
All qas .-.ample line... f om;.he sample taps:o and includi ng the first flow cortrol va3ves are seismic cateqory 1 except for the seccn<3ary containment sample line which has no control valve before it en+ers the sample panel. The sam "le lines from the RHR syst em are seism'c category 1 thro>>qh bc.h syst..m isola-ion valves and a flow res.rictina orifice. The sample line from the
]et p>>mn instrument system is seismic cateqo" y 1 to the flow control/isolation valve. All containment isolation valves ups+ream of +he sample . ap" can be cverridd -. f rom the control room. All isolatior. and control valves shown in Fiqure9.3-9a which are wi. hin the 0 bounda y are controlled Ly a single permissive swi ch 'n the control room and individually controlled at +he samplinq control panel located ad jacent to the sample station.
The soleno'd isolation and control valves which are part of the post accident sample system to +he Q bounda..y will b~
environmentally qualified. The qas sample lines are heat traced to prevent precipitation of mois+ure and the re ultant loss of iodine in the samnle lires.
- 18. l. 21. 3. 2. 3 Pinina Station The pipina station, which is installed with'n the reactor buildinq, include s mple coolers and control valves which iletermire +he liquid sample flow path to +he sample s~ation. The location fo the pip'nq station 's shown ir, "-ia>>=e 1.2-20.
Coolinq water comes from +he Reactor Buildinq Closed Coolinq l<a. e r S yst e m.
- 18. 1.21. 3. 2. 4 Sample Station and Control Panels The location of the sample station, control panels ard associated equipmen is shown in Fiqure 1.2-4. The sample station consists of a wall mounted frame and enclosures. Included wi+hin the sample station are equipment trays which contain modularized liquid and qas samplers. The lower liquid sample pcrtion of the sample station is shielded with 6 inches of lead brick, whereas the upper qas,ampler has 2 irches of lead shieldinq. Tne control instrumentation is installed in two control panels. One
- 18. 1-48
SSES-FSAR of these. panels con.ains the conductivity, and radiation level readout... The other control panel contains the flow, pressure, and tempe=ature indica':.o s, and vario>>s cont ol valves and swi.ches. A qranhic display directly below the main cortrol panel which shows -he status of ".he pumps and valves at all times. The panel also indicates +he relative posi.'on of the oress>>re qauqe" and other items of ccncern zo the opera~or. The use of th's nanel will .mprove ope"ator comprehersion and assist in tro>>hie shoot'nq operations. The various sample 1'ne- ard eturn lines enter .he sample station enclcsure (which i' mounted flush aqainst the secondary containment wall):hrouqh the back by way of a penetration in the s~eam ?unnel wall.
- 18. 1. 21. 3. 2. 4. 1 (."as Sa m ol er The qas sample sys+em is desiqned to opcrateat. pressu 'e rarging from sub-atmospheric to the desiqn pressures of the primary containment one hour after a loss-of-coolant acciden . The aas samples may he passed throuqh a part iculate filter and s-'ver zeolite cartridqe for determination of particulate activity and
~otal iodine ac+ivity by subsequent qamma spectroscopic analysis.
A radiation monitor is mounted close to the filter tray to measure the activity buildup on the cartridqes. Al ternately," tho.
sample flow bypasses the iodine sampler, is chilled to remove moisture, and a 15 milliliter grab sample can he ..aken for determination of qa.,eous act'vity and for gas ccmposition by gas chromatoqraphy. The qas is collec+ed in an evacuated vial using hypodermic needles in a manner analogous to the normal off-gas samples. Hhen purainq the drywell arid wetwell qas sampl~ lines o ob+ain a representative sample, the flow is returned .o the wetwi ll; howeve", during pu"qirq of the secondary containment line and when flushinq the ample panel lines with air o nit=oqen, flow is returned:o secondary containment. The sample ta.ion desian allows for flushinq of the entire sample panel line from ..he fo>>r posi+ion selector value .hrough the needles wi+h either air, nitrogen, or the aas to he sampled. This capability will minimize any possible cross contamination between
'the various sam ples.
18, 1. 21. 3. 2. 4. 2 T,iguid Sampler The liauid sample system is desianed to opera+e at pressures f rcm 0 to 1500 psi. The design p>>rqe flow of 1 qpm is sufficient to maintain turbulent flow in the sample line and serves to alleviate cross contamination hetween sample . The purge flow is returned to the suppression pool. The liquid sampling system is desiqned to allow routine demineralized water flushinq of the system lines from a point between the two ccolers in the pioina
- 18. 1-49
SSHS-FSAR station through the samplinq needle=-. Using the hydro-test connection which is outside he sample panel, it is also possible to backf lush all the liquid sample lines through the sample tap noint. This will allow for clearir'q of pluqqed lines. All liquid samples are taken in+o 15 milliliter septum bottles mounted on samplinq needles. Tn the normal lineup,:he sample flows throuqh a conduct'vity cell (O.l to 1000 micromhos/cm and throuqh a ball valve bored out to 0.10 milliliter volume. After flow throuqh the sample panel is established, the ball valve is rotated 90o and a syringe, connected to a line external to the panel, is used to fl>>sh the sample plus a measured volume of diluent (generally 10 milliliters) through +he valve and into the sample bottle. This provides an ini+ial dilution of up to 100: l.
The sample bottle is contained in a shielded cask and remotely positioned on the sample needles throuqh an opening in the bottom of the sample enclosure. Alternately, the sample can be diverted throuqh a 70 milliliter bomb to obtain a larqe pressurized c'ulated and volume. This 70 m'lliliter volume can be depressurized into a qas samplinq chamber. A 15 milliliter qas sample can then be obtained through a hypodermic needle fo qas chromatoaraphic and radioisotopic analyses of the dissolved gases associated with the 70 milliliter liquid volume. Ten milliliter aliquots of this degassed liquid can then be taken for off-site (or on-site dependinq on activity level) analyses which reguire a relatively larqe undiluted sample. This sample is obtained remotely using the large volume cask and cask positioner throuqh needles on the under ide of the sample station enclosure.
- 18. l. 21. 3. 2. 4. 3 Sample Station Ventila tion The sample s+ation enclosure is vented to secondary containment via the main steam line tunnel. Ventila+ion is motivated by differential pressure between the turbine and reactor buildings.
The ven+'lation rate equired for heat removal durinq operation is about 40 scfm. The ventila,.ion duct is sized for less than 100 scfm a 1/4 inch of water differen>ial pressure when the enclosure .is opened for maintenance. Standby air flow will be about 3 scfm and can be reduced by tap'ng all openings. A pressure qauqe is attached to the sample station enclosure to mon'tor the pressure differential between the enclosure and the general amplinq area in the turbine building. This w'll assure the operator that airho ne activity in the sample enclosure will be swep+ into secondary containment.
1.8. 1. 21. 3. 2. 4. 4 Sample St at ion Sump The sample sta+ion is provided with a sump at the bottom of the sample enclosure which will collect any leakage within the
- 18. 1-50
SSES-FSAB enclosure. This sump can be isolated and pressurized, discharqinq into the sample station liquid return line +o and hence into the suppression pool.
- 18. l. 21. 3. 2. 4. 5 Sample Handling Tools and T" anspor".
Containers Appropria+e sample handlinq cools anl1 transporting casks are provided. Gas vials are ins+alled and removed by use of a vial positioner through the front of the gas sample . The vial is then manually dropped into a small shielded cask directly from the positioninq tool. This allows +he operator to maintain a distance of about three feet from the unshielded vial. This cask p ovides about 1-1/8 inches of lead shieldinq. A 1/8 inch diameter hole is'rilled in the cask so tha. an aliquot can be withdrawn from the vial with a qas syringe without exposing the analyst to the unshielded vial.
The particulate and iodine cartridqes are removed via a drawer arranqement. The quantity of activity which is accumulated on the cartridge is con+rolled by a combination of flow orificing and time control of the flow valve opening. In addition, the deposition of iodine is monitored during samplinq usinq a radiation detector installed in the sample staticn next to the car+ridqe. These samples will hence be limited to activity levels which will no+ require shielded sample carriers.
The small volume (diluted) liauid sample cask is a cylinder with lead wall thickness of about two inches. The. cask weiqhs a.
approximately 65 pounds and ha a hand'le which allows it to be carried by one person.
The 10 milliliter undiluted sample is taken in a 700 pound lead shielded cask which is transported and positioned by a four-wheel dolly. The sam'pie is shielded by about 5-1/2 inches of lead.
$ 8.$ .21.3.2.4.6 Sam@le Station Power Supple The PASS isolation and control valves, sample sta+ion control panels, and auxiliary eguipment are connected to an Ins rument AC Distribution Panel which is powered from an Engineered Safeguard System (FSS) bus. Followinq a loss o f of f-site power, the ESS bus is powered from the on-site diesel qenerators and backed up by batteries. The Feactor Buildinq Closed Cooling Rater System, which is needed for the sample coolers, is also powered from the emerqency diesel qenerators followinq a loss of off-site power.
Compressed air for the air-operated valves comes from compressed
SS ES-FSAB air cyl'nders, thus eliminatinq any dependence on "he plant rompress~d air system.
18.1.21.3.3 Description of Sample Preparation/Chemistry and Nuclear Counting Facilities After the samples are obtained from the sample station, they will be transported to a sample propara+ion/chemistry area. There they will be diluted as necessary and appropziate aliquots taken for chemical and radioisotopic analyses. The radioisotopic analysis will be done in a separate countinq area where backqround radiation can be kept to a minimum. Two different fac'lities will be available to plant personnel tc perform the above tasks. The primary facility is the existinq chemistry laboratory and countinq room which is at elevation 676', the qround level of the control tructure. A backup sample preparation/chemis+ry area and countinq ream is provided in the Emerqency Operations Facility (EOE) which is located 2500 feet south-west of the control structure. In addition to these on-site and near-site facilities, which a e in ended to handle the qas samples and the dilu"ed liquid samples, prior a anqements have been made with an independent off-site laboratory for analysis of the undiluted 10 ml liquid samples.
18,1,21,3,3.1 On-Site Chemistry Lahoratorg and Counting Boom The plant shieldinq study results, pre. ented in Suhsec ion 18,1.20.3, show that followinq an accident, tho. chemistry laboratory will be a Zone II area (<100 mR/h) . Therefore, +he existinq facilities will be accessible at lens+ foz intermit+ent use followinq an accident. The mos. direc- route between the ample station and these facilities is throuqh areas of the turbine buildinq which should be Zone I areas (<15 mR/h) followinq an accident. The chemistry labcratory is equipped to provide the capability to handle the qas samples and the 0.1 ml diluted liquid samples. The maximum activity of these samples vill be 0.7 Ci the fractional and 0.3 Ci, respectively, usinq one-hour decay and releases of core inventory specified by llUHEG-0578 (see Section 18.1.21.3.5).
The laboratory maintains a dedicated inventory of items such as lead bzicks fo shieldinq, qas syringes, qloves, reaqents for analyses, etc.. which will be needed in case of an accident. The laboratory is equipped with a qa chromatoqraph, pH mete",
conductivity meter, turbidimeter and other instrumentation needed to perform the required analyses. This equipment however, may not be dedicated exclusively to post-accident analysis. Supplied air or self-contained breathinq masks will be available in the
- 18. 1-52
SSES-FSAB even". of high act'vi.y levels in the ventilation supply or acc'dental sp'lls in the laboratory.
The existinq countinq facility'oca ed adjacent to the chemistry laborato"y is well equipped to handle t.he gamma spectra analyses reauired f or post-acciden. samples. The ccuntinq zoom is eauipped with two Ge (Li) detectors with four inch lead shields connected to a computer based analyzer system. The system has automatic peak sea ch and isotope identificat'on capabilities.
The Ge(Li) detector and shelf assembly in the lead shield can be well i. olated and the capability to purqe the volume w'hin the shieM with compressed qas will be provided. This will help prevent atmospheric noble qas activity released during an accident from swampinq the detector.
18.1.21.3.3.2 EOF Sample Preparation/Chemistry and Coun.ing Facilities The sample preparation and countinq rooms located in the near-site EOF serve as backups to the on-site facilities. The EOF is 2500 fee+ from the control structure and is directly accessible from the site by road. Travel time from the sample station to the EOF is less than 30 minutes. The backup facilities will be act'vated whenever the on-site facility beccmes inaccessible or if additional lab space or counting equipment is needed o handle
+he increased work load in the on-si+e facil'ty resulting frcm an accident. The sample preparation/chemis..ry room is furnished with a radicisotope labo atory hood, about 14 feet of laboratory cabinets and benchtop working space, a small sink dzaininq to a removable carboy, and at least a 5-qallon supply of demineralized water in plastic carboy mounted on the wall over the sink. The hood is eauipped with a HEPA fil+er unit. Although some analytical instrumentation may he kept in this room, i. is not !
meant to completely duplicate that in the on-si; e laboratory.
However, the facility is fully equipped tc handle the neces"a"y dilutions and manipulations to p epare samples which come directly from the sample station for gamma spect oscopic analysis. Additional instrumentation foz the reauired chemical analyses will be brought from the on-site labora ory as needed.
Chemical reagents, qlassware and ot.her miscellaneous equipment will be stocked in the facili+y. A supply cf lead bricks is also kept in th's room for use as temporary shieldinq. A lead brick cave foz storage of samples is also provided. 'The EOF counting room will contain as a minimum a high resolution gamma-ray spectrometer system. The system is capable of characterizing and quantifyinq the gamma activities of reactcr coolant and containment atmosphere samples. The intent is to make this system similar to the on-site system.
- 18. 1-53
SSES-FSAB T he EOF has its own diesel generator which will be capable of supplying the electrical power needs for the fac.'ity during loss of off-site power.
- 18. 1. 21. 3. 3.
~ 3 Arrancrements fo= S t e Analyses Off-Site key part of +he SSZS approach to post-acc'dent sampling is the establishment of prior arrangements with an off-site independent laboratory for confirmatory and supplemental analyses. The capability of the off-si.e laborato y will also he used =o meet he reauirement for chloride analysis. The reason for usinq the off-site laboratory for chloride and as a backup for other analyses is to prevent having to handle and analyze undiluted coolant sample which may have activity levels in the curie per milliliter range. The on-site and EOF facili+ies are not des'neR to handle sources of this maanitude. The analyses of undiluted samples can be done in a safer manner by laboratories with facilities and personnel spec'fically built and trained to handle hiqh-activity sources. The following is a description of the siqnificant features of the off-site laboratory:
formal mechanism exists to allow for initiation of pos+-
acc'dent services at any time (24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s/day).
Hritten procedures must be controlled anR maintained for each of the analyses described in Table 18.1-6. The analysis procedures must be qualified for use at the activ ity levels qiven in the table. This requirement may be satisfieR by referencinq the appropriate li-:.erature, by calculations, or by undertakinq a testinq program.
c) Laboratory equipment and facilities fo .he required analyses must be available and ma'n+ained in working order such that analyses may be completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the
=eceipt of the sample.
R) Provision will be made for the practice or exercise of each aspect of the off-site analysis work at the option of the utility.
Equi@ment will be available fo" the timely transmission and receipt of information and results (telecopier and/or telex)
- 18. 1-54
SSKS-FSAR
- 18. 1. 21. 3. 4 Summa~a Desc"iption of Procedures
- 18. 1, 2g, 3. 4.1 Sample Collertion and Transcort p ocedures After a Recision is made to obtain a sample, the des'gnated sample station operators (2) will proceed +o +he sample station with the neressa"y equipment.
Since all the pos -acciden+. sample lines (except for the secondary containmen atmosphere) tap off-1'.nes which are isolated following a containment isolation sianal, the sample station operator must confizm.with the ccntrol "oom +hat the necessary isolation valves are open. (A telephone extension to the control room vill be installed close to tho sample control panels for this purpose). The control room must also activate the ~~Accident Sample Station Permissive Switch" to allow the sample station operator control of the "isolation and control valves" which are part of the post-accident sampling system.
After switchinq the "Master Shu+off Valve Control" to the "open" position, the operator is ready to open the valve(s) controlling flow from the desired source to the sample station. Af ter openinq .he necessary control valve (s), the operator goes to the "sample station control panel". This panel controls the valves which ar~ part of the pipinq s+ ation and those in the sample station enclosure in the turbine building.
Followinq a series of presamplinq checks and procedures includinq: adjustment of the enclosure damper .o insure adequate roolina, checks of demineralized water and nitrogen supplies, flushinq of system with demineralized water, dra'ning the trap and sump, etr., the system is ready fo" obtaining the samples.
18,$ .21.3.4.1.1 ProceRure for Obtaining Gas Sample A standard 14.7 milliliter off-gas vial is placed in the gas vial positioner and inserted into the qas port on the front of the sample station. The desired sample location is selected by switch and the qas is circulated until the sample lines are flushed out with the qas being sampled. She vial and a small volume of tubing remains unflushed; however, the vial and this tubinq volume are then evacuated. The sample is then drawn into
+he vial by pressing and holding a pushbut+on switch. Ifofcross-con.amination is suspected due to incomplete evacuation the vial, the evacuation and fill sequence can be repeated usinq air or nitrogen flush before takinq +he f iral sample, or the sequence can be repeated vith the desireR sample qas until the operator is assured that he has a representative sample. Fcllowinq an ai or
- 18. 1-55
SSES-FSAH nitrogen purqe of the sample lines, the qas vial positioner is then removed from the port and +he v'al inserted into the qa=
vial cask. The lenqth of he vial positicner allows the operator o r main about three feet from the vial during this operation.
The. cask has a 10-inch carryinq handle and can be easily carried by one person down the stairs in the turbine buildinq to the chemistry labo ra tor y.
- 18. 1. 21. 3. 4. 1. 2 Procedure for Obta'ning an Iodine Pa"ticulate Sa mp3.e.
'he il desired f ter cartridge (s) are placed in to a cartr idge retainer which is placed into the gas filter drawer. This drawer slides into an openinq in the front, of the sample station enclosure. The appropriate critical orifice is also chosen and placed in the cartridge retainer. This will determine the flow rate throuqh the sampler and +hereby control the amount of activ'ty deposited on the filters. The operator then selects a sample location and flushes the sample line except for a short piece of tubinq qoinq to the sample drawer. However, this line, can be flushed with air or nitroqen prior to sampling if cross-contamination between samples is suspected. In addition, as part of the normal samplinq procedures, this line is flushed with air or nitroqen after completion of each sample sequence and should therefore be free of contamination for the followinq sample. The operator has the option of usinq an automatic timer o obtain amples with collection times between 0 and 30 seconds or of manually timinq the sample for lonqer collection +imes. Afte" startinq the sample collection sequence, the operator will be able to follow activity buildup on the f ilters by observing the radiation level readout on the cont"ol panel from the probe inserted nex+ to the cartridges in the gas sample panel. After sample collection is completed, the cartridges are evacuated usinq the vacuum from the qas pumos and then flushed with air or nitrogen to remove the noble gases. The filter drawer is withdrawn and the ca" ridge retaine" with filters is placed in a plastic baq. The baq is then closed, and dependinq on the measured dose rate, it is carried by hand or attached "o a pole and carried to the chemistry laboratory. No shielding cask is provided for these samples ince it is possible to regulate the amount of activity deposited on them. In addition, for ease of countinq, it is desirable to keep the activity levels on these samples low.
18, 1. 21. 3. 4. l. 3 Procedure for Obtaining a Diluted Liquid Samole A 15 milliliter sample bottle with a neoprene cap .is placed in the small volume cask which is then placed into a positioner
- 18. 1-56
SSHS-FSAP attached to the sample station support frame. The sample needles are exposed by pull'q out the lead shielding drawe" ur.der the sample sta .ior. enclosure. The cask holding "he'ample bottle is then swung into position under the sample station and the sample bottle raised intc position so the needles penetrate the neopren.o.
cap. After aliqninq the p"oper valves, . he sample lines from the selected source throuqh the piping staticn are flushed with return flow to the wetwell. Afte" these lines are flushed, the bypass valve in the. pipinq tation is clcsed and the "ample flows to .he sample station hrouqh the cal'brated volume sample valve and back to the wetwell. After sufficient flushing, the calib"ated valve 's rotated 900 into aliqnment w'th +he line to the sample bottle. A syringe filled with uo to 10 ml of demineralized water is connected onto a line at the front of the sample station and this water is injected to wash the sample captured in the ball valve into the sample bottle. The syringe is then removed, filled with air, e-a+tached and the ai-injected to force out all water remaininq in the line through the sample needle and into the sample bottle. The rinsing action of the water followed by the air purqe of. the line should reduce cross-contamination between different samples. The calibrated sample valve is returned to the purge positicn and the sample l.ines, from the second cooler in the pipinq station, through the sample valve and back to the suppres or pcol, are rir.sed with demineralized water. The operator then returns to the sample s+ation, remotely lowers the sample bottle into the cask, screws a top pluq with carrying handle into the cask. The cask is then carried down the stairs to the chemistry laboratory. Although one person can carry the cask, a pole with a hook in the middle will be available to allow two people to carry the cask more easily.
18.1.21.3.4.1.4 Procedure for Obtaining a Large Liquid Sample
/undiluted} and/or a Dissolved Gas Sample A standard off-qas sample vial is placed in the qas vial positioner and inserted into the dissolved gas sampling port on the f ont of the liquid sample panel and a 15 milliliter sample bottle is placed in the large volume sample cask. The ample cask is positioned unde" the sample enclosure using a four-wheeled cart. The cask is raised into position and the sample bottle raised out of the cask and onto two needles usinq a remote mechanism. Shen the cask is p operly oositioned, the operato s will be shielded from the sample durinq all subsequent operations. After attaininq the prope" valve lineup, the sample lines are first flushed through the pipinq sta.icn and then through +he sample station lines including the 70 milliliter hold up cylinder and qas breakdown circulation loop. After completinq the flush cycle a f'xed volume of the pressurized liquid is isolated.and a measured amount of a tracer gas is injected. The
- 18. 1-57
isolated volume is hen depressurized by openinq a valve to a previously evacuated 15 mill'i+er qas collection chambe". The operator now has the option of ei.her collectinq <<he dissolved qas sample in an evacuated vial or releas'ng i.. -..o the suppression pool atmosphere. If a dissolve.l theqas same sample is collected, it is handled and transpor+ed in manner as the containment, qas sample discussed previously. The operator also has the op ion of collectinq a 10 millilite sample of the degassed liauid or allowinq it to he flushed tc the suppression pool durinq the subsequent demineralized water flush cycle. If a larqe volume sample is desired, it is drawn . nto the evacuated 14.7 millilite sample bottle. To minimize cross-contamina ion, the system can be cycled several times through all the ahove steps before takinq the final larqe volume sample. The dissolved qas and liquid sample system is then flushed wi+h demineralized water to minimize radiation levels while removing samples from
+he station.
The sample bottle is then remotely lowered from .he needles into the shielded sample cask which is lowered on the cart and pulled out from under ?he sample enclosure. A lead plug is then inserted " n .he opening of. the cask and the cask can be easily moved to the elevator in the control structures usinq the positioninq cart. By usinq this eleva+or no steps are encountered when moving the cask from the sample station to qround level. The shieldinq study results (see Subsection 18.1.20.3) indicate that this elevator should be accessible from a radiation level standpoint. In case of loss of off-site power, the elevator will be out of service since no emergency power is provided. However, +he undiluted sample is only essential for dete"mininq the chloride concentration which is not required un+il four days after sampling. Thi.. will allow a reasonable time for the restoration of off-site power. However, if af er two days off-site power is not restored, arrangements can be made to lower the sample cask from the tu bine operating floor to ground level through one of three open hatches. Since the
.undiluted sample is to be ent to an off-site laboratory, prior arrangements will be made to have a shipping container sent from the off-site laboratory or have one available on-site. The current intent is to have several shipping containers built which will hold the larqe volume casks, thus avoidinq the exposure which would result from tryinq to transfer the sample from the samplinq cask to another container.
- 18. 1-58
18.1.21.3. 4. 2 Chemical/Radiochemical Procedures 18, l. 21. 3. 4. 2. 1 Introduction The PASS provides a means of obtaininq primary coolant, suppress'on pool, and primary ard secondary contairment air samples for radiochemical and chemical analysis following a major reactor accident. Because of the extremely hiqh rad.oactivity levels associated with extensive fuel damaqc, the PASS and its auxiliary suppor. was developed with +he philosophy of providing the capability of obtaining the necessary samples and of performinq on a timely basis those analyses, as required, for immediate plant needs, or as defined by regulatory requirements.
Procedures and arrangemen+s will be established for shipping samples to facilities having the experience and equipment appropriate to performinq detailed and accurate chemical analyses on multi-Curie level samples.
The analytical procedures chosen vill satisfy the philo .ophy of performina only those analyses as reauired for operational support, of minimizinq personnel exposure and contamination hazards, and of dependinq upon outside analysis for extensive analysis and long-ranqe operational needs. Test- were performed by Gereral Electric to assess the effects of high fission produc+
levels on the suqqested analytical methods. The type of fuel damage associated with the release of meqacurie auantities of iodine and other activities also has the potential for releas'nq kiloaram quantities of stable or very lonq lived fission Droducts. Zt is conceivable tha+ +he primary coolant might con+ain 10-20 ppm of iodide and bromide. Also, the release of a major fraction of the core inventory of cesium and rub'dium may sliahtly raise the primary coolant pH. Such releases will also cause an increase in the coolant conductivi+y while radiolysis of the water will probably contribute to the formatior. of low levels of hydroqen peroxide. Depending upon the concentrations, these are all possible analytical interferences w'th the required analysis. Of these, the iodide/bromide inter fe ence with the chloride procedure is probably the most severe. However, sirce the reauirement .for chloride analysi- will be satisfied hy sendinq the samples o an off-site laboratory, the chloride procedu e heing proposed for the on-site laboratory is only to obtain a rouqh upper limit. .he effects of radiation interference have been qenerally evaluated and are summarized in Subsection 18. l. 21. 3. 6.
- 18. 1-59
SSES-FSAR 18, 1,21. 3. 4. 2. 2 Sample Prenaration All sample bottles, iodine cartridges, etc., will be numbered or otherwise identified prior to samplinq. This will eliminato.
unnecessa" y exposure as a result of handling high level samples fo the purpose of attachinq labels. A centralized loqqinq system will be developed to keep track of sample aliquo~
iden+ification, dilution factors, sample disposition, etc.
Liquid samples will be taken at the sample station in septum type bottles and +ransported to the analysis facility in lead con,.ainers. Sample aliquots are then taken from the septum bottles for analysis or further dilution. Aliquotinq and transfer will be performed usinq shielded containers, or behind a lead brick pile. Calibrated hypodermic syringes will be used for aliquotinq the hiqher activity samples. Tongs or other holdinq/clampinq devices will he available for holding the sample bottle during the transfer and dilutions to reduce hand and body exposure. tJnles prohibited by the intended analysis, dilutions
~
will be done usinq very dilute (about 0.01N) nitric acid as th diluent o minimize sample plate-out problems.
Beactor coolant activity levels on he orde of 1 to 3 Curies per qram would require a dilution factor of lx105, or larger, for qamma ray spectroscopy samples. As an example, a typical series of dilutions miqht be 0.1 ml (100 lambda) added to 10.0 ml at the sample station, followed by fu" her diluting of 0.1 ml to 100 ml in the laboratory. An aliquot of 0.1 ml wculd then be taken from the second dilution for countinq purposes.
Gas sample.>> are taken at the sample stat'on in the same 14.7 ml septum bottle used in the normal offqas sampler. A lead carrier is furnished with a small hole at the septum end so ha+ a qas sample can be withdrawn from the carrier usinq a hypodermic syrinqe without havinq to handle the hot+le.
Samples taken from the qas sample bottle wi1 ei+her be injected into a qas chromatoqraph for analysis or used to dilute the gaseous activity for gamma spectroscopy purposes. The dilutions will be performed in a manner analoguous to the liquid samples.
Fractional milliliter samples can be transferred to new 14.7 ml qas bottles without concern for sample leakaqe due to pressurization. For larqer volume aliqucts a gas syringe vill be used to draw a partial vacuum in the bottle pr'or to sample transfer.
Since there is no initial di1ution of the qaseous ac+ivity at t he sample station, extensive dilution may be required in the laboratory.
- 18. 1-60
SSHS-FSA!
- 18. 1. 21. 3. 4. 2. 3 Chemical Analv -is
- a. introduction The chosen procedures are not necessarily the mo"t sensitive no" the most accurate. They were chosen primarily on the basis of simplicity, stabil'ty and availability of reaqent, minimum radiation exposure, and least likely to cause ma d'or contamination problems. They hav>> been tested for radiation so.nsi:ivity and are suitable for use at the PASh desiqn hasi source term of 2.8 Ci/qm, and where applicable, with the de" iqn basis 0.1 ml to 10 ml dilution at the sample station.
- b. Boron Analysis Carminic Acid Method The chosen HACH method closely follows the ASTN D3082-74, "Standard Test'Method for Boron in Hater, Ne..hod A Ca"minie Acid Colorimetric Method." The HACH procedure is suqgested because the eagents and standards are available in small quantities, are conveniently packaged, and can be quickly prepared. Xt is estimated that the complete analysis, includinq reaqent Preparation, can be performed in 40 minutes.
This method was tested to be satisfactory for use at the maximum expected activity levels. The analysis is designed for boron concentrat'ons in the ranqe of O.l to 10 cpm cf boron. This sensitivity is particularly suited to the sample station's 0.1 ml to 10.0 ml dilution" since this corresponds to a range of 100 to 1000 ppm in the undilut d coolant.
- c. Chloride Analysis Turbidimetric Method (see also the discussion on conductivity)
The chosen method was developed by the General Electric Reactor Chemistry T aininq qroup. The p"ocedure is very similar to a HACH Chemical Co. procedure, "Turbidimetric Determination of Trace Chio"ide in Water>>.
The minimum quantity of measura ble chloride by this procedure is 0.5 q. I+ 5 ml of the O.l to 10 ml p imary coolan. dilution is used for analysis, the minimum measurable concentration would be 10 opm.
Usinq the 10 ml. direct primary coolant sample greatly'ncreases the sensitivity for measurinq chloride. A one ml of aliquot of this sample could be analyzed at the 0. 5 to 1.0 ppm level.
Tests of the radiation sensitivity of the method showed that activity levels comparable to the PASS desiqn basis source terms resulted in the equivalent of 1.8 ppm Cl- in he primary coolant
SSr,S-FSAP.
for the 0.1 to 10 ml dilution. This was deemed to be insiqnificant, importantly, as it is below the sensitivity limit, and more interference from the large amount of stable fission product halides potentially associated with the source terms will fa out-shadow the radiation ef feet itself.
Tests were also performed on the addition of 500 >g of boron added to 0. 5 to 20 pq of chloride. No 'ter ference wa observed wi.h the turbidimetric procedure.
Nea'surement of pH Indicator pape= for pH will be used for activity levels below 10K.
of he desiqn basis source terms.
The i radiation tests indicated that at 10% of the design basis source terms, the color stability was adeguate given only a drop of solution and less'han a 5-minu.e exposure.
Usinq this method, pH measurements can be taken a. the small volume sampler by placinq a piece of the paper into the sample bottle and using an air filled syrinqe to blow several approximately 0.1 ml aliquots from the sample valve into the bot.le to moisten the paper.
This type of samplinq approach can also be used to obtain a small sample for possible electrochemical pH measurement. Lazar Research Labs, Inc. manufactures a micro-pH electrode which functions on microliter samples. This electrode or similar micro-probe is curren+ly being evaluated for use at source term greater than 10'A of design basis.
Indicator pape" for pH can cover +he ranqe from 1-11 and distinguish differences of 0.25 pH uni s.
At very low conductivities, conductivity it self may be the hest ind'cator of the pH. For instance, at 0. 2 micromho/cm, he pH is bounded by 6.3 to 7.6, which i well wi.,h in the technical specifica ions for normal operaton. Thus the conductivity g
should serve as an adequate indicator of pH as long as conductivity is sufficiently low that it is impossible to be outside the technical specifications limit.
- e. Conductivity Measurements The Post Accident Sample Station is equipped w'th a 0.1 cm-~
conductivity cell. The conductivity meter has a linear scale with a six position range selector switch to give conductivity ranges of 03, 010, 030, 0100, 0300, and 01000 mic omho/cm when usinq the O.l cm-~ cell. This conduc+ivity measurement system vill be used to determine the primary coolant or suppression pool conductivity. During normal operation the BHP.
- 18. 1-62
SSES-F S AR technical specifications require maintaining the primary coolant helcw 1 micromho/cm, and conductivi:y measurements are the primary method of coolent rhemical control.
Conductivity measurements are, of course, non-specific, but they serve the impo"tant funct'on of indicatinq chanqes in chemical concentrations and conditions. Perhaps even more important, in the cas.. of the B!<8 primary coolant, the ccnductivity measurements can es"ahlish upper limits cf possible chemical roncentrations and can eliminate the need fcr additional analyses. For example, if. the conduct'vity is measured to be 5.0 micromho/cm, the upper limit on th~ chloride concentration is 1.4 ppm ~
The conductivity measurement can also be used tc bound the possible range of pH values. This relationship is shown in Fiqure 18. 1-11.
At a specific conductance of 1.0 micromho/cm t he pH mus. be between 5.6 and 8.7. Furthermore, a pH of 5 and a .pecific conductance of. 1.0 is an impossible situaticn since the conductivity is not larqe enough to support a hydrogen ion concentration of 10-5N. Fiqure 18.1-11 can, therefore, be used to great advantage in checkina on aqreement between pH and conductivity measurements and possibly eliminating the need for pH measurement if the conduc.ivity is very low. In qene"al, accurate pH measurements are difficult to make in very low conductiv ty water as the 'pedanre of the solution may be siqnificant compared to the impedance of the mea urinq device, and conduct.ivity measurements are usually considered a better indicator of the maximum H+ or OH- conrentration.
- 18. 1.21.3. 4 2. 4 Radiochemical Analysis--Gamma Bag Spectroscopy Af er the samples have been hrouqht to the chemistry laboratory and appropriately diluted, they can be carried without shielding to the countinq room which is adjacent to the chemistry laboratory. The appropriate dilution factors will be somewhat dependent upon the detector and shelf arrangements available. A prior determination of the maximum desirable dose rates for the various shelf confiquration will be made to minimize this problem. The present high resolution, high efficiency Ge(Li) detec ors, coupled with the multichannel analyzers, and computer data reduction in the on-site countinq rocm vill easily handle the analysis of these samples.
The qas samples will be counted in the standard of f-qas sample vials and the liquid samples will be counted in the standard sample bottles used durinq normal operation since calibration curve., for these geometries will be available and reqularly
SS.".S-FSAP.
updated. Calibration curves w'l also be available for +he particulate fil'er and iodine cartridge qeometr'es. In qeneral,
".he'ounting of the post-accident samples will follow th~ normal countinq oom procedu es. A special pos"-accident lib=ary vill have to be developed for use by +he compu+er peak search and identifica+ion routine .o supplemen. the normal iso.ope libra"y.
.he pos -acciden+ peak search and identif ication library will con.ain the principal gamma rays of the followinq isotopes in addition to the standard activated corzosicn products:
Noble qases: Kr-85, Kr-85m, Kz-87, K"-88, Xe 131m'e 133~ Xt 133m'e 135 Iodines: I-3 31, I-132, I-133, I-135 Cesiums: Cs-134, Cs-137 Others: Ba/La-140, Ce-141, Ce-144, Hu-106, Te-129, Te-129m, Te-131, Te-131m, Np-239 If the levels of noble qases in the ambient atmosphere surroundinq the detector is hiqh enough to cause signif icant interference or overload the detector, a compressed ai" oz nitzoqen purqe of the detector shield volume will be maintained.
18,1. 21,3,4,2. 5 Gas Analysis-Gas Chromatogra2hy qas chromatoqraph will .be used to measure hydrogen, nitrogen and oxyqen concentrations in containment atmosphere and dissolved aas samples. The qas chromatoqraph will be located in the chemistry laboratory and ven+ed to a laboratory hooR. Samples for qas analysis will be used undilu+ed frcm the sample vials and injected into the qas chromatoqraph. Since the sample sizes needed for the analysis will ranqe from 0.1 to 1 milliliter, may be necessary to place a temporary lead shield around the inst ument. The analysis of the drywell, wetwell, and secondary containment samples will be done usinq standard procedu" es.
Calibration curves for. the instrument will he pzepared and periodically updated.
In the mixture of hydroqen, oxyqen, nitroqen, and possibly krypton, +he analysis sensitivity should be sufficient to detect any of these constituents at the 0.1'5 by volume level, or lower, providinq the Kr:N ratio in this mixture does not vary by more than a factor of 10 in eithez direction. At the 0.5't level th analysis should be accu ate to within 20% cf the measureR concentration. At concentrations above 1%, ",.he analysis should be accurate to within 5% of the measured concentration.
The dissolved qas sample will contain krypton or other tracer in addition to oxyqen, nitroqen, and possibly hydroqen. Although the analysis of the dissolved qas sample for hyd ogen should be
- 18. 1-64
SSES-FSAB K
reliable, the analysis for oxyqen and nitroqen presents several difficulties. The major problem is due <o the incomplete evacuation of the sample vial which init ial'y ccntains air. A partial vacuum (4-5 Dsia) is drawn on +he v-al befo e the sample
's taken, however, this leaves a significant amount of air in the v'al. This may not be a significant problem if the amount of dissolved oxyqen or nitrogen tripped from the ccolan+ is large compared to that left in the evacuated vial, since a correction can be made based on the pressure measurements taken before and after taking the sample. Powever, dissolved oxyqen and nitrogen is not required by HUBEG-0737, which states tha t determination of dissolved hydroqen qas in the coolant is adequate. Tn case the need should arise, a procedure will he established tc tap of C the sample 1'e in the sample station and run this to an in-line oxygen monitor. The flow would then return to the liquid return line to "he wet well.
- 18. l. 21. 3. 4. 3 Storage an d Disposal o C Sa mo 1.
Short term sample storage a eas will be provided in the chemistry laboratory and countinq rooms in botn the cn-site and EOF facilities. An area for lonq term s.orage of the samples will be designated at a later date. Low level wastes generated by the chemistry procedures will be flushed to radwaste in the on-site chemistry laboratory and collected in removable carboys in the EOP. The carboys will. then he taken +o an or.-site location for d'sposal to the radwaste system. Ul..imate procedures for disposal of the samples will be determined later; however, after a sufficiently long decay period,,he activity levels vill be siqnificantly reduced. This will ease exposure problems durinq disposal.
18.1.21.3.4.4 System Tes+ing and Operator Vrainina To ensure the lonq-term operability of +he PASS, it tested semiannually. Samples will be taken from all gas sample will be points; however, the number and type of liquid samples taken will be based on the operatinq status of the reactor at the time. The semiannual functional testinq will also serve to maintain operator proficiency. In addition to the scheduled tests, the system will be used for operator train'g on an as-needed basis.
To ensure an adeguate pool of qualified PASS operators, a formal traininq proqram will be established. Thi proqram will be part of the chemistry technician qualif ica+ion program. All plant chemistry echnicians and chemistry management personnel will be required to show competence in +he operaticn of the sample station and the chemical analysis procedures.
- 18. 1-65
SSES-FSA!
- 18. 1.21.3. 4.5 Core Damage E.timat ion Procedure A revised core damaqe estimation procedure to bc used on both units will be developed prio" to .he start up fcllowinq the first refuelinq outage of Uni. 1.
- 18. l. 21. 3..5 Dose Rate Analvsis Radioactivity source terms were calculated fo use in desiqn of the PASS shieldinq. These source terms are fc" a LOCA assuminq a release of fission product activity as defined by HUREG 0578.
Source terms were calculated for a three year reactor operation a+ 3293 AHt. Fo the purposes of specifying shieldinq design source terms, a decay period of one hour has been assumed between reactor shutdown and initial samplinq. Although there is no decay pe iod specified in )1UREG 0578 the source .erms calculated for PASS still esult in a conservative desiqn. The PASS is desiqned to limit operator whole body exposure to 100 mRem as a result of taking and analyzing the sample. NUREG 0737, on the other hand, limits the opera. or exposure to less than 5 Rem whole body exposure for the entire operation.
Usinq a one hour decay and the fractional releases o core inventory specified by VUBEG 0578, the primary coolant and primary containment atmosphere fission prcduct concen rations are calculated to be 2.6 Ci/qm and 0.046 Ci/cc, respectively. Usinq these fission product concentrations, qamma radiation source
+erms were de+ermined in terms of tleV/sec fo" ten gamma energy qroups. These radiation source +e ms were used for shieldinq desiqn and sample dose rate calculat'ons. Assuming point sources, the calculated dose rates per unit volume of coolant and containment atmosphere are 125 R/h/gm and 1.9 R/h/cc at 4 inches, respectively Thus, the 0.1 milliliter reactor sample would have a maximum exposure rate of about 12 R/h at 4 inches and 14.7 millil'ter vial of containment atmosphere at STP would have an exposure ra..e of 25 R/h at, 4 inches. Usinq the calculated source terms, dose
=ate estimates resultinq from ac.ivity in +he sample station and sample casks were calculated for var'ous distances. The results are qiven ' Table 18.1-7. These dose rates wil1 he used in a ime-motion study to estimate the total in+eqrated dose expected during samplinq and analysis after the sample station is opera tiona l.
- 18. 1-66
FS AR 18,1,21,3,6 Trradiation Fffect On Analytical Procedures Some scopinq tests were performed to s.udy the effect of high fission product levels on the proposed analytical procedures.
The core invento"y of. in<lividual nuclide beta energies in term.-
of HeV/second/klr t after one hour decay was taken from the same CTHDER run as used to calculate, he PASS activi .y sou ce te ms.
The YUBFG-0578 release fractior.s were used tc determine the fraction of the core inven ory dissolved in +he primary coolant.
The "all other" catoqory was iqnored as at a 1% release f=ac+ion the dose contribution from these nuclides is neqligible compared to the 504 halogen anrl 100% noble qas releases. The results a=e shown in Table 18. 1-13. For the sake of simplicity, i was assumed that the qamma enerqy deposition in the sample was negligible compared +o the beta energy depositior.. It was also assumed that 1007> of the. beta energy was absorbed in the sample.
The net result, 1.92xl06 Rads/hr, is conservative as the gamma enerqy absorpt'on for small samples would be much less than the beta enerqy escapinq the solution.
Dose rates approachinq 2X10~ B/h are availahle in the VNC Co-60 ir"adiation facil.ty. A+ 93 erqs/g/R/h, this corresponds to 1.8xl06 Bads/hr, and approximates the calculated maximum energy deposition possible for the reactor coolant. Tests were run to de+ermine the effec+s of radiation on the conduc+ivity, pH, chloride, and bo "on analytical procedures., he true enerqy deposition within the irradiated sample holders was dete mined by Fricke dosimetry usir.q the sample holders as dosimeters. Except for conductivity and pH measuremen,.s, +he dose rates were considerably larger than would be encountered with the PASS sou ce terms. These hiqhez dose rates were used to achieve a better measurement of. the radiation effect, and it was then assumed +hat this effect would be linear with dose rate. It is hoped to verify this assumption in later studies.
18,1,21,:3. 6.1 Conductivity Cell A 0.1 cm Balsbaugh conductivity cell and stainless steel holder was irradiated at va ious position 'n .he 4 1/4-in. dia. Co-60 irradiation tube. The flow path from this. conductivitv cell was connected to a 0.1 cm Beckman conductivit y cell downst earn of the cell under irradiation. Both static and Slowing irradiation tests were performed. The. flow tests were performed at ca. 125 cc/min with a 3 to 4 min flow delay between the Balsbaugh and Beckman cells. The Beckman cell, therefore, served to dete mine if there we e any relatively long lived adia+ion products remaininq in solution. An in-line thermometer was mounted in the flow system downstream of the Beckman cell.
- 18. 1-67
SSES-FSAR
- 18. 1.21.3. 6. 2 Purification A Gelman Hater-I purification unit was installed in the conductivity cell flow loop. The output conductivity of the water from the purification unit was 0.055 pS/cm, as indicated by the purification units built in the conductivity meter. The vater flow was from the purification unit through the tvo conductivity cells under study and back to the reservoir of the purification unit. The output of the conductivity meter associated with the irradiated cell vas continuously recorded.
The hiqhest radiation field in the 4 1/4 in. irradiation tube, as measured by a Victoreen R Heter, vas 7.4x105 R/h. The actual cell enerqy absorbtion rate at this position vas determined by removinq the conductivity element and using the cell holder as a Fricke dosimeter container. The result, 9.8x105 Rads/hr was also used to convert the R/h measurements at the other elevations to Rads/hr by assuming a constant ratio betveen the field intensity and the energy absorbtion. (This is not strictly true as the photon energy- distribution varies vith the elevation in the irradiation facility. Consequently, the fraction of the photons penetrating the stainless steel cell holler will vary slightly.)
The results of these experiments are summarized in Table 18.1-14.
There vas apparently some pickup of impurities from the flow loop materials as 0.-10 pS/cm vas the lowest loop conductivity observed. The. 0.06pS/cm at the output of the purification unit vas confirmed by connectinq one'of the flow cells immediately as the. output.
In the case of the flowing measureme'nts, there vas a steady increase in conductivity from 0.11 to 0.65 pS/cm as the irradiation intensity increasel from 1.3xl0>> to 6.6x10 Rads/hr.
The conductive species which were formed were relatively stable as there was little difference between the conductivity as measured at the irradiated cell and the downstream cell. In fact, when the flow was stopped and the conductivity of the irradiated cell vas allowed to come to equilibrium, the cell could be removed from the radiation field and the conductive would remain constant, at least up to several hours, the longest period observed. The flov vas secured at each irradiation intensity and the conductivity vas monitored until a steady-state condition was attained. From the data in Table 2 that a maximum conductivity is attained at about it 2.2 would appear pS/cm, anl that the conductivity diminishes with increasing radiation intensity. The steady-state difference in cell behavior at 6.6xl0~ and 9.8x105 Rads/hr is unexplained.
It is suspected that the conductivity is due to the formation of hydrogen peroxide, but this has not been confirmed. It is obvious that there vill be some radiation effect on the conductivity at very high fission product concentrations. This 18 1-68
e
SSES-P SAR does not appear too serious, hovever, as 2.2 pS/cm corresponds to a NaCl concentration of 1.0 ppm. The concentration of stable fission products, particularly I-127 and I-129, associated with the high Curie concentrations vill at the same time result in considerably higher conductivities.
- 18. l. 21. 3. 6. 3 Conductivity of 10 pram Chloride QCL-) Solution Irradiation tests vere performed to determine the radiation effect on the conductivity of a dilute NaC1 solution. It vas anticipated that if the pure water conductivity increases under irradiation vere due to the formation of H202, this might be suppressed by the presence of the Cl- ions. In this experiment the NaCl solution vas pumped from a reservior through the two conductivity. cells and. back to the reservior. A common conductivity bridge vas used to alternately determine the conductance of each cell, and thereby eliminate any bias between different bridges.. The testing was done at the highest available irradiation level, 9.8xl05 Rads/hr. The solution temperature, as indicated by a flov thermometer dovnstream of the unirradiated cell, ranged from 59.5 to 60.2~P. Several alternate conductivity readings vere taken on each cell approximately five minutes after each change in condition, and when the cell conductances had reached a steady value The average result for each co'ndition is given in Table 18.1-15. 'The difference between the cell readings for any given set conditions is attributed, to errors in the stated cell constants. The conductivity of the flowing stream increased by approximately 0.6 pS/cm for both cells before and after irradiation,. which may be the result of the generation of some long lived species. This possibly is supported by the Beckman cell, which although located outside the radiation field, showed a 0.6 p S/cm increase in conductivity during irradiation.
The puzzling observation was the large drop in conductivity of the static solution during irradiation. This should he investigated further.
- 18. 1. 21. 3. 6. 4 pH.-
Solutions of pH 3.8 and, 10.0 were made up using HCl and NaOH, respectively. LO-Ion pH test'aper vas placed in aliquots of these solutions and the solution was inserted into the 9.8xl05 Rads/hr position (as determined by Pricke dosimeter). A 10.0 minute exposure for a total dose of 1.6xlO< Rads completely destroyed the color in the acid solution and reduced the color intensity of the basic solution to a pale green. This test vas then repeated using a 1.0 min exposure at the same intensity level for an exposure of 1.6xLO~ Rads. This exposure shifted both solutions about 1/2 pH unit to the more acid side. The t8 1-69
SSES-FSAR results would not necessarily indicate that pH indicator paper cannot be used at the highest dose rates, but more importantly, that the paper cannot be immersed in a relatively large volume of solution. If the paper were merely moistened by a drop or so of solution, most of the beta particles vould escape the paper with little energy deposition and the paper would not be surrounded by a highly radioactive solution with the resultant beta field and water excitation products. This subject is still under consideration.
At source terms on the order of 10% or less of the maximum~, the irradiation effect, for even an immersed strip, would be tolerable at exposures less than 5 min, as it would result in less than an-0.5 pH unit shift.
Some measurements vere also made to determine the effect. of irradiation on pH electrodes. Long leads are needed on the pH
'lectrodes in order to reach in the Co-60 irradiation facility, and these electrodes were not available. He intend. to order some nev- electrodes and vill continue this study. In the meantime, we have irradiated a glass membrane pH electrode to 1.6xl0~ Rads at
~
a 9.85x10~ Rad/hr intensity and found it, still functions
~
following irradiation.
~ ~
18 l,gl,3.6.5 Tuzbidmetgic Chloride Procedure Using the maximum source term of 2x10~ Rads/hr, ml diluted primary coolant sample vould have an internal beta exposure of 2xlov 'Rad/hr. The turbidimetric method calls for a total volume of 25 ml. Therefore, even if the entire 10 ml of diluted sample were used, the dose rate of the final analysis solution would be less than 8xlO~ Rad/hr. Test solutions containing 0, 1, 5, and 20 /gm of chloride in 25 ml vere processed through the chloride test methods in pairs. During the 15 min turbidity-formation period, one sample of each set vas irradiated at an absorbed dose rate of 4.4xlov Bad/hr as determiued by yricke dosimetry. rhe The originally calculated source term was 1.9x10~ Rads/hr.
Thirty'-five percent of this source intensity,, hovever, is ..
due to noble gases vhich would escape solution in the sampling process, A 10% source term for pH measurement would then be approximately 1 2x10~ Rads/hr and a 5-min
! exposure would correspond to a lxlO~ Rad energy absorption, vhich is approximately the exposure causing a 0.,5 pH shift.
18 1-70
SSZS-PSAR maximum observed radiation effect was a difference of about 10 turbidity units between the irradiated and unirradiated 1 gm Cl solutions. This difference is equivalent to about 10 p /gm of chloride in the 25 ml of solution being processed. Assuming this increase in turbidity is proportional to the dose, the maximum effect would be (10 p qm) {8x10~/4. 4xlO~) = 0. 18 pgm. If only 0. 1 ml of reactor vater were used for the oriqinal'ample, this would be equivalent to 1.8 ppm of Cl- in the primary coolant. This error is probably insignificant as the interference from all the stable iodine associated vith the high radiation intensity is likely to be far larger.
The test data also indicates that as little as 5 p gm of Cl- in the 25 ml of test solution inhibits the formation of the radiation-induced turbidity. It is suspected that the increased turbidity is due to the precipitation of silver peroxide and the 5 p qm'CI- inhibited the formation of hydrogen peroxide. In any event, it was concluded that the test method is useful for highyly radioactive solutions above the 10 ppm level, or for less radi;oactive solutions above the 1 ppm level. For low activity samples which- do not need to be diluted and vhere at least a 1 ml of sample is available, the method is useful above the 100 ppb level.
18.1 21.3.6 6--Carminic-Acid Boron Analysis Using the maximum source term-of 2x10~- Rad/hr, an 0.1 ml to 10 ml diluted primary coolant sample- would have an internal beta exposure of ca. 2xl0~ Rad/hr The colorimetric method calls for a total'olume of 25 ml. Therefore, even if the entire 10 ml of diluted solution vere used, the dose rate of the final analysis solution would be less than 8xlO~ Rad/hr. Test solutions containinq 0 and 20 pgm..of boron vere processed through the boron test methods in pairs. During the 40-min color development phase, one sample of each pair was irradiated at an absorbed gamma-radiation dose level of 4.4xlO~ Rad/hr as determined by Pricke dosimetry. The maximum irradiation effect observed was a difference of 0.854 absorbance units between the irradiated an unirradiated blank solutions. This difference is eguivalent to about 27 pqm of boron in 25 ml.of solution being processed.
Assuming this difference i:s absorbance is proportional to the dose, the maximum affect= vould be (27 > gm) (8xlO~/4.4xlOs) = 0.49 gm. If only 0.1 ml of reactor water vere used for the original sample, this is equivalent to a 5 ppm error in the primary coolant analysis This error is totally negligible in terms of the levels of boron required for reactor shutdown..
'8
~ 1-71
S SES-PS AR
$ 8 $ ,22 TRAINING FOR MITIGATING CORE DAMAGE /II. B 4g 18.1.22.1- Statement of Requirement Licensees are required to develop and implement a training program to teach the use of installed equipment and systems to control or mitigate accidents in which the core is severely damaged.
Shift technical advisors and operatinq personnel from the plant manaqer throuqh the operations chain to the licensed operators shall receive all the training indicated in Table 18.1-8..
Managers and technicians in the instrumentation and control, health physics, and chemistry departments shall receive training commensurate with their responsibilities.
Applicants for operating licenses should develop a training program prior to fuel loading and complete personnel training prior to full-power operation.
1~81~22~2 - Interpretation None reguired.
18,1 22~3 Statement of response A course titled "Mitigating Core Damage" has been, developed and is available to all shift technical advisors and operations personnel from the plant manager through the operations chain to and including licensed operators to requirement. A course outline is fulfill provided this training in Table 18.1-9.
Managers and technicians in instrumentation and controls, health physics, and chemistry are given training commensurate with their responsibilities during accidents which involve severe core damage.
18 1-72
SSES-FSAR 18 1 23- - ~ELIEF AHD SAFETY VALVE TEST REQUIREMENTS
~ /II D 1$
18,1.23.1 Statement-of requirement Boilinq-vater reactor licensees and applicants shall conduct testinq to qualify the reactor coolant system relief and safety valves under expected operating conditions for design-basis transients and accidents.
Licensees and applicants shall determine the expected valve operating conditions through the use of analyses of accidents and anticipated operational occurrences referenced in Regulatory Guide. 1.70, Revision 2.. The single failures applied to these analyses shall he chosen so that the dynamic forces on the safety and relief valves are maximized. Test pressures shall be the highest predicted by conventional safety analysis procedures.
Reactor coolant system relief and safety valve qualificaiton shall include qualification of associated control circuitry, piping, and supports, as well as the valves themselves.
Preimplementation reviev .will be based on EPRI, BRR, and applicant submittals vith regard to the various test programs.
These submittals should be made on a timely basis as noted below, to allow for adequate reviev and to ensure that the following valve qualification date can he met:
Final BRR Test Program October 1, 1980 Postimplementation review will be based on the applicants~ plant-specific suhmittals .for gualification of safety relief valves.
To properly evaluate these plant-specific applications, the test data and results of the various programs vill also be required by the folloving dates:
BIR Generic Test Program Results July 1, 1981 Plant-specific submittals confirming adequacy of safety and.
relief valves based on licensee/applicant preliminary review of generic test program results July', 1981 Plant-specific reports for safety and relief valve qualification October 1, 1981 Plant-specific submittals for piping and support evaluations January 1, 1982 18.1.2~3 2 Intermretation-Hone required.
- 18. 1-73
SSES-FSAR 18.1 23.3 Statement-of Response PPSL is participating in the BHR Ovner's Group (BHROG) program to test safety/relief valves (SRVs). Hyle Laboratories in Huntsville, Alabama has been contracted to design and build a test facility., The design is complete and construction is well underway. The facility will be capable of hiqh and lov pressure valve tests.
Documentation of the BWROG testing program vas sent to the NRC on September 17,, 1980 by a letter from D.B. Haters to R.N. Vollmer.
A summary of this document is provided belov.
An enqineering evaluation was done to identify the expected operatinq conditions for SRVs during design basis transients and accidents.. This evaluation 'indicates the SRVs may he required to pass lov pressure liquid as a result of the Alternate Shutdown Mode (described in Subsection 15.2.9). No other significantly probable event, even combined vith a single active failure or single operator error, produces expected operating conditions that justify qualification of SRVs for extreme operating conditions.. Therefore a test program vas developed to demonstrate the SRVs'apabilities as may be necessary during the Alternate Shutdown Mode The test results vere submitted by a letter to A. Schwencer from N. H. Curtis on Jul.y 1, 1981 (PLA-865). A plant specific SRV qualification report was submitted to- the NRC on October t, 1981 (PLA-940) , This report includes all necessary evaluations of piping and supports.
I'8 1 g4- - - SAFETYQRELXEF VALVE POSITION INDICATION /II D 3)
- 18. 1~/~41- Statement of R~euirement Reactor coolant system relief 'and safety valves shall be provided with a positive indication in the control room derived from a reliable valve-position detection device or a reliable indication of flov in the discharge pipe.
The basic requirement is to provide the operator with unambiguous indication of valve position (open or closed) so, that appropriate operator actions can be taken The valve position, should be indicated in the control room. An alarm should be provided in conjunction with this indication.
18 ..1-74
SS ES-FS AR The valve position indication may be safety grade. If the position indication is not safety qrade, a reliable single-channel direct'ndication powered from a vital instrument bus may be provided if backup methods of determining valve position are available and are discussed in the emergency procedures as an aid to operator diagnosis of an action.
The. valve position indication should be seismically qualified consistent with the component or system to which it is attached.
The position indication should be qualified for its appropriate environment (any transient or accident which w'ould cause the relief or safety valve to lift) and in accordance with the Commission order on May 23rd, 1980 (CLI-20-81).
, It is important that the displays and controls added to the control room as a, result of thi*s requirement do not increase the potential for operator error., A human-factor a'nalysis should be performed taking into consideration:
(a) the use of this information by an op'erator during both normal and abnormal plant conditions, /
{b) integration into emergency procedures.,
I (c) : integration into operator training, and
{d) other alarms during emergency and need for prioritization of alarms .
Documentation should be provided that discusses each item of the clarification, as well as electrical schematics and proposed test, procedures in accordance with the proposed review schedule, but in no case less than four months prior to the scheduled issuance of the staff safety evaluation report. Implementation must be completed prior to fuel load.
18.1.24 2 Interpretation None required..
18,1,2~43 Statement =of He~sonse-Each of the safety/relief valves (SRVs) (16 per unit) is provided with a safety" grade acoustic monitoring system to detect flow through the valve. An acoustic sensor is mounted on the discharge piping, downstream of each valve.
18 1-75
SSES-FSAR The monitors are grouped into two divisions with 8 valves each.,
Each division has group annunciation for valve opening and for
~ ~
division loss of power. A red annunciator window is provid,ed for
~ ~
valve opening and white annunciator window for loss of power on a front row control panel for these annunciations. Each division is
~
powered from a 1E vital instrument bus.
Individual indication of an open valve is provided by a red light (1 light for each valve) on a front row control room panel (10601). Individual indication of valve position is also available on a back row control room panel ~here the siqnal conditioninq instruments are located.
The acoustic monitoring system is designed to be safety grade.
This equipment has been qualified to IEEE-344-1975,. IEEE-323-1974 and NUREG-0588 in accordance with the .Commission, order: on May 23,.
1980 {CLI-20-81)
Additional desiqn information is presented in Subsection 76..1b 1 7 A human factors review of the front row.control panel on which these indicators are located has been completed. This same i analysis has been applied to the SRV position indicators added to this panel.
The. use of tailpipe temperature detectors in. the emergency
~
procedures. is discussed in a let'ter from N. H. Curtis to B. J.
~ ~ ~
Younqblood on April 30,. 1981 {PLA-736) ..
~
18.1 25 AUXILIARY FEEDQATER SYSTEM EVALUATION /II E 1.~1 This requirement is not applicable to Susquehanna SES.
18 1.26 AUXILIARY FEEDWATER SYSTEM INITIATION AND FLOP
.= '. - -=/II E.1 2}-
This regui;rement is not applicable to Susquehanna SES 18 1,27- - - ZHEQGENCY POSER FOR PRESSURIZER
~ HEATERS~II E-3 lg This requirement is not applicable to Susquehanna SES.,
18 I'-76
SSES-ES AR 18 1 28 DEDICATED HYDROGEN PENETRATIONS /II E. 4 1}
18,1.28.1 - Statement of Reguirment Plants using external recombiners or purge systems for postaccident combustible qas control of the containment atmosphere should provide containment penetration systems for external recombiner or purge systems that are dedicated to that service only. These systems must meet the redundance and single-failure requirements of General Design Criteria 54 and 56 of Appendix A to 10 CPR 50, and that are sized to satisfy the flow requirements of the recombiner or purge system.
The procedures for the use of combustible qas control systems followinq an accident- that results in a degraded'ore and release of radioactivity to the containment must be reviewed an revised, if necessary.
Operatinq license applicants must have design changes completed by July 1, 1981 or prior to issuance of an operating license, whichever is later.
- 18. l. 28. 2 -'Interpretation Hone required.,
I
- 18. 1~2. 3 Statement of response Susquehanna SES desiqn includes 100% redundant internal hydrogen recombiner systems for postaccident combustible gas (hydrogen) control. Therefore this requirement is not applicable to Susguehanna SES..
~8$ 29-" -CONTAINMENT-.ISOLATION DEPENDABILITY /II E~ 4 2) 18.1-,29 1- Statement of Reauirement t
(1) Containment isolation system designs shall comply with the recommendations of Standard Review Plan (SRP) Section 6.2.4 (i.',e, that there be diversity in the parameters sensed for the initiation of containment isolation).
(2) All plant personnel shall given careful consideration to the definition of essential and nonessential systems, identify 18 1-77
SSE S-F SAR each system determined to be essential, identify each system determined to be nonessential, describe the basis for selection of each essential system, modify their containment isolation desiqns accordingly, and. report the results of the reevaluation to the NRC.
(3) All nonessential systems shall be automatically isolated by the containment isolation signal.
(4) The design of control systems for automatic containment isolation valves shall be such that resetting the isolation signal will not result in the automatic reopening of containment isolation valves. Reopening of containment isolation valves shall require deliberate operator action.
{5) The containment setpoint. pressure that initiates containment isolation for nonessential penetrations must be reduced to the minimum compatible with normal operating conditions.
Containment purqe valves that do not satisfy the operability criteria set forth in Branch Technical Position CSB 6-4 or the Staff Interim Position of October 23, 1979 must be sealed closed as defined in SRP 6.2.4, item IX.3.f during operational conditions 1, 2, 3, and 4. Furthermore, these valves must be verified to be closed at least every 31 days.
Containment purge and vent isolation valves must close on a high radiation signal.
Applicants for an operating license must be in compliance with positions 1 through 4 before receiving an operating license.
Applicants must be in compliance with positions 5 and 7 by July 1, 1981, and position 6 by January 1, 1981 or before they receive their operatinq license, whichever is later for each position.
18.1.29.2 Interpretations
~
From item 4, the opening of containment isolation valves must require a deliberate operator action.
Prom item 5, the containment isolation setpoint pressure should be optimized to prevent -unnecessary isolations during normal operations. However, containment isolation must not be prevented or delayed during an accident.,
18'-78
SSES-FSAR 18.1.29.3 Statement of-Response n> Containment isolation signals are actuated by several sensed parameters (refer to Table 3.3.2-1 in the Technical Specifications) . This complies with SRP Subsection 6.2.4, Para qr aph XX-6.
(2) Each process line penetrating containment was reviewed to determine whether it is an essential or non-essential line for purposes of isolation requirements. The classification for each line is given in Table 18.1-10.
Justification for the classification as an essential or non-essential line was also developed and is provided in Table 18..l-ll., Systems identified as essential are those which may be required to perform an indispensable safety function in the event of an accident. Non-essential systems are those not required, during or after an accident. Since instrument lines are not governed by isolation signals but are equipped with a manual isolation valve followed by an excess flow check valve outside the containment, the review of these lines was limited to ensure compatibility with the penetration listing in Table 6.2-12a..
(3) ~ All lines to non-essential systems are provided with isolation capability, All isolation valves in these lines, except the reactor water clean-up system (RMCU) discharge (G33-1FC042 and 1F104 receive auto-isolation signals 'alves (refer to Table 18..1-10)., The isolation function for the R%CU discharge lines is provided by three series check valves (141-1F010A,B, HV-14107A,B and 633-1F039A,B) which prevents back flow from the reactor vessel., The RVCU discharqe isolation valves are not closed to prevent the loss of the filter cake in the RSCU filter demineralizer system and injection of resin into the vessel on restart of the RWCU system.
(4) All containment isolation valves, except those listed below, will not automatically open on logic reset.
a) The RCXC and HPCX turbine steam supply line isolation valves (HV-1F007, HV-1F008 ~ HV-1F002 and HV-1F003) are normally open valves and will close upon a steam line break isolation signal These valves are essential valves and do not receive a containment isolation signal.. Reopening of these valves wi11 occur if the hand switches are not placed in the closed position by the operator prior to actuation of the reset switch .and the isolation parameters have cleared.
These valves are equipped with key-locked maintained
'contact switches to insure that these valves are open
'h
SSES-FSAR
'during ECCS initiation. If a pipe break condition vere detected, then these valves vill be automatically closed. After the pipe break problems are cleared these valves can be reopened to their normal emergency positions by deliberate operator action using the key-locked reset svitches for each system. The operator is required to ensure that the valve svitches are in the correct position prior to operating the keylock reset svitch.
The inboard HPCI and RCIC isolation valves each have a pressure equalization valve (HV-1F100 and HV-1F088) around them.. The equalization valves are normally closed and are only used to equalize the pressure around the inboard isolation valve in order to open them. If open,'the valves vill close upon a steam line break isolation signal. Reopening of these valves will occur if the hand svitches are not placed in the closed position by the operator prior to actuation of the reset svitch and the isolation parameters have cleared.
As vith the HPCI/RCIC isolation valves the egualization valves vill reopen upon deliberate manual logic reset using the key-,locked reset switches. These valves must open in order .to allow the inboard isolation valves to reopen to their. normal emergency positions when the pipe break problems have cleared. If the equalization valve svitches are not in the open position the operator must manually open them to equalize the pressure around the inboard HPCI/RCIC valves.
The RHR containment isolation valves (HV-lF016A,B, and HV-1F028A ~ B) .associated with the drywell and suppression pool spray lines vill reopen if their handswitches are placed in the open position prior to actuation of the reset svitch, the LPCI injection signals are clear, and the LPCI injection valves are closed. These spray line valves are normally closed and are provided key-locked hand switches and receive an .isolation signal as described in Tables 18.1-10 and 18.1-12., If the valves were open before an LPCI injection event, these valves vill automatically close and can not be reopened still if the 'LPCI injection signals exist or the LPCI injection valves are still open.. This is to insure that the 'LPCI injection function-will not be inadvertently jeopardized. by opening of the spray line isolation valves. If these spray line valves were closed before the LPCI injection event, the valves vill remain closed after reset even after all injection signals are clear and the LPCI valve are closed. 'njection 18 ~ .1-80
SSES-FSAR As noted in Table 18.1-10 only the outermost valve is considered a containment isolation valve for these penetrations. The three inboard valves HV-lF021A, HV-1F-27A and HV-1F024A are spring return to <<AUTO<<
switches and will not automatically reopen after logic reset and all signals clear. These inboard valves have not been considered containment isolation valves because they can not be leak tested in the <<forward" direction. Since these valves effectively function as containment isolation valves, a logic reset will not automatically result in a breach of containment integrity for these penetrations.
- 5) The BWR Owners'roup has performed a generic analysis which is summarized as follows. The containment isolation analytical setpoint pressure for Nark I, II,.
and III containments is approximately 2 psig (drywell pressure). Under normal operating conditions, fluctuations in the atmospheric barometric pressure as well as heat inputs (from such sources as pumps) can result in containment, pressure increases on the order of 1 psi:. Consequently,. the isolation setpoint of 2 psiq provides a 1 psi margin above the maximum expected operating pressure. The 1 psi margin to isolation has proved to be a suitable value to minimize the possibility of spurious containment isolation. At the same time, it. is such a low value (particularly in view of the small drywell volume of Nark I, II, and III containments) that i4 provides a very sensitive and positive means of detecting and protecting against breaks and leaks in the reactor coolant system. No change of the setpoint is necessary for these containment types.
concurs with.this position. Therefore, no modifications to the containment isolation pressure setpoint are necessary in response to this requirement.
6)
'PGL The design of the containment atmosphere purge valves was reviewed against Branch Technical Position CSB6-4.
This review identified several valves that do not meet these criteria. These valves will be qualified to meet this criteria as stated in a letter to B. J.. Youngblood from N. R. Curtis on April 1, 1981 (PLA-700). Valves in Unit 1 will be fully qualified prior to the startup following first refueling. Valves in Unit 2 will be qualified prior to Unit 2 fuel load.
{7) Two redundant safety grade radiation monitors are installed down stream of the Standby Gas Treatment System. A high radiation level trips the Standby Gas Treatment System. This signal is used to close the
SSES-PS AR following containment isolation valves in the vent and purge system: HV 15703 HV 15704 HV 15705~ HV 15711~
HV-15713, HV-15714, HU-15721 HV-15722'V-15723, HV-15724, HV-15725, SV-15736A, SV-15737, SV-15767 and SV-15776A.
The radiation setpoint is set to so that the 10CPR 100 limits are not exceeded. The high radiation alarm for these detectors is annunciated on control room front row panel 1C653. The rad'iation level measured by these detectors is recorded on control zoom backzow panel lc600..
18 1 30 - ACCIDENT-MONITORING INSTRUNENTATION /II P 1$
18 .1.30,1 Statement of Requirement The following equipment shall be added:.
(1) Noble gas effluent radiological monitor; (2) Provisions for continuous sampling of plant effluents for postaccident releases of radioactive iodines and particulates and onsite laboratory capabilities; (3) Containment high-range radiation monitor; (4) Containment pressure monitor; (5) Containment water level monitor; and (6) Containment hydrogen concentration monitor.
It is important control room as that the displays and controls added to the a result of this reguirement not increase the potential for operator error. A human-factors analysis should
= be performed which considers:
(a) the use of this information by an operator during both normal and abnormal plant conditions,
{b) integration into emergency procedures, (c) inteqration into operator training, and fd) other alarms during emergency and need for prioritization of alarms.
Each piece of equipment is further discussed below.
18 1-82
SSES-PSAR
- 8. 1.30 1. 1 Noble Gas Ef fluent Nonitor Noble gas ef fluent monitors shall be installed with an extended range designed to function durinq accident conditions as well as during normal operating conditions. Multiple monitors aze considered necessary to cover the ranges of interest.
Noble qas effluent monitors with an upper range capacity of 10~ pCi/cc (Xe-133) are considered to be practical and should be installed in all operating plants.
(2) Noble gas effluent monitorinq shall be provided for the total ranqe of concentration extending from normal condition (as low as reasonably achievable concentrations to a maximum of 105 p Ci/cc (Xe-133) .
Hultiple monitors are considered to be necessary to cover the ranges of interest. The range capacity of indi;vidual monitors should overlap by a factor of ten.:
Licensees and licensing applicants should have available for review the final design description of the as-built system, including piping and instrument diagrams together with either (1) a description of procedures for system operation and calibration, or, (2) copies of proceduzes for system operation and calibration..
License applicants will submit the above details in accordance with the proposed review schedule,'ut in no case less than four.
months prior to the issuance of an operatinq license.,
18.1,30.1. 2 - Sampligg and Analysis of Plant Effluents Because iodine gaseous effluent monitors for the accident condition are not considered to be practical at this time, capability for effluent monitoring of radioiodines for the accident condition shall be provided with sampling cond.ucted by adsorption on charcoal or other media, followed by onsite laboratory analysis.
Licensees shall provide continuous sampling of plant gaseous effluent for postaccident releases of radioactive iodines and particulates to meet the requirements of Table II.P.1-2 in NUREG 0737. Licensees shall also provide onsite laboratory capabilities to analyze"or measure these samples. This requirement should not be construed to prohibit design and development of radioiodine and particulate monitors to provide online sampling and analysis for the accident condition. Xf gross gamma radiation measurement techniques are used, then provisions shall be made to minimize noble gas interference.
18 1-83
SSES-FSAR The shielding design basis is given in Table II.F. 1-2 o f NURZG 0737.'he sampling system design shall he uch that plant personnel could remove samples, replace sampling media and transport the samples to the onsite analysis facility vith radiation exposures that are not in excess of the criteria of GDC 19 of 5-rem vhole-body exposure and 75 rem to the extremities durinq the duration of the accident.
The design of the systems for the sampling of particulates and iodines should provide for sample nozzle entry velocities vhich are approximately isokinetic {same velocity) vith expected induct or instack air velocities. For accident conditions, sampling may be complicated by a reduction in stack or vent effluent velocities to below design levels, makinq it necessary to substantially. reduce sampler intake flov rates to achieve the isokinetic condition., Reductions in air flow may well -be beyond
'the capability of': available sampl'er'flov controllers to'maintain isokinetic conditions; therefore, the staff will accept flow control devices vhich have the capability of maintaining isokinetic conditions with variations in stack or duct design flov velocity of +205. Further departure from the isokinetic condition need not be considered in design. Corrections for non-isokinetic sampling conditions, as provided in. Appendix C of ANSI 13.1-1969 may be considered on an ad hoc basis.
Effluent streams. vhich .may contain air,with entrained water, e.g.-
air ejector discharge, shall have provisions, e.g., heaters, to ensure that t1ie adsorber is not degraded while providing a representative.. sample..
license applicants vill submit final'esign details in accordance with the proposed reviev schedule, hut in no case less than four months prior to the issuance of an operating license.
18,1,30. 1,3 ~
Containment High-Range Radiation Monitor In containment radiation-level monitors with a maximum range of 108- rad/hr shall be installed. A minimum of'two such monitors that are physically .separated. shall be provided. Monitors shall be developed and'ualified to function in an accident environment.
The specification of 10~ rad/hr. in the above position =was based on a calculation of postaccident containment. radiation levels that include, both particulate '{beta)., and photon '(gamma) radiation., A radiation detector that, responds to both beta and "gamma radiation cannot be qualified to post-LOCA (1'oss-of-coolant accident) containment environments hut gamma-sensitive instruments can be so qualified. In order to follov the course of an accident, a containment. monitor that measures only gamma 18 ~ 1-84
radia+ion is adequate. The requirement. was revi"ed in ..he October 30, 1979 let+ez to provide fo- a pho-'on-only measuzemen.
with an upper zanqe of 10'/hr.
The monitors shall be loca+ed in containmen (s) in a manner as to provide a reasonable assessment of area zad'atior. conditions inside containment. The monito s shall be widely separated so as to provide independent. measurements and shall ~'view" a la ge f-action of the con+ainment volume. loni+ors should no . be placed in areas which are pro ected by massive shielding and should be reasonably accessible foz replacement, maintenance, or calibration. Placement high in a reactor buildinq dome is not recommended because of potential ma'ntenance difficulties.
The monitors are reaui ed to respond to gamma photons wi+h enezqies as low as 60 keV and +o provide an essentially flat response for gamma energies between. 100 keV ar.d 3 HeV, as specified in Yabl'e II.F.1-3 of NUBFG 0737. i1onitors that use
".hick shield 'g to increase the upper ranqe will under-estima+e postaccident radiation levels in containmen by several orde"s of magnitude because of their insensitivity tc low energy aammas and are not acceptable.
License applicants will submi+ the zeauired documentation in accordance with the appropriate review schedule, but in no case less than four months prior to the issuance oi the staf f evaluation epor for an operatinq license.
- 18. 1. 30. l . 0 Con tain men 0 Pressure 'lonit or A continuous indicatior. of con ainment pressure shall be provided in the control room of each operating reactor. measurement and indication capability shall include three times the desiqn pressure of, the containment for concrete, four times the de iqn pressure for s+eel, and -5 psiq for all containments.
Operatinq license applicants with an operating license dated before January 1, 1982 must have des'n changes completed hy January 1, 1982; those applicants with license dated after January 1, 1982 must have all desiqn modifications completed before they can receive their opera inq license. Documentation is due 6 months for the expected date of operation.
$ 8. 1. 30. 1. 5 Containment Mat.ez Level monitor continuous indication of containment water level shall be provided in the control room for a11 plants. A wide ranqe instrumen+ shall be provided +o cover the range from the bottom to 5 feet above the normal water level in the suppression pool.
SSES-FSAR The containment vide-range water level indication channels shall meet app=op"iate desiqn and qualif.cation c"itezia. The narrow-ranqe channel shall meet the requirements of Regulatory Guide 1.89.
For BJB pressure-suppression containments, the emergency core coolinq system suction line 'nle s may,be used as a startinq reference point or the na row-"ange and wide-ranqe wate" level monitors, instead of the bottom of the suppression pool.
The accuracy requirements of tl e wa er level monitors shall be provided and justified to be adequate for their intended funct on.
Operatinq license applicants with an operatinq license date before July 1, 1981 must have desiqn changes completed by July 1, 1981, whereas those applicants with license dates past July 1, 1981 mu t have all design modifications completed before they can receive their operaitng license.
Submittals from operatinq reactors licensees and applicants for operatinq licenses {with an operating license date before January 1, 1982) shall be provided by January 1, 1982. Applicants with operatinq license dates beyond January 1, 1982 shall "provide the required design information at least 6 months befo e the expec ed date of operation.
18.1.30.1.6 Containment Hydrogen t<onitor A continuous indication of hydrogen concentration in the containment atmosphere shall be provided in the control zoom.
Measurement capability shall be provided over the range of 0 to 10% hydroqen concentration under both posit've and neqa.ive ambient pressure.
Operating license applicants with an opezatinq license date before January 1, 1982 must. have design chanqes completed by January 1, 1982 must have all design modifications completed before they can receive their operatinq license.
Opezatinq reactors and applicants fo" operating license receivinq an operatinq license before January 1, 1982 will submit documentation before Januazy 1, 1982. Appl-'cants with operatinq licen e issued after January 1, 1982 shall provide the required desiqn information at least 6 months prier to th expected date of operation.
- 18. 1-86
SSHS-FS AB 18.1.30.2 In.eroretation NoneI I
~
r I required.
- 18. l. 30. 3 Statement of Response The response for each equipment requirement is qiven below. All equipment will be installed hy the required dates. A human factors evalua+ion will be pe formed for changes that involve control room ins.r>>mentation. Drawinqs showing the loca.ion of equipment were submitted in a letter from N. R. Curtis +o A.
Schwencer on June 15 {PLA-842) .
For modifications to plant sys ems and components such as addition of new post-accident monitorinq capability, procedures are developed or revised as necessa y and appropriare trainirg is provided when he final design documents are approved and the equipment is available for use.
18.1,.30. 3. 1 Nob1.e Gas Hf fluent Nonito" Hach of the five plant vents are monitored by an Fberline Nodel FAAM {Fixed Airborne Activity Nonitor) . The FAAN's analyze representative samples which are provided Ly isokine+ic probes which are in compliance with ANSI 13.1-1969. Each FAAN has three noble qas detecto s which provide ove lappina -anges of 1 x 10-7 Ci/cc to 1 x 10~ 9 Ci/cc for Xe-133 qas. The sample stream is filtered by a H"PA filter and a charcoal f ilter, which are con+ained in a SA-13 assembly before passing the noble qas detectors. The charcoal filter can be replaced with a silver zeolite filter when required.
The plan+ effluent noble qas data is continuously monitored and stored in solid state memory. The flow through +he sample line is also measured and stored in solid state memory. The FAAN then calcula+es and stores activity per unit, of volume. This information can be displayed upon request. and is periodically printed out for record keepinq purposes. This information is disolayed and recorded on back"ow panel 1C669.
fliqh activity alarms for the reactor and turbine buildings are annunciated on control room front row panel 1C651. High activity alarms for the Standby Gas Treatment System are annunciated on control room front row panel 1C601.
The low-ranqe noble qas channel is calibrated using Kr 85 and Xe 133 aas tandards traceable to the National Bureau of Standards..
SSES-FSAP.
The mid-ranqe noble qas channel is calibrated usina a Cs 137 stick-source. The hiqh-ranqe noble qa channel is calibrated u inq'a Kr 85 qas s,.andard traceable o the National Bureau of Standards.
The system is Powered from non-class IE instrument AC power.
independent battery backup is provided which is capable of providinq powe" for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
18,1,30.3.2 Samolina and Analysis of Plant Ef=luents Each of the five plant vents has a continuous isokine.ic sample drawn from it in accordance with ANSI-N13.1. Each sample is then taken through short runs of heat traced tubinq to a Eberline Model FAAN (Fixed Airborne Activity monitor) . In the FAAil the sample stream hen passes th ough a HEPA filter which removes particulates. Opon leavinq the HEPA filter the sample s,.ream passes through a charcoal filter which removes iodines. When required this filter can be replaced with a silver zeolite filter. Capabilities for purging the sample line with compressed air are provided under manual control. The sample s.ream is next measured f cr noble qas activity and then "eturned to the plant vent. Durinq normal operation the HEPA and charcoal filters are monitored Ly radiation detectors and this information is presented to the operator in the control rccm. ~Jnder accident conditions these detectors will saturate and the filters must be removed, placed in a shielded container, and analyzed in a laboratory. The FAAtl also has provisions for ot-.aining a qrab samples~
The isokinetic sample is in compliance with ANST.-N13.1-1969. To accomplish this, each vent has an air profile (final qa .
treatment) station to eliminate turbulent and'ctatinq qas flow.
The averaqe stack velocity and volume are then measu"ed by means of a multipcint, self-averaqinq Pitot transverse station. An air flow controller theh simultaneously withdraws a multipoint sample under isokinetic flow conditions by means cf an isokinetic sample rack. This isokinetic sample is then directed to the Final Airborne Activity iionitor.
The system is designed such that plant personnel can remove samples, replace sample media and transport the samples in shielded conta'ers to an analysis facility. Radiation exposures for this process are not in excess of 3 rem whole-bcdy exposure and 18.5 rem to the extremities during the duration of the accident. These exposures are based on the proper use of plant procedures f or removinq sample media and doses f rom the shielding s udv presented in Section 18.1.20.
- 18. 1-88
SSZS-FSAR procedures fo" analyzing samples both normal and accident condi ions are described in Subsection 12.'.3.5.5. The equipment used to analyze these samples is described in Subsection 1.2.5.2.7.1. Additional instrumentation and procedures for samplinq and analyzinq implant iodine a" e described in Subsection
- 18. 1. 70.
18.1.30. 3. 3 Containment High-Range Radiation Monitor Redundant Class 1E in-containment radiaticn monitors are Provided. The mone tors are General A~omi c high range radiat ion monitors. These monitors are capable of measurinq radiation levels of 1R/h o 1 x 108 R/h" (Gamma) for photon energies of between 80 KeV to 3 NeV. An accuracy of + 20% is obtained on lower decades.
The detec o-s a "e unshielded and physically sepa at".d on opposite sides of the reactor pressure vessel.
Logarithmic indicatinq recorders are provided for Channels A and 8 on front row panel 1C601.
A common red hiqh radiation annunciator fo both channels is provided on control room front row panel 1C601. A common white system trouble 1'qht is also provided for both channels on control rocm ront row panel 1C601.
The containment radiation monitorinq system is designed to be safety qrade. This equipment is qualif'ed to XZEE-344-1975, IEEE-323-1974 and NUREG-l.588 in accordance with the Commission order on ilay 23rd, 1980 (CLI-20-81) .
18.1. 30.3. 4 Containment Pre. sure ihdonitcr Two Class lE redundant drywell chamber pressure measurements is provided as follows:
SERVICE RANGE LOCA Ran qe 0 to 65 psia HI Ranqe 0 to 250 psiq The LOCA and HI ranqes are divided into two division Continuous, individual indica ion of all four Division I and II pressure measurements are provided by indicating recorders for the operation on front row panels 1C601.
18 1-89
SSFS-FSAB Normal operatinq pressures in the drywell and wetwell are mor.i.ored by a -1 to +3 psiq instrumen+. installed 'n each chambe . An indicator on con+ ol panel lC601 displays these I
pressures. A selector switch is provided 'to allcw the operator to monitor either drywell oz wetwell pressuzee These instruments are non-safety grade with the exception of the transmit.+ers, which are designed to meet con+ainmen+ p"essure hour.dary service.
The accuracy of +hese instruments is + 2'5 of full scale.
The containment accident ranqe pressure mor.itors are desiqned to be safety qrade. This equipment are qualified to TEEE-344-1 975, IEEE-323-1974 and NUBEG 0588 in accordance with the Commission order on Nay 23rd, 1980 (CLI-20-81).
- 18. l. 30.3. 5 Contai nmen t Mater Level Monitor I
Bedundant wide and narrow ranqe safety q ade instruments are installed to continuously monitor suppressicn pcol water level.
The channel A measuremen+s will be displayed on control room front row panel 1C601. The chanr.el 8 measurements will be recorded on front row panel 1C601.
The narrow ranqe instruments measure between 18 and 26 feet. The wide range instruments measure between 4.5 and 49 feet. This covers the requ.zed range of from the lowest ECCS suction to 5 feet above normal ~ater level. Normal water level is approximately 23 feet.
The accuracy of these instruments is + 2j of full scale.
18.1.30.3. 6 Containment Hydrogen i<oni oz Cont'nuon and redundant indication and "ecoril inn of hyd" oqen are provided on control room f ror.t row panel 1C601. These instruments have a ange of 0 to 30K.
The containment hydroqen monitorinq system is designed to be safety grade. The equipment are qualified to ZEEE-344-1975, IEEE-323-1974 and NUBEG-0588 in accordance with the Commission order on Nay 23rd, 1980 (CLI-20-81) .
The accuracy of these instruments is + 2% of full scale.
18.1.31 INSTBUNENTATION FOB DETECTION OP INADEgUATF. COBE COOLING 3II-2=2)
- 18. 1-90
- 18. 1.31.1 Statement of Requirement Licensees shall p ov'e a desc'ption of any addi+ ional instrumen+ation or controls (primary or backup) proposed for the D lant .o supple ment existing instr umentat ion (includinq primary coolant sa+ura+ion monitors) in order to provide an unambiquous, easy-+o-in+erpret indication of inadeaua+e core cooling (ICC) .
description of. the functional design reaui ement'or the system shall also be included. A de cription cf the procedures to be used wi.h the proposed equipment, the analysis used in developing "hese procedures, and a schedule for installing the equipment shall be provided.
18.1.'31.2 Intergg~tation None regu red.
$8 3 . 3$ . 3 STATEMENT OF RFSPONSE
- 18. 1.31.3. 1 Introd uction PPGI. has participated in the BNR Owners 'roup (B'rt ROG) study to specifically address ICC concerns. The purpose of the study was to evaluate means of providinq reliable information to detect:
the approach towards ICC, the exis ence of ICC, and the retu "n to adeauate core cooling. The study considered local and core-wid e ICC, the reliability of existirq in. trumen a..ion, and t he impac+
of additional instrumentation.
>he BMHOG first evaluated the relationship between reactor water level and adequate core coolinq. In order to clearly demonstrate that reactor water level is a viable indicator of ICC and due to the complexi+y of the isue, the BNROG scoped work in".o wo activities. The first was to evaluate the reliability of. the existinq BMH reactor water level measurement systems. The second was a study of ICC. The report resultinq from the first part was transmitted to NRC in a lette from T. J. Dente to H. R. Denton on Auqust 13, 1982. The report resulting from the second part was transmi+ted to NRC. These reports provide substantial evidence to conclude that reactor water level is the most suitable parame+er for operational contrcl to avoid and mitiqate ICC.
- 18. 1. 31.3 2 Reac+or Pater Level Instrumentation Report
SSL'S-FSAH The BMHAG studied four types of reactor water level ins+rumentat'on which aze representative of existinq design".
The conclusion of he study was +hat the instrurenta-ion used "throuqh manv years of operating exper'ence ha ve demonstra ted very hiqh degrees of capability to provide required information in va"ious- cond'tions of reactor operation. Almost without exception, the information presented tc +he operator is not ambiquous, and trips, initiations, and othe r signals taken from
'the level measurement systems have cccurred as require~."
The report qoes on however to note a few reported events resultinq in spurious siqnals and erroneous info mation -.o the opera+or, none of which resulted in serious ccnsequence . The report indicates the desirability of an overall reassessment of the level system vulnerabilities aqainst a list of potential aroas o+ improvement. The report concludes =hat "no modifications should be made to any specific system until a thorouqh plant specific analysis is conducted.
Interaction of systems in a specific plant design can siqnificantly affect the degree of desiqn chanqe necessary to improve a system and may possibly demonstrate that a design chanqe is not required.>'8
- 1. 31. 3,3 Inadequate Core Cooling Report The followinq concepts are extracted from the report on ICC.
- 1) Defin'tion of ICC ' terms of fuel and clad peak temperatures.
Clad tempe aturesin the range of 13000F to 15000F may 1'kely result in the release of qaseous fission products in the fuel to clad qap by means of perforation produced by weaken'nq of the f>>el cladding.
At temperatures in excess of 18000F, clad me"..al-water chemical heat reaction commences and accelerates the heat rate. The report suggests tha. ICC miqht be defined as reachinq peak temperatures between 13000F and 18000F in an averaqe fuel bundle.
The relationship amonq reactor power, coolant inventory (water .level), and recirculation flow which results in ICC is developed. The most extens've development of ICC results from operation at critical heat flux within the normal operating power ranqe. Critical heat flux is treated extensively in presen+ safety analyses and its occurrence is prevented by a ubstan+ial requlatozy methodoloqy includinq power-flow tzip lines, limiting powerdistribution, and reactor trip systems. This
- 18. 1-92
SSZS-FSAR rationale is ex+ended down to zero flow and zero powerincludinq TCC conditions which may accompany high void f ract ion pum pod recircula tion flew.
From the above, the zepor: concludes that the ICC requirements oz NUBFG 0737, i+em XI.F.2 and ?egulacory Guide 1.97 were no+ meant o bo applicable to normal power "ange operation critical hea;. flux conditions, hu+ apply to BHB's only at decay power condi+ ons.
Hater level as an indicato of ICC.
Applyinq only decay power condign,ions, a scenario was developed based on =eactor scram, recirculation pump trip, reactor pressure vessel isolaticn, and loss of all makeup water systems (saf et.y and non-sa fetassumed y) . The steam produced by sensible and decay heat i to be lost from the reactor pressure vessel at constan.
pressure. Th'4 time hi. tory of water level in this condition is =hown in Fiqure 18.1-14. The relationship between watez level and peak cladding tempera+ure (which is an indicator of ICC) is shown in Fiqure 18.1-
- 15. Sensitivity of this relationship to core uncovery times is shown +o be very flat (see Figure 18.1-16) .
The assumption of a constant pressure (1,000 psia) was shown to be conservativeas compared tc a similar scena "io at low pressure (100 psia) and a saw too h shaped pressure function indicative of periodic safety/relief valve operation.
Accordinqly it is concluded tha+ reactor vessel water level is a valid indicator of ICC includinq approach to, existence of, and re+urn f rem those condition I
Local ve sus global detection of ECC.
A literature review indicated that co-e damaqe will not propaqate once the core is recovered with wa.er. A scenario is postulated that results in local fuel damage durinq t.he existence of global IFF, where the b'lockage prevents ubsequent cooling of the damaged channel. Damage propagation subseauent t,o global ICC recovery will be restricted to those bundles where sufficient fuel damaqe occurred duzinq the global ICC to totally cut off the bundle water flow after recovery.
The use of. instrumentation to detect this existence of local ICC was considered and rejected because bundle damage sufficient to cause complete blockage of cooling subsequent to recovery would also destroy any instrument placed .herein.
Addit ional Tnstrumentar ion.
SSFS-FSAH In addition to water level,:here are a number of other existinq instrument systems which prov'de infor mation relative to the question of ICC. These include core spray flow rate, flows to and from the reactor vessel, primary containment radiation levels and hydrogen concen ration levels, and activity sampling in reac+o" coolant ~ater and the supp"essicn pool.
Risk significance of ICC.
The contribution of water level measuremen'ystem fai3ure to core melt p obahili y was ovaluated based on modifvinq an existinq PRA for a BMR;4 plant wi.h NAHK II containment. The basic approach was to modify the event trees to identify the risk contributed by the wate" level system. Najo" conce"ns conside"ed were:
loss of level indication due to loss cf'eference leg under h'qh dzywell temperature and low vessel pressure conditions; concurrent or commcn failures of level instruments, and reference leg breaks. The results a=e considered to be representative of the Susquehanna desiqn.
lt is shown that water level measurement failures contribute less than 13'K of the overall probability of core melt. Improvements in the level measurement svstem can reduce the contribution of level instrument failure to overal risk down to 13'$. These improvements include reduction or mit'gation of errors caused by hiqh drywell temperatures, validation of level signals, and increas'nq the probability of timely ADS operation by manual actuation. Susquehanna has combined elements of these imp ovements in its design includinq reduced and equal vertical drops within primary containment: for both the reference and variable legs of the multiple ins+rument channels which mitigate the effects of high drywell temperatures. In addit'on, the Susquehanna Emergency Operatinq Procedures (which are based on the BWBOG Emergency Procedure Guidelines) provide assurance of timely manual ADS operat ion.
Cost/Benefit of Additional Instrumenta"ion.
An evaluation of alternative or diverse means of detectinq ICC was conducted. Thirty-,hree concepts, listed in Table 18. 1- 16 were evaluated with many of these concepts beinq di carded afte the preliminary evaluation. Finally, four devices were selected for further evaluation of performance and cost. These devices included: in-core thezmccouples in the LPHll tubes; heated junction thermocouples as a point level measurement inside the LPBt< tubes; s+eam dome thermocouples; source range moni.ors as an ICC detection ~device. A cost/benef it ana ly es, described
- 18. 1-94
SSES-L" S AR in the reoort, was perfo med cn these instrument system add'tions u"irg a tech nique proposed in SECY 513 "Plan for Early Recoqni+ ion of Safety Issues:
August 25, 1981. The results of :hat analysis showed tha+ +he addit'on of alternat ive ICC detection devices could be assigned a low prio" ity when comparod to other LNR safety issues.
- 18. 1. 31. 3. 4 Conclusior.
The BMROG study shows that knowledqe of water level withir. +he core is uniquely suitable and suf icier.-,. for the monitoring of the adequacy of core coolinq under accident conditions. The existinq wa" er 1evel measurement systems are highly rel'able systems in providinq information to the operator but tha.
individual level measurement systems should he evaluated for possible imp ovements particularly with regard to loss of drywell coolinq (which can produce flashing) and inst umen. line breaks.
Modifications can be made to reduce the probability of reactor water level irstrumenta ion failure and thereby decrease its contribution to core melt from 13% to 3g. However, the Susquehanna desiqn already includes a significant port'n of the improvements identi ied by the BVROG.
The addition of backup, diverse ICC detection devices is shown to have a very small additional contribution +o overall risk "eduction. Further, the safety priority analysis of .hese devices indicates a score in the lower end of the lcw pr'ority ranqe. Therefore, no additional instrumentation should he considered recessary for the detection of ICC because of its neqliqible cont"ibution to plant safety.
Symptom based procedures have been developed and implemented a+
Susquehanna Unit 1. These procedures vill assis the operator in Subsection 18. 1.8 for detectinq the a@proach to ICC. Refer +o the response to equirement I.C.l.
In addition,'PGL has developed a D'splay Cont"ol Sub-system (DCS) format to promote operator detection cf inadequate core coolinq. The format consists of three dis+inct functional areas:
a qraphic representation of reactor water level, a twenty minute reactor water level trend, and wate level supportinq data.
The qraphic display will provide a qualitative representation of reactor water level from -150 to + 170 inche= relative .o instrument level zero. Seve"al vessel components are statically depicted as points of reference. The water level indication is normally displayed in yellow, however, of level decreases to or below -38 inches it will tu rn f rom yellow to red.
- 18. 1-95
SS ES- FS AR The reactor wa+er level trend portion of the display will provide a twenty-minute his ory, in one minute inc=ements, of + he water
.rend. Slowly increasinq or decreasinq levels should be apparent.
from thi= trend. The trend display wi'-1 t u=n f "om yellow to ro.d if the level decreases to or below -38 inches.
Other supportive data, which may be useful in moni, orinq reactor water level, has also been provided.
The format is subject to possible revisions or refinements, however, the fundamental concept of qraphically indicatinq reactor water level will always be provided by the display. A typical format sample is provided in F'qure 18.1-13.
18,1.32 EMERGENCY POHFR FOH PHFSSUP'IZEB EQUIPMENT QI'I. G. 1}
This requirement is not applicable to Susquehanna SES.
18 ~ 1. 33 RP.VIEW ESF VALVES /II.K. 1. 5$
No requirement stated in NURFG 0737. Refer to Subsection 18.2. 25 which contains the response to the equi ement in NUHEG 0694.
- 18. 1 34 OPFHABILITY STATUS /II.K. 1. 10$
No requirement stated in NUHEG 0737. Refer to Subsection 18.2.26 which contains +he response to the "equi"ement 'n NUBEG 0694.
18-1 35
~ TRIP PRESSURIZER LOW-LEVEL COINCIL'ENT SIGNAL HISTABLES (II. K. 1.
17'his requirement is not applicable to Susquehanna SES.
18 1.36 OPERATOR TRAINING FOR PROMPT MANUAL REACTOR TRIP (II.K. l. 20)
This requirement. is not applicable +o Susquehanna SFS.
- 18. 1-96
SSZS-FSAR
- 18. l. 37 AUTOMATIC SAFFTY GR ADF ANTICIPATCRY REACTOR TBXP gI X. K. l. 21}
This requ'rement is not applicable to Susquehanna SES.
18, 1,38 AUXILIAFY ff FAT REMOVAL SYSTEM. PROCEDURFS /II-K-1 22}
No requiremen+ stated in NUREG 0737. Befe to Subsection 18.2.30 which contains the response to the requirement in NUREG 0694.
3,8 ~ 1,39 /FACTOR VFSSHL LEVEL PROCEDURES /II.K.1.23}
No requirem nt stated in NUREG 0737. Refer -o Subsection 18.2.31 which con+ains the response to the requirement in NUREG 0694.
- 18. 1.40 COMMISSION ORDERS ON BABCOCK AND WILCOX PLANTS /IX. K. 2}
The e requiremen+s are not applicable to Susquehanna SFS.
This requirement i not applicable to Susquehanna SES.
- 18. 1.42 RFPORT ON POHEB-OPERATED BELIE." VALVE FAILURES I.
(I K. 3. 2}
This requirement is not applicable <<o Susquehanna SES.
18.1.43 REPORTING SAFETY/RELIEF VALVE FAILURES AifD CffALLF.VGES /II K. 3. 3}
No requirement stated in NUREG 0737. Refer to Subsection 18.2.33 which contains the response to the requirement in NUREG 0694.
- 18. 1-97
SSES-FSAH
- 18. 1.44 AUTOiIATIC TRIP OF HFACTOH COOLANT PU 1PS DURING A LOCA (II. K. 3. 51 This rrequirement is not applicable .o S'usquehanna SES.
18.1.45 EVALUATION OF POWER-OPERATED BELIEF VALVF, OPENING PROBABILITY /II.K. 3.7g This equiremen.. ' not. applicable to Susquehanna SES.
18.1.46 PROPORTIONAL INTEGRAL DFRIVATIVE CONTROLLER MODIFICATION /II K. 3. 9)
This requirement is not applicable to Susquehanna SES.
- 18. 1. 47 PROPOSED ANTICIPATORY TRIP NODIFICATICN /II.K. 3. 10$
This requirement is not applicable to Susauehanna SES.
/
- 18. 1. 48 PO'eJEF.-OPERATED RELIEF VALVE FAILURE RATE /II.K. 3. 11}
This requirement is not applicable to Susquehanna SLS.
- 18. 1.49 ANTICIPATORY REACTOF TRIP ON TURBIWF. TRIP /II K. 3. 12$
This requirement is not applicable to Susquehanna SES.
18.1.50 SEPARATION OF BIG(( PRESSURE COOLANT INJECTION AND REACTOR COBE ISOLATIOVi COOLING SYST EN INITIATION LEVELS (I l. K. 3. 131
- 18. l. 50.1 Statement of. Requirement Currently, the reactor core isolation coolinq (BCIC) system and t.he hiqh-pressure coolant injection {HPCI) system both initiate on the same low-water-level siqnal and hoth isolate on the same hiah-water-level siqnal. The HPCI system will restart on low water level but the RCXC system will not. The BCIC system is a
- 18. 1-98
SSES-FSAR low-flow sys+em when compared to the HPCI system.
1evels o the HPCI and RCIC system should be separated so that The 'it iation the RCIC system initiates at a higher wa er level than the HPCI system. Further, the ini+'ation logic of the RCIC syst em should be modified so that the RCIC sys" em will res..art on low wat,er level. These chanqes ha ve the poten-.ial tc "educe the number o f challenqes to "he HPCI system and could result in less stress on the vessel from cold water injection. Analvses should performed to evaluate these chanqes. The analyses shou ld be submit'ed to the NRC staf f and changes should b= implem e n te<'. if justified by the analyses.
All applicants for operatinq license should submit the results o" an evaluation ar:d proposed modificatior. fcur mcnths prior to the expected issuance of. the sta f safety evaluation report for an operatinq lice'nse or four months prior to the listed implementation date (July 1, 1981), whichever is la er.
18.1.50.2 Interpretation None required.
18 1.50.3 Statement of Response PPGL concurs with the BWR Owners~ Group posi.ion on the separation of +he HPCI and RCIC .etpoints which was .ransmitt d to the NRC by lette" from R. H. Buchholz (GE) to D. G. Eisenhut (NRC), October, 1, 1980 (<FN-169-80) .
This letter forwarded a GE study which shcwed that HPCI and RCIC initiations at +he curren+ low wa:er level setpoints is within the desiqn basis thermal fatigue analysis cf the reactor vessel and its internals. Separat inq HPCI and RCIC se-..poin+s as a means of reducing thermal cycles has been shown to be of neqliqihle benefit. In addition, raising the RCIC setpoint or lowering the HPCI setpoint have undesirable consequer.ces wh'ch outweigh the benefit of the limi.ed reduction in thermal cycles. The"efore, when evaluated on this basi, PPGL conclude that no chanqe in RCIC or HPCI setpoints is required.
PPGL also concurs with the B<ilR Grcup position that RCIC restart automatically followinq a trip of the system at Owners'hould hiqh reactor vessel water level. This position was transmitted to the NRC by le+ter from D. B. Maters (8MROG) + c D. G. Rise. nhut (NRC), December 29, 1980.
PPGL will implement the recommended option 2 which is described in detail in the GF. study f orwarded with the B~~ 8 Owner" 'roup
- 18. 1-99
SSF,S-FSAB posi ion. Implementation is discussed in a letter from N.
Curtis to B. J. Younqblood on Pay 20, 1981 (PLA-792).
18.1.51 tiODIFY BREAK-DETECTION LOGIC TO PFEVZNT SPURIOUS ISOLATION OF HIGH PHESSVHE COOLANT INJECTION AND RFACTOR COBE ISOLATION COOLING /II.K.
3.15'8,1.
- 51. 1 S atement of Requirement The hiqh-pressure coolant injection (HPCI) and reactor core isolation cooling (HCIC) systems use,differential Pressure sensors on elbow taps in the steam lines to their turbine drives to detect and isolate pipe breaks in ~he systems. The pipe-break-detection circuitry has resulted in spurious isolation of the HPCI and HCTC systems due to the p essure sp'ke which accompanies startup of the systems. The pipe-break-detection circuitry should be modified so tha+ pressure spikes resulting f rom HPCI and HCIC system initiation will not cause irad vertent sys+em isolation.
All applicants for operatinq license should submit documentation four months prior +o the expected issuance of the staff safety evaluation report for an operatinq license or four month" prior to the listed implementation date (July 1, 1981), whichever is late 18.1. Sl. 2 Inter oretat ion None required.
$ 8.1.51.3 Statement of, Response The BHB Owners'roup has performed an evaluation and recommends the followinq modification to the steamline break detection loqic. In order to minimize inadve"tent HPCI/HCIC isola ion due to pressure transient" durinq system initia.ion, a time delay relay, set at approximately three (3) seconds, has been installed in the steamline high differential pressure ci"cuitry. The time delay featu e assures that the steamline b.eak isolation signal is, in fact, due to continuous high steam flow. See Subsections
.7. 3. 1. 1a. 1. 3. 4 and 7. 6. 1a. 4. 3. 3. 42.
The time delay relay is class lZ, with an adjustable time delay se+tinq of 0-5 seconds. This classificaticn is compatible with the system's existinq circuitry. Two time delay relays are
- 18. 1-100
SS ES-.FS AH required for the trip system loq.'c for hoth the HPCI and RXCT systems ~
A desiqn assessment s study shall confirm the app'opr'ate ~ime-delay settinq. Implementa t ion i= discussed in a letter from N.
H. Curtis to B. J. Younqblood on Hay 20, 1981 (PLA-792) .
- 18. 1.52 FFDUCTION OF CHALLENGFS A'8D FAILURES O." BELII'.F VALVES
/II. K. 3. 16) 18.1.52.1 Statement of Requirement The record of relief-valve failures to close for all boilinq-water reactors (BPRs) in the past 3 years of plant operation is approximately 30 in 73 reactor-years {0.41 failures per reactor-year). This has demonstrated that the failure of a relief valve to close would be the most likely cause of a small-break loss-of-coolant accident (LOCA). The high failure rate is the result of a hiqh relief-valve challenqe rate and a relatively high failure rate per challenqe (0.16 failures per challenge). Typically, five valves a e challenged in each event. This "esult. in an eauivalent failure rate per challenge of 0.03. The challenge and failure rates can be reduced in the followinq ways:
(1) Add'tional anticipatory scram on loss of feedwater, (2) Bevised relief-valve actuat'on setpoints, (3) Increased emergency core cooling (ECC) flow, (4) Lowe" opera+inc pressures, (5) Ea"lier initiation of ECC systems (6) Heat removal throuqh emergency condensers, (7) Offset valve setpoints to open fewer valves per challenge,
{8) Installa+ion of additional relief valves with a block- or isolation-valve feature to eliminate opening of the safety/relief valves (SBVs), consistent wi h the ASi)E Code, (9) Increasinq'he high steam line flow setpo't for main steam
.line isola tion val ve { HS I V) closure, (10) Lowerina the pressure setpoin+ for NSIV closure, (11) Beducing the testing frequency of the HSIVs,
SSF,S F.SAB (12) Ao e-s-..ringent valve leakage c" i e ia, and Q3) Farly 1 removal off leakinq valves An investiqation of the feasibility and ccntrai ndications of.
reducinq challenqes to the relief valves by use of the aforementioned methods should be conducted. Other methods should also be included in the feasibility study. Those changes which are. shown +o reduce relief-valve challenaes without corn promisinq the performance of the relief valves or o+he: systems should be implemented. Challenqes .o +he relief valves should be reduced substantially (by an order of magnitude).
il Results of the evaluation shall be submit.ed by Ap" 1, 1981 fo" s.aff review. The actual modification shall be accomplished durinq the next scheduled refuelinq outaae following staff approval or no la+er +han 1 year followinq s,.aff approval.
',Jodification +o be implemented should be doc>>men:ed at the time of i mplemen ta ion.
- 18. 1. 52. 2 interpretation None required.
18.1.52. 3 Statementt off Response The BNR Owners'roup (BaROG) has per formed an evaluation and developed recommendations to comply wi,.h this requirement. These ecommendations were transmitted by a letter from B. D. ttaters to D. G. Kisenhut on March 31, 1981. This evaluation shows that Crosby SP.Vs (as vill be installed in Susquehanna) have a probability of stickinq open which is approximately a factor of ten less than the three staqe Tarqe Bock valves. Lt is ou under tandinq that the qoal of this requirement is to reduce the probability of a stuck open SRV by a factor of 10 relat'e to a reference valve, which is the Tarqet Pock valve. Therefore we meet the in+ent of this requirement without modif ications.
Implementation of the modification proposed by the BRROG will not siqnificantly reduce this failure probability. Therefore no modifications are necessary in response to this requirement:.
- 18. 1-102
SSES-FS AH 18 1 53 REPOR 'N OUTAGFS OF EMERGHVCY COR COOLING SYSTEMS GATI K. 3. 17$
- 18. l. 53. 1 8 tat emen+ o f Recru irement Several components of the emerqency core-ccclinq (ECC) systems are permitted by +echnical specifica.ions ~c have suhstan ial outaqe time" (e. q., 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for one diesel-generator; 14 days fo the HPCI system) . In add'tion, there a=e no cumulative outage time limitations fo" ECC systems. Licensees should .,ubmit a report detailing outage dates and lenqths of cutages or all ECC systems for the last 5 years of operation. The ressort should also include the causes of the outaqes (i.e., ccntroller failure, spurious isolation) .
18.1.53.2 Interpretation None required.
- 18. 1. 53. 3 S ta tern en t o f Resoonse PPGL will subm'+ a report which summarizesr erne" gency core cooling system outaqes accumulated during the first five years of operation.
18 ~ 1 54 MODIFICATION OF AUTOMATIC DEPRESSURIZATION SYSTEM LCGIC (II.K.3.181 18.1.54.1 Statement of Requirement The automatic depressurization system (ALS) actua.ion logic should be modif ied to eliminate the need fc" manual actuation to assure adequate core coolinq. A feasibility and risk assessment study is required to determine the optimum approach. One possible scheme that should be considered is ADS ac uation on low reactor-vessel water level provided nc high-pressure coolant injection or hiqh-pressure coolant system flow exists and a low-pressure emergency core coolinq sy tern is runninq. This logic would complement, not replace, +he exist inq ADS ac,.uation logic.
Applicants for operatinq license shall provide results of feasibility studv 1 year prior to issuance of operatinq license.
SSZS-FS AR A desc"iption of t4e proposed modifica.ion for staff approval is required four months prior to issuance of an ope"ating license.
18.1.54.2 Tnternre+ation The ADS actuation loaic may not be au .omatically actuated for steam line break.. (SLB) outside containment. The operator must manually actua-.e the ADS after diaqnos'nq that an SLB has occurred. The ADS actuation logic should be modified to provide automatic actuation fo" all Design Basis Accidents.
- 18. 1.54. 3 Statement of Response PPGL has committed to the NRC (PLA-1312) tc mod'y the Automatic Dep essurization System (ADS) loqic in accordance w'th Option 4 of the BVROG study dated October 28, 1982. (Letter to Darrell G.
Kisenhuth NRC f rom T. J. Den+e BNR Owners 'roup BHROG-8260) . This op~ion bypasses the high drywell pressure port.'on of
+he current ADS actuation logic after a specific time interval and adds a manual switch which allows the oporato" to prevent an automatic ADS actuation. The additional loqic does not af feet automatic ADS response to pipe breaks inside the drywell. The analysis that led to the decision to implement this option is based on an assumption that the 2A fix is an acceptable resolution of the ATMS issue.
The hiqh drywell pressure reauiremen+ is bypassed by installing a second ("byoass") timer that is actuated cn low reactor water level (Level 1) . Hhen +his timer runs out, the hiqh drywell pres ure trip is bypassed and the ADS is initiated on a low reactor water level siqnal alone. A manual ADS inhibi+ swi ch is also provided to aid the operator in the execution of certain
..teps in the emergency Opera+inq Procedu es. "o inhibit the ADS with the current loqic, +he operator must continuously reset the two-minu e delay timer or tu n off all of the low pressure ECCS pumps. Thus, tho addition of a manually-operated inhibit switch would allow the operator to inhibit ADS ace,uation under ATHS condi+ions with a single action instead of having tc repeatedly reset the existinq two minute timer.
This option with procedural control provided by +he Fme qency Operatinq Procedures allows desirable operational control while
~
providing automatic actions in time to prevent excessive fuel heatup.
The HRC has concluded (letter from A. Schwencer to V. V. Curtis dated 4/25/83) that Option 4 's an acceptable method of modifying
- 18. 1-104
SSFS-FSAR
..he ADS loqic. However, t.he followinq additional in orma'ion is beir.q submitted to the NRC to complete tne review on Susquehanna:
Justification for the bypass timer set-;.inq.
b) A periodic testinq plan fo", he "imer.
c) Address the use of the manual inhibit switch in their emergency procedures.
A sur veillance plan for. the swi. ch.
As stated in a le+ter from N. M. Curtis tc A. Schwencer on June 17, 1981 (PLA-851), the requ'red system modifications will be installed prior to +he startup followinq the first r=fuelinq outaqe for Unit 1 and prior to fuel load for Unit 2 contingent on the results of the NRC review and contingent upon delivery of qualified equipmen
- 18. 1 55 RESTART OF COBE SPRAY AND LOH PRESSURE'OOLANT INJECTION SYSTFRS /II K 3 21)
- 18. l.55. l Statement of Requirement.
The core-sprav and low-pressure, coolant-injection (LPCl) system flow may be stopped by the operator. These systems will not restart automatically on loss of water level if an initiation signal is still present. The core spray and LPCZ system' loqic should be modified so that these systems will re .tart.,
requi=ed, .o assur~ adeauate core coolir.q. Becau e .his design modification affects several core-cooling modes under accident conditions, a prelim'na"y design should he submitted for staff review and approval prior to making the ac+ual modification.
Al.l applicants for operatinq license should submit documentation four months prior to the expec-ed issuance of an operatinq license or four months prior to the listed implemen+ation date, whichever i later.
18.3.55.2 Interoretation None required.
- 18. 1-105
SSES-FS AR 18.].55.3 Statement of Response PPGL concurs wi"h the BA'R Owners'roup position which was forwarded to the NRC by letter from DE B. Viaters (BKHOG) to D. G.
Fi enhut {NRC), December 29, 1980.
he BHROG repo"t states that the cur"ent ECCS design rep=esents the optimum approach to BNH safety. No modifications to existinq LPCI and core sp=ay systems are necessary in response to this requirement.
- 18. 1.56 AUTOMATIC SViITCHOVFB OF HFACTOR CORE ISOLATION COOLING SYSTFA SUCTION /II. K. 3.22/
- 18. 1.56.1 Statement of Requirement The reactor core isolation cooling (BCIC) system takes suction f om the condensate storage -ank with manual switchover o the suppression pool when,.he condensate s+oraqe tank level is low.
This swi.chover should be made automatically. Until the automa+ic switchover is implemented, licensees should verify tha clear and coqent procedures exist for the manual switchover of he HCIC system suction from the condensate storage tank .o .he suppr ssion pool.
Documentation must be submitted four months prior to issuance of the sta+ safety evaluation report or four mon" hs prior to the implementation date, whichever is later. Nodifications shall be completed by January 1, 1982.
18.1.56.2 Interoretation None required.
18.1.56.3 Statement of lesgonse Automatic switchover of the RCIC suction from the condensate storaqe tank (CST) to the suppression pool cn low CST level has been installed at Susquehanna SES.
- 18. 1-106
SSES-FSAR 18.1.57 CONFIBN ADEQUACY OF SPACE COOLING FOB HIGH PRESSURE COOLA NT IN JECTICN AND REACTOR CORE ISOLATION COOLING SYST FNS gI I. K. 3. 24/
18.1.57.1 S at.o.ment r of Requirement t Lonq-term operation of the reactor core isola..icn coolinq (RCIC) and hiqh-pressure coolant injection (HPCI) systems may requi e space coolinq to maintain the pump-coom .empera+u=es within allowable limits. Licensees should verify the accep:ability of the consequences of a comolete loss of alte na+inq-cu=rent (AC) power. The PCIC and flPCI systems should be desiqned to withstand a complete loss of offsite AC power to thei support systems, ncludinq coolers, for at least 2 hours.
All applicants for operatinq license should submit documentation fou" months pr'or to the expected issuance of he staff safety evaluation report for an opera+inq license cr four months prior to the listed implementation date, whichever is later.
- 18. 1.57. 2 Intergre+ation i tthat HPCI and RCIC room coolinq can be maintained to Confirm for enable continuous operation durinq a loss cf offsite AC power r
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
18.1.57.3 Sta .ement of Response The HPCI and RCIC room unit coolers and their s>>ppo t systems a"e desiqned +o withstand the consequences of a ccmplete loss of offsite AC power since these are powered from onsite diesel qenerators. Each HPCI and PCIC room i, provided with a 100%
capacity redundant unit cooler. Refer to Subsection 9.4.2.2.
18.1.58 EFFECT OF LOSS OF A TFRNATING-CURRENT POHEB ON RECIRCULATION PUf
Short-Term ADS Operation Accumulator capacity is sufficient for each ADS valve to provide two actuations aqainst 31.5 psig (70Ã of 45 psig) drywell pressure (see PSAR Subsection 5.2.2.4. 1 and response to Question 211.67) . (b) Lonq-Term ADS Operability of 100 Days The safety related nitrogen storaqe system contains adequate qas in storaqe (N -bottles could be replaced periodically to provide capacity for at least 100 days operation of the ADS. Justification for meeting these criteria is given below. Short-Term ADS Desian Basis Short-term is defined for this discussion as the time required to depressurize the reactor to the residual heat removal (RHR) shutdown cooling pressure permissive setpoint, stabilize the reactor water level and place the reactor in the shutdown cooling mode. Each ADS accumulator is presently sized to provide two ADS e safety/relief valve design pressure. (S/RV) actuations at 70Ã of drywell This is'quivalent to six actuations of the ADS S/RVs at atmospheric pressure in the drywell. The ADS valves are designed to operate at 70% of drywell design pressure because that is the maximum pressure for which rapid reactor depressurization throuqh the ADS valves is required. (qreater drywell pressures are associated only with the short duration primary system blowdown in the drywell immediately following a large pipe break). Por large breaks which result in higher drywell pressure, sufficient reactor depressurization occurs due to the break to preclude the need for ADS., One ADS actuation at 70% of drywell design pressure is sufficient to depressurize the reactor and allow inventory makeup by the low pressure ZCC systems. However, for conservatism, the ADS accumulators are sized to allow two ADS actuations at 70% of =-drywell design pressure. This design provides sufficient nitrogen to the ADS valves to permit depressurization until the RHR shutdown cooling mode can be initiated. Preoperational testinq of the ADS valves at 705 of design drywell pressure is not practical because pressurizing the drywell during the ADS it valve would reguire testing. Thus, an equivalent number of valve actuations at SSES-FSAR atmospheric pressure is normally included in the ADS system test specification. Long-Term ADS Design Basis The basis for the long-term ADS requirement is derived from the long-term cooling acceptance criterion (Criterion 5) of 10CFR50.46. Criterion 5 states: <<Long-Term cooling. After any calculated successful initial operation of the ECCS, the cal-culated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core. This criterion requires that either ADS be operable in conjunction with the low pressure ECCS pumps or that RHR shutdown cooling and water makeup capability he operable, to ensure long-term core cooling. The primary purpose of lonq-term ADS is to keep the reactor pressure low enough so that low pressure ECCS systems can be used to keep the core cooled. The ADS is not required after the decay heat is low enough so the vessel will not be pressurized above the shutoff head of the low pressure ZCCS pumpso The duration for which the ADS must be available is dependent on factors such as the power of the reactor at the time of the LOCA, break- size and location, available injection systems, and availability of RHR shutdown cooling. The long-term ADS design requirement is 100 days. This is based on a judgment of the time required to make any necessary repairs to the RHR shutdown cooling system or ADS, thus ensuring the core would be kept cool. Based on the 10CFR50 requirement, a long-term depressurization capability is provided by supplying nitroqen to the ADS accumulators using a safety grade system. The safety related nitrogen storage (N bottles) system contains adequate gas in storage for 30 days after a postulated DBA.. However, these nitrogen bottles could be replaced periodically by bringing portable N -bottles to provide long-term operation of the ADS. (At Susquehanna, these bottles are located in an area that is accessible following a loss-of-coolant accident.) From the above discussion, PPSL concludes that the Susquehanna design of ADS pneumatic supply system meets the intent of NUREG-0737, Item XI.K.3.28. 18 1-111 SS ES-PS AR 18 1 61 REVISED SHALL-BREAK LOSS OF COOLANT ACCIDENT HZTHODS /II.K 3 30$
- 18. 1.61.1 Statement of requirement The analysis methods used by nuclear steam supply system vendors and/or fuel suppliers for small-break loss-of-coolant accident (LOCA) analysis for compliance vith Appendix K to 10 CPR Part 50 should be revised, documented and submitted for NRC approval.
The revisions should account for comparisons with experimental data, including data from the LOFT test and Semiscale Test facilities. The Bulletins and Orders Task Force identified a number of concerns regardinq the adequacy of certain features of small-break LOCA models, particularly the need to confirm specific model features (e.g., condensation heat transfer rates) against applicable experimental data. Thes'e concerns, as they applied to each. light-water. reactor (LWR) vendor's models, vere documented in the task force also concluded that, in light of the TMI-2 accident, additional systems verification of the small-break LOCA model. as required by II.O of Appendix K to 10 CFR 50 was needed. This included providing experimental verification of the various mades of single-phase and twa-phase natural circulation predicted to occur in each vendor's reactor during small-break LOCAs. Based on the cumulative staff requirements for additional small-break LOCA model verification, including both integral system and separate effects- verification, the staff considered model revision as the appropriate method for reflecting any potential upgrading of the analysis methods. The purpose of the verification vas to provide the necessary assurance that the small-break LOCA models vere acceptable to calculate the behavior and consequences of small primary system breaks. The staff believes that this assurance can alternatively be provided, as appropriate,,by additional. justification of the acceptability of present small-break: LOCA models vith regard to specific staff concerns and recent test data. Such justification could supplement or supersede the need for model revision. The specific-staff concerns regarding small-break LOCA models are pzavided in-the analysis sections of the,BSO Task Force reports for each LWR'endor, (NUREG-0635, -0565, -0626, -0611, and 0623). These concerns should be reviewed in total'y each holder of an approved, emergency core cooling system model and addressed in the evaluation as appropriate.. The recent tests include the entire Semiscale small-break test ~ series and LOFT Tests (L3-1) and L3-2). The staff believes that SSES-FSAR the present small-break LOCA models can be both qualitatively and quantitatively assessed against these tests. Other sepazate eff ects tests (e. q., ORNL core uncovery tests) and future tests, as appropriate, should also be factored into this assessment. Based on the precedinq information, a detailed outline of the proposed program to address this issue should be submitted. In particular, this submittal should identif y (1) which ar eas of the models, if any, the licensee intends to upgrade, {2) which areas the licensee intends to address by further justification of acceptability, (3) test data to be used as part of the overall verification/upgrade effort, and (4) the estimated schedule for performing the necessary work and submitting this information for staff review and approval. Licensees shall submit an outline of a program for model justification/revision by November 15, 1980. Licensees shall submit additional information for model justification and/or revised analysis model for staff approval by January 1, 1982. Licensees shall submit their plant-specific analyses using the revised models by January 1, 1983 or one year after any model revisions are approved. Applicants shall submit appropriate information in accordance with the licensing review schedule. 18.1.61.2 Entezozetation ' None required. F 18.1.61.3 Statement Of Resoonse PPGL considers that the reactor vendor, General Electric, is the most appropriate party to work with the staff in resolving staff concerns with small break LOCA models foz BHRs. Accordingly, the staff should direct their questions regarding the scope and schedule for this requirement to General Electric (attn. R. H. Buchholz, Manager, BMR Systems Licensing) . Copies of correspondence on this item should be sent to PPGL so that we may remain cognizant of the progress of the program to resolve the staff's concerns on this requirement. 18 1 62 PLANT-SPECIFIC CALCULATIONS TO SHOW COMPLIANCE METH 10CFR PART-50 46 gEE K 3 3~1
- 18. l. 62. 1 Statement of Requirement Plant-specific calculations using NRC-approved models for small-break loss-of-coolant accidents (LOCAs) as described in item 18 1-113
SSES-'FSAR II.K.3.30 to show compliance with 10 CFR 50.46 should be submitted for NRC approval by all licensees. 18.1.62.2 Interpretation None required. 18.1.62.3 Statement of Response Plant specific calculations will be performed, following if required, NRC approval of LOCA model revisions required by item XI .K'.3.30.(see Subsection 18.1..61)., 18 1.63 EVALUATION OP ANTICIPATED TRANSIENTS WITH SINGLE 'PAILURE TO VERIFY NO FUEL CL'ADDING FAILURE /II K 3 44) 18,1,,63.1 Statement of Peguirement Por anticipated transients combined with the worst single failure an assuming proper operator actions,. licensees should demonstrate that the core remains covered, or provide analysis to show that no. significant fuel damage results from core uncovery. Transients which, result in a stuck-open relief valve should be included in this category.. All applicants for operatinq license should submit documentation four months prior to the expected issuance of the staff safety evalaution report for an operating license or four months prior to the listed implementation date, whichever is later. 18,1 63,2 Inte ryr et at io'n None required., 18.1.63~3'-Statement of-gesponse-The BWR Owners'roup has prepared a generic response to this requirement. The report was transmitted to D. G. Eisenhut by a letter from D.. B. Waters on- December 29, 1980. This response contains an evaluation of analyses performed to demonstrate the core remains covered or no significant fuel damage occurs from an 18 1-11 4 SSES-FSAR anticipated transient with a single failure. PPGL has reviewed this response and finds it is applicable to Susquehanna SES. The report concludes that the core remains covered for all evaluated combinations of anticipated transients and single failures. 18 1 64 EVALUATION OF DEPRESSURIZATZON WITH OTHER THAN THE AUTOMATIC DEPRESSURIZATION SYSTEM QZI K 3.45}
- 18. 1.64.1 Statement of geguirement Analyses to support depressurization modes other than full actuation of, the automatic depressurization system (ADS) fe.g.,
early blowdown with one or two safety relief valves) should be provided.Slower, depressurization. would reduce the possibility of exceeding vessel,'integrity limits by rapid cooldown. All appli.'cants for operating license should submit documentation four months prior to the expected issuance of the staff safety evaluation report for an operating license or four months prior to the listed implementation date, whichever is later.
- 18. 1. 64. 2 Interpretation None reguired.
18.1.64.3 Statement of Response The BQR Owners'roup submitted a generic response to this requirement., This response was transmitted by letter to D. G. Eisenhut from D..B., Haters on December 29, 1980.. PPSL has reviewed this response and find it applicable to Susguehanna SES. The report concludes that no improvement can be gained by a slower depressurization and actually could be. detrimental to core cooling. Therefore no additional action is necessary in response to this requirement.. 18 1 65- HICHELSON-CONCERNS /II.K 3 46} l8~.65.1 Statement- of g~euirement A number of concerns related to decay heat removal following a very small break LOCA and other related items were questioned by 18 1-115 SSES-FSAR Mr. C. Michelson of the Tennessee Valley Authority. These concerns were identified for PHRs. GE. was requested to evaluate these concerns as they apply to BQRs and to assess the importance of natural circulation during a small-break LOCA in BWRs. 18.1.65.2 Interpretation None required. 18.1.65~3 Statement of Response The General Electric Company- has responded to the guestions posed by Mr. Michelson.. This response was sent by letter from R- H. Buchholz to D. F.. Ross on February 21, 1980. These responses are applicable to Susquehanna SES and no further response is necessary. 18 $ 66 EMERGENCY PREPAREDNESS-SHORT TERM /III A 1 lg No requirement stated in NUREG 0737. Refer to Subsection 18.,2.38 which contains the response to the requirement in NUREG 0694. 't '8 1 67 -UPGRADE EMERGENCY SUPPORT FACILITIES gIIT A 1 2) 18.1.67 1 Statement of Q~euirement A detailed statement of the requirement can be found in NUREG-0696. The implementation schedule was announced in Generic Letter 81-10 on February 18, 1981.. This schedule is as follows: Design information for emergency response facilities should be provided in connection with the operating license review process. These facilities shall be operational'y October 1, 1982 or prior to fuel load, whichever is later. Interim facilities, as described in NUREG-0694 .shall be provided by fuel load.
- 18. 1 67. 2 Inter2retat ion None required.
18.-1-116 SSES-FSAR 18.1.67.3 Statement of Resoonse The proposed method of respondinq to this requirement was submitted by a letter to B. J. Younqblood from N. W. Curtis on April 2, 1981 (PLA-704) . Details on the emergency response facilities are presented in the Emergency Plan. 18 1.68- EMERGENCY- PREPAREDNESS-LONG TERM ZIII A.21 18 1 68 1 Statement of Reauirement Each. nuclear facility shall upgrade its emergency plans to provide- reasonable assurance that adequate protective measures . can and will be taken in the event of a radiological emergency. Specific criteria to meet this requirement is delineated in NUREG-0654 (FEHA-REP-1), >>Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparation in Support of Nuclear Power Revision 1: NUREG-0696, >>Functional Criteria for Plants.>>'UREG-0654, Emergency Response Facilities; >> and the amendments to 10 CFR Part
- 50. and Appendix E to 10 CFR Part 50 regardinq emergency preparedness, provide more detailed criteria for emergency plans, design, and functional- criteria for emergency response facilities and establishes firm dates for submission of upgraded emergency plans for installation of prompt notification systems. These revised criteria and rules supersede previous Commission guidance for the upgradinq of emergency preparedness at nuclear power facilities.
Requirements of the new emergency-preparedness rules under paraqraphs 50.47 and 50.54 and the revised Appendix E to Part 50 taken together with NUREG-0654 Revision 1 and NUREG-0696, when approved for issuance, go beyond the previous reguizements for meteorological programs. To provide a realistic time frame for implementation, a staged schedule has been established with compensating actions provided for interim measures. Specific milestones have been developed and are presented below. Milestones are numbered and taqqed with.-the following code; a-date h-activist c-minimum acceptance criteria. They are as follows: (1) a. Fuel load. b.. Submittal of radiological emergency response plans. 0 SSES-FSAR C>> A description of the plan to include elements of NUREG-0654, Revision 1, Appendix 2. (2) a. Fuel load. b., Submittal of implementing procedures. c>> methods, systems, and equipment to assess and monitor actual or potential offsite consequences of a radiological emergency condition shall be provided. (3) a. Fuel load. b., Tmplementation of radiological emergency response plans.
- c. Four elements of Appendix 2 to NUREG-0654 with the exception of the Class B model of element 3, or Alternative to item (3) reguiring compensating actions:
A meteorological measurements program which is consistent with the existing technic'al specifications as the the baseline or an element 1 program and/or element 2, system of Appendix 2 to NURZG-0654, or two independent element 2 systems shall provide the basic meteorological parameters (wind direction and speed and an indicator or atmospheric stability) on display in the control room. An operable dose calculational methodology (DCH) shall be in use in the control room and at appropriate emergency response facilities. The followinq compensatinq actions shall be taken by the licensee for this alternative: (i) Zf only element 1 or element 2 is in use: o The licensee (the person who will be responsible for making offsite dose projections) shall check communications with the cognizant National Reather Service (NMS) first order station andbasisN%S toforecasting ensure that station on a monthly routine meteorological observations and forecasts can be accessed. o The licensee shall calibrate the meteorological measurements program at a frequency no less than guarterly and identify a readily available source of meteorological data (characteristic of site conditions) to 18 1-118 SSES-PSAR which they can gain access during calibration periods. 0 During conditions of measurements system unavailability, an alternate source of meteorological data which is characteristic of site conditions shall be identified to which the licensee can gain access. 0 The licensee shall maintain a site inspection schedule for evaluation of the meteorological measurements program at a frequency no less than weekly. o It shall be a reportable occurence if the meteorological data unavailability exceeds the goals outline in Proposed Revision 1 to Regulatory Guide 1.23 on a quarterly basis. (ii) The portion of the DCM relating to the transport and'iffusion of gaseous effluents shall be consistent with the characteristics of the Class A model outlined in element 3 of Appendix 2 to NUREG-0654 (iii) Direct telephone access to the individual responsible for making offsite dose projections (Appendix E to 10 CPR Part 50{IV) (A) (4)) shall be available to the NRC in the event of a radiological emerqency. Procedures for establishing contact and identif ication of contact, individuals shall be provided as part of the implementing procedures. This alternative shall not be exercised after July 1, 1982. Purther,. by July 1, 1981, a functional description of the upgraded programs (four elements) and schedule for installation and full operational capability shall be provided (see milestones 4 and 5). (4) a. March 1, -1982. b Installation of Emergency Response Pacility hardware and software.
- c. Pour elements of Appendix 2 to NUREG-0654, with exception of the Class B model of element 3.
(5) a. July 1, 1982.. b Pull operational capability of milestone 18 1-119 SSES-PSAR Ce The Class A model (designed to he used out to the plume exposure EPZ) may be used in lieu of Class B 'model out to the ingestion EPZ. Compensating actions to be taken for extending the application of the Class A model out to the. ingestion EPZ include access to supplemental information (meso and synoptic scale) to apply judgment regarding intermediate and long-range transport estimates. The distribution of meteorological information by the licensee should be as described in Table 18.1-13 by July 1, 1982. {6) a. July 1, 1982 or at the time of the completion of milestone 5, whichever is sooner.
- b. Mandatory review of the DCM by the licensee.
co Any DCM in use should be reviewed to ensure consistency with the operational Class A model. Thus-, actions recommended during the initial phases of a radiological emergency would be consistent with those after the TSC and EOP are activated. (7) a. September 1, 1982. b., Description of the Class B model provided to the NRC N Co Documentation of the technical bases and justification for selection of the type Class B model by the licensee with a discussion of the site-specific attributes., (8) a. June 1, 1983.
- b. Full operational capability of the Class B model.
- c. Class B model of element 3 of Appendix 2 to NUREG-0654, Revision 1 Applicants for an operating license shall meet at least milestones 1, 2, and 3 prior to the issuance of an operating license. Subsequent milestones shall be met hy the same dates indicated for operating reactors. Por the alternative to milestone 3 the meteorological measurements program shall he consistent with the NUREG-75/087, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,>> Secton 2.3.3 program as the baseline or element 1 and/or element 2 systems.
18 1-120 SSES-FS AR
- 18. 1.68.2 Interoretation None required.
$ 8.1.68.3 Statement of Response Responses to these requirements are incorporated into the Emergency Plan. 18 1,69 INTEGRITY OF SYSTEMS OUTSIDE CONTAINMENT LIKELY TO CONTAIN RADIOACTIVE NATERIAL /III.D.l,l) 18.1.69.1 Statement of Requirement Applicants shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as-low-as-practical levels. This program shall include the following: {1) Immediate leak reduction. (a) Implement all practical leak reduction measures for all'ystems .that could carry radioactive fluid outside of containment. {b) Measure actual leakage rates with system in operation and report them to the NRC. (2) Continuing Leak Reduction Establish and implement a program of preventive maintenance to reduce leakage to as-low-as-practical .levels. This program shall include periodic integrated leak tests at intervals not to exceed each refueling cycle. This requirement shall be implemented prior to issuance of a full-power license. Applicants shall provide a summary description, together with initial leak-test results, of their program to reduce leakage from systems outside containment that would or could contain primary coolant or other highly radioactive fluids or gases during or following a serious transient or accident. Applicants shall submit this information at least four months prior to fuel load SSES-FSAR
- 18. 1. 69. 2 Interpretation None required.
18.1.69.3 Statement of Response Program summary description: 1.1 The followinq systems will be leak tested {the frequency is indicated in { ) after each each item): A Residual Heat Removal (18 months) B . Reactor Core Isolation Cooling C Core Spray D High Pressure Core Injection E'o. Scram Discharge p Reactor Water Clean-up~ G, Standby Gas Treatment Hi Containment Air Monitors I % Post Accident Sampling Initial leak-test, results will be available when the first. measurements are made,, prior to completion of the startup test program. + NOTE: . The RWCU system wiU. not have significant post-accident radioactivity because the suction is isolated by containment isolation signals (refer to Table 18.1-10).. However, this system may conceivably be used in some post-accident scenarios, and will therefore be leak tested. 1 2 The followinq systems contain radioactive material but are excluded from our program {justification for exclusion follows each item): A. Main Steam identified by NEDO-24782 as not to be regarded as containing highly radioactive fluid following an accident. B Peed water same justification as A. C Main Steam Line Drain this system is isolated following a LOCA. D Reactor Water Sample this system vill not be used following an accident, a separate R 18 .1-122 SSES-FS AR post-accident sampling station is being developed in response to item II.B.3. E. Recirculation Pump Seal Rater (from CRD pumps) 'ines are protected by check valves and an excess flow check valves. F. Floor 6 Equipment Drains this sytem isolated followinq a LOCA and vill not be used folloving an accident. G. Suppression Pool Clean-up 6 Drain same justification as F. 1-3 Method for obtaining actual leak rates A ~ Water - leakage will be collected in a graduated measuring device and timed to determine GPM leak rate. Implementing procedures vill establish criteria for initiation of leak rate quantification. B. Steam an estimate of the size of the leak vill be made (i.e.. equivalent pipe diameter steam. flow) . Flovrate will be determined using. standard Hand. book data. This vill be converted to a GPM flovrate using the specific volume of the steam at the given conditions. The two gaseous systems are tested as follovs: A. Standby Gas Treatment System This system is subject to filter efficiency testing in accordance. vith the Technical Specifications which includes "DOP" and refrigerant injection. 1 B. Containment Air Monitors These are tested while the system is under normal running conditions by checking each mechanical joint v ith liguid soap.. Consideration vas given to the Standby Gas system regarding the. incident at North Anna Unit 1 in 1979. The standby gas piping and duct work from to the filters are gas tight and do not the'ontainment include any pressure relief devices which would allov gases to escape to the Reactor Building.. The piping is rated at 150 psig and the duct work is HVM-GS-G (High Velocity Medium, Pressure Galvanized Steel Gas tight) . 18 1-123 SSES-PSAR In liqht of the above, the actions stated in 1.1.G and
- 2. A have resulted.
- 4. Technical Specifications references this program which includes an acceptance crite ia of 5 GPM %otal leakage rate for the systems listed in 1.1 with the exception ofo Standby Gas Treatment which is limited to the acceptance criteria stated in Technical Specifications Subsection 4.6.5.3 and B. The containment air monitors which has an acceptance criteria of zero leakage as determined by a liquid soap test.
$ 8 $ 70 . INPLANT IODINE,RADIATION MONITORING /III.D 3 3i 18.1.70.1 Statement of Reauirement Each licensee shall provide equipment and associated training and procedures for accurately determining the airborne iodine concentration in areas within the facility where plant personnel may be present during an accident., Effective monitoring of increasinq iodine levels in the buildings under accident conditions must include the use of portable instruments using sample media that will collect iodine selectively over xenon (e.g., silver zeolite) for the following reasons: The physical size of the auxiliary and/or fuel handling building precludes locating stationary monitoring instrumentation at all areas where airborne iodine concentration, data might be required. (2) Unanticipated.isolated "hot spots" may occur in locations where no stationary monitoring instrumentation is located. T (3) Unexpectedly high- background radiation levels near stationary monitoring instrumentation after an accident may interfere with filter radiation readings. (4) The time required to retrieve samples after an accident may result in high personnel exposures if these filters are located in high-dose-rate areas.
- 18. 1-124
SSES-ZSAR After January 1, 1981, each applicant and licensee shall have the capability to remove the sampling cartridge to a low-background, low-contamination area for further analysis. Normally, counting rooms in auxiliary buildinqs will not have sufficiently low backqrounds for such analyses following an accident. In the low backqround area, the sample should first be purged of any entrapped noble gases using nitrogen gas or clean air free of noble gases. The licensee shall have the capability to measure accurately the iodine concentrations present on these samples under accident conditions. There should be sufficient samplers to sample all vital areas. 18.1.70 2 Interpretation PPSL is in basic agreement with the technical discussion as outlined in this requirement. It should be noted that Susquehanna SES is a BMR and does not possess an auxiliary ~ building. Consequently,, it is premature to suggest that our counting facilities within the control structure vill be inadequate to effectively count air samples. Additionally, purqinq of the air sample cartridqes may not be necessary if ef'fective collection media is used for radioiodine air sampling. an 18..1.70.3 Statement of Response PPSL vill meet the requirements defined in this item., To summarize the program, three (3) particulate and gaseous continuous air monitoring systems are provided for air sampling plant areas where personnel may be present during accident conditions. The systems are cart mounted for ease of relocation. Grap samples are obtained using the equipment specified in Subsection 12.5.2.6.3. During accident conditions silver zeolite cartridges vill be used for radioiodine analysis in conjunction with tvo (2) Eberline stabilized assay meters {SAM-2) or equivalent; C Air samples are evaluated as specif ied in Subsection 12 5.3.5. 5. In addition to initial training provided for Health Physics personnel, periodic drills are conducted in accordance with the Susguehanna Emergency Plan Section 8.1.2 (See Amendment 25 of Operatinq License Application). Analysis of iodine cartridges vill be performed in a low background, low contamination area. During accident conditions, preliminary analysis will be performed by onsite radiation monitoring teams in the counting room, if accessible using a SAM-2.. Pinal analysis vill be .Performed in the emergency off-site facility vhere appropriate sensitivity can be achieved. prior to 18 1-125 SS ES-P SAR analysis, cartridges will be purged using station service air or bottled nitrogen, if necessary to reduce noble gas interference. 18 1 71 CONTROL ROOM HABITABILITy REQUIREMENTS ZIII D 3.43 18.1.71.1 Statement of Requirement Licensees shall assure that control room operators will be adequately protected against the effects of accidental release of toxic and radioactive gases and that the nuclear power plant can be safely operated or shut down under design basis accident conditions (Criterion 19, >>Control Room," of Appendix A, "General 'esiqn Criteria for Nuclear Power. Plants,>> to 10 CPR Part 50). All licensees must make a submittal to the NRC regardless of whether or not they met the criteria of the Standard Review Plans (SRP) sections listed below. The new clarification specifies that licensees that meet the criteria of the SRPs should provide the basis by referencing past submittals to the NRC and/or providinq new or additional information to supplement past submittals. 18.1.71.1.1 Requirements for Licensees that Meet Criteria All licensees with control ~ ~ rooms that meet the criteria of the following sections of the Standard Review Plan: ~ ~ 2.2.1-2.2.2 . Identification of Potential Hazards in Site Vicinity
- 2. 2. 3 Evaluation of Potential Accidents; 6 4 Habitability Systems shall report their findings regarding the specific SRP sections as explained below. The following documents should be used for quidance:
(a) Requlatory Guide 1.78, >>Assumptions for Evaluating the Habitability of Regulatory Power Plant Control Room During a Postulated, Hazardous Chemical Release"; (b) Regulatory Guide 1.95, "protection of Nuclear Power Plant Control Room Operators Against an Accident Chlorine Release>>; and, (c) K. G.,Murphy and K.,M.,Campe, >>Nuclear Power Plant Control Room Ventilation System, Design for Meeting General Design Criterion 19,>> 13th AEC Air Cleaning Conference, August 1974.,
- 18. 1-126
SSES-FSAR Licensees shall submit the results of their findings as well as the basis for those findings by January 1, 1981. In providing the basis for the habitability finding, licensees may reference their past submittals. Licensees should, however, ensure that these submittals reflect the current facility design and that the information zeguestel in Attachment 1 of NUREG 0737 is provided. 18.1.71.1.2 Requirements for Licensees that Do Not Neet Criteria A11 licensees with control rooms that do not meet the criteria of the above-listed references, Standard Review Plans, Regulatory Guides, and other references shall perform'he evaluations and identify appropriate modifications, as discussed below.. Each licensee submittal shall include the results of the analyses of control room concentrations from postulated accidental release of toxic gases and control room operator radiation exposures from airborne radioactive material and direct radiation resulting from design-basis accidents The toxic gas accident analysis should be performed for all potential hazardous chemical releases occurring either on the site or within 5 miles cf the plant-site boundary.. Regulatory Guile 1.78 lists the chemicals most commonly encountered in the evaluation of control room habitability but is not all inclusive.. The design-basis-accident (DBA) radiation source term should be for the loss-of-coolant accident LOCA containment leakage and, engineered safety feature (ESP) leakage contribution outside containment as described in Appendix A and B of Standard Review Plan Chapter.. 15 6.5. In addition, boiling-water reactor (BMR) facility evaluations should add any leakage from the main steam isoaltion valves (MSIV) (i. e., valve-stem leakage, valve seat leakage, main steam isolation valve leakage control system release) to the containment leakage and ESP leakage following a LOCA.,, This should not be construed as altering the staff recommendations in Section D, of Regulatory Guide 1.96 (Rev. 2) regarding MSIV leakage-control systems. Other. DBAs should be reviewed to determine whether they might constitute a more-severe control-room hazard than the LOCA. In addition to the accident-analysis results, which should either identify the possible need for control-room modifications or provide assurance that the habitability systems will operate under all postulated conditions to permit the control-room operators to remain in the control room to take appropriate actions required by General Design Criterion 19, the licensee should submit sufficient information needed for an independent evaluation of the adequacy of the habitability systems. Attachment 1 of NUREG 0737, item III.D.3.4 lists the information that should be provided along with the licensee~s evaluation. 0 18 1-127 SSES-FSAR 18.1.71,1.3 Documentation and Implementation Applicants for operatinq licenses shall submit their responses prior to issuance of a full-power license. Modifications needed for compliance with the control-zoom habitability requirements specified in this letter should be identified, and a schedule f or completion of the modifications should be provided. Implementation of such modifications should be started without awaitinq the results of the staff review. Additional needed modifications, will be if any, identified by the staff during its review specified to. licensees.
- 18. l. 71. 2 In terre tation None requi red.
18.1.71.3 Statement of Resoonse The control room habitability system design includes protection of the control room from radioactive and toxic gases. Subsection 6.Q provides a description of this system and compliance to habitability requirements. Potential hazards from nearby ~ ~ facilities are discussed and evaluated in Subsection 2. 2. ~ ~ ~ ~ References for the information. required for the NRC control room ~ ~ habitability evaluation aze provided in Table 18.1-17. ~ ~ 18 1-72 = REFERENCES
- 18. 1-1 Letter, D. G. Eisenhut (NRC) to S. T. Rogers (BRR Owners'roup), regarding Emergency Procedure Guidelines,. October 21, 1980 18 1-2 U.S. Nuclear Requlatory Commission, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations" USNRC Report NUREG-0578, July 1979, Recommendation 2.1.6b.
U.S. Nuclear Regulatory Commission, >>NRC Action Plan Developed as a Result of the TMI-2 Accident,>> USNRC-0660, Vols. 1 and 2, May 1980,Section II.B.2. 18 1-4 Letter from D G. Eisenhut (NRC) to All Licensees of Operating Plants and Applicants for Operating Licenses and Holders of Construction Permits,
Subject:
18 1-128
SSES-PS AR Preliminary'larification of TMI Action Plan Requirements, dated September 5, 1980.
U.S. Nuclear Requlatory Commission, "Clarification of TMI Action Plan Reguirements,>> USNRC Report NUREG-0737, November, 1980, Item II. B.2.
U.S. Nuclear Regulatory Commission, IE Bulletin No.79-01B, >>Environmental Qualification of Class IE Equipment", January 14, 1980.
U.S. Nuclear Regulatory Commission, "Interim Staff Position on Environmental Qualification Report NUREG-0588, December 1979.
USNRC Standard Review Plan 6 4, >>Habitability Systems",
Revision l.,
US'NRC Regulatory Guide 1.3, >>Assumptions Used for Evaluatinq the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Hater Reactors",
Revision 2, June 1974.
USNRC Regulatory Guide 1.7, >>Control of Combustible Gas Concentrations in Containment Pollowing a Loss-of-Coolant Accident," Revision 2, November 1978.
USNRC Regulatory Guide 1.89, >>Qualification of Class IE Equipment for Nuclear Power Plants,>> November 1974..
Code of Pederal Regulations, 10CPR Part 50, Appendix A, GDC 19, Revised as of January 1, 1980.
C. Michael Lederer, et al., Table of Isotopes~ Lawrence Radiation Laboratory, University of California, March 1968 D. S. Duncan and A. B. Spear, GRACE I An IBM 704-709 Program. Design for Computing. Gamma Rag Attenuation and.
Heating in Reactor Shields~ Atomics International,
{June 1959) .
D. S, Duncan and A. B. Spear, for Computing-Gamma Rag GRACE II An Attenuation IBM 709 and Heating Pgogram-in Cylindrical and Spherical Gecmetries, Atomics International, November 1959.
Memorandum of Telephone Conversation, S. Pord of LIS to N.,Anderson of NRC's Lessons Learned Task Porce,
Subject:
TMI Reguirements at SHNPP, April 9, 1980.
18 1-129
SSES-PSAR 18.1-17 USNRC Regional Meeting Minutes, Region I, Sub ject: TMI Review Requirements at SHNPP, April 9, 1980.
18.l-l8 USNRC Regional Meeting Minutes, Region IV and V,
Subject:
TMI Reviev Reguirements, 9/26/79.
18 1-130
SSES-FSAR TABLE 18.1-1 INTERIH RE UIRED SHIFT STAFFING One Unit, Two Units Two Units Three Units Control Control Two Control Two Control
'ne One Operating Status Room Room Rooms Rooms One Unit Operating+ 1 SS (SRO) 1 SS (SRO) 1 SS (SRO) 1 SS (SRO) 1 SRO. 1 SRO 1 SRO 1 SRO 2 RO 3 RO 3 RO 4 RO 2 AO 3 AO 3 AO 4 AO Two Units Operating+ NA 1 SS (SRO) 1 SS (SRO) 1 SS (SRO) 1 SRO 2 SRO 2 SRO ) Only 1 SRO S 4 ROs required 3 RO 4 RO 5 RO ) if both units are operated 3 AO 4 AO ) from one control room 5 AO All Units Operating+ NA 1 SS (SRO) 1 SS (SRO) 1 SS (SRO) 1 SRO 2 SRO 2 SRO 3 RO 4 RO 5 RO 3 AO 4 AO 5 AO All Units Shut Down 1 SS (SRO) 1 SS (SRO) 1 SS (SRO) 1 SS (SRO)
~
1 RO 2 RO 2 RO 3 RO 1 AO 3 AO 3 AO 5 AO SS - shift supervisor RO - licensed reactor operator SRO - licensed senior reactor operator AO - auxiliary operator NOTE: ~
(1) In order to operate or supervise the operation of more than one unit, an operator (SRO or RO) must hold an appropriate, current license for each such unit.
(2) In addition to the staffing requirements indicated in the table, a licensed senior operator will be required to directly supervise any core alteration activity.
(3) See item I.A.l.l for shift technical advisor requirements.
- Nodes 1 through 3.
Rev. 27, 10/81
SSES-FSAR TABLE 18.1-2 INITIAL CORE ISOTOPIC INVENTORY
~Zsoto e Curies ~Isoto e Curies ~isoto e Curies I131 8.66+7 Y-"-93 1.82+8 TE-129 2.38+7 I"-132 1.29+8 Ywww94 1.61+8 TE131M 1.31+7 I-"133 1.99+8 Y"""95 1.84+8 TE-131 7.74+7 I134 2.32+8 ZR-95 1.84+8 TE" 132 1.29+8 I 135 1.82+8 ZR""97 2.86+8 TE133M 1.40+8 I 136 9.22+7 NB-95M 3.81+6 TE-133'E-134 8.93+7 BR 83 1.52+7'.74+7 NB"-95 1.91+6 2.05+8 BR"-84 NB-97M. 1.78+8 CS"137 1.13+7 BR 85 3.84+7 NB""97 1.85+8 CS-138 1.90+8 KR-83M 1.55+7 MO 99 l. 8&8 CS-139 1.93+8 KR-85M 3.87+7 MO-101 1.49+8 CS-140 1.76+8 KR 85 1.31+6 MO-102 1.19+8 CS.-142 9.22+7 KR 87 7.44+7 MO-105 f 2.05+7 BA137M 1.75+8 KR"-88 1.04+8 TC-99M 1.63+8 BA"139 1.87+8 KR 89 1.37+8 TC-101 1.49+8 BA-140 1.87+8 XE133M 5,.0& 6 TC-102. 1.23+8 BA"141 1.87+8 XE"133 1.98+8 TC-105 2.65+7 BA-142 1.71+8 XE135M 5.36+7 RU"103 8.93+7 LA-140 1.87+8 XE-135 1.87+8 RU-105, 2.68+7 LA-141. 1.90+8 XE-137'E"138 1.79+8: RU-106 9. 84+7 LA"143 1.74+8 1.76+8 RU>>107 5.65+6 LA-142 1.74+8 SE'81 4.17+6 RH1 03M 8.93+7 CE-141 1.90+8 SE-83M 8.63+6 RH105M 5.62+6 CE-143 1.75+8 SE 83 6.55+6 RH-105 2.68+7 CE-144 1.45+8 SB 84 2.92+7 RH-106 1.16+7 CE-145 1.15+8 RB 88 1.07+8 RH-107 5.65+6 CE-146 8.81+7 RB 89 1.42+8 SN-127 3.27+6 PR-143 1.75+8 RB 90 1.72+8 SN-126 1.75+1 PR-144 1.49+8 RB 91 1.62+& SN"128 1.10+7 PR-145 1.15+8 RB 92. 1.31+8 SN"130 5.95+7 PR-146 9.14+7 SR""89 1.42+8 SB"127 3.87+6 ND-147 6.70+7 SR"-90 1.14+7 SB-128 1.64+7 ND-149 3. 24+7 SR 91 1.73+8 SB-129 2.20+7 ND-151 1. 19+7 SR 92 1.58+8 SB"130 5.95+7 PM-147 3.45+7 SR 93 1.67+8 SB-131, 8.03+7 PM-149 3.24+7 SR 94 1.28+8 SB-132 9.97+7 PM"151 1.25+7 Ywmm90 1.72+8 SB-133 1.01+8 SM-151 2.70+5 Y' 91M 1.01+8 1.04+6" SM-153 4.70+6 TE127M'E"127 1
'wmm9 1.70+8 3.87+6 Ymmm92 1.76+8 TE129M 1.04+7 (1) Based on 1000 reactor operating days at 3440 MMt. Reference GE Internal Report Document, "Summary of Fission Yield for U-235, U-238, and PU-239,"
published by Meek and Rider, June, 1977.
8.66+7 = 8 66 10 Rev. 27, 10/81
SSES-CESAR TABLE 18.1-3 RADIATION ZONE CLASSIFICATION Radiation Maximum Zone Dose Rate I 15 mR/hr II 100 mR/hr III 500 mR/hr IV c 5 R/hr 50 R/hr VI 500 R/hr VII < 5000 R/hr VIII > 5000 R/hr Notes:
- 1. Based on maximum contact dose rate for zones containing radiation sources.
- 2. Based on maximum field dose rate for'ones with radiation fields caused by sources located outside the area.
Rev. 27, 10/81
SSES-FSAR TABLE 18.1-4 VITAL AREAS Radiation TID State of Occu an Frere ~Sbol Zone (rem)
Continuous Main Control Room 18.1-5 A. 1 0.24 Technical Support Center 18.1-5 A.2 1.6 Operations Support Center 18.1-5 A.3 1.6 North Gate House (ASCC) 18.1-1 A.4 0.1 Security Control Center 18.1-1 A.5 0.1 Emergency Operations 18.1-1 A.6 0.1 F-ility(')
Post-Accident Sampling
- 1) Sample Station 18.1-5 B.l 1 (2)
- 2) Chemistry Lab 18.1"3 B.2 0.1(3)
- 3) Plant Vent Sample 18.1-8 B.3 0.1(4)
Station (1) Information regarding the EOF may be found in Appendix I of the SSES Emergency Plan.
(2) TID is determined per event, that is, for a one time, one man task initially performed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> post-accident and lasting 30 minutes.
(3) TID is determined per event, that is for a one time one man task initially performed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> post-accident.
(4) TID is determined per event. Duration is thirty (30) minutes for filter removal and transport. Task initially performed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> post-accident.
Rev. 27, 10/81
SSES"FSAR TABLE 18.1-5 PRINCIPLE DOSE RATE CONTRIBUTORS IN PIANT AREAS Structure Area Dominant S stem (Source)
Core Spray Core Spray (C)
Sump. Room RHR Cooling Mode (B)
Elev. 670'-0". Wetwell Drywell. (A) to 683'-0" RHR RHR Cooling Mode (B)
Access RCIC (C)
Corridor Truck Port RHR Cooling Mode (B)
Railroad Port 'HR Cooling Mode (B)
Other Areas Core Spray>(C), RCIC (C),
HPCZ (C)
Elev. 683'-0" Drywell Drywell (A) o 719P 1 Equip. Areas RHR Cooling Mode (B)
Equip. Removal RHR Cooling Mode (B)
Areas Core Spray Piping Area Core Spray (C)
Elev. 719'-1" Drywell Drywell (A) to 749'"1" Main Steam Core Spray (C)
Tunnel NE Equipment RHR Spray. Mode (C)
Airlock SW Equipment Core Spray (C)
Airlock 'HR South Switch Spray Mode (C)
Gear Room CRD Hatch RHR Spray Mode (C)
Elev. 749'-1" Drywell Drywell (A) to 770'>>1" Penetration Core SprayR(C), RRR Spray Rooms and Mode (C) other Areas
- 2) Control Building All Areas Core Spray (C)
Elev. 656'-0" to 806'-0" Elev. 806'-0" All Areas. Standby Gas Treatment to 818,'-0" Systems:.
Rev. 27, 10/Sl
SSES-FSAR TABLE 18.1-5 (page 2 of 2)
NOTES:
(1) At one hour post-accident, the steam source (D) will dominate area radiation levels. Following reactor steam activity depletion, radiation levels will be due to contained source (C).
(2) Radiation levels are based on system/source proximity, however, in this case each system noted contains source (C). Therefore, radiation levels may be determined as a function of time by referring to the same curve on figures 18.1-9 and 18.1-10 for source (C).
Rev. 27, 10/81
SSES-F SAR TABLE 18. 1-6 CHEMICAL AND RADIOCHEMICAL ANALYTICAL CAP ABILITIES RZgUIRED OF OFF-SITE LABORATORY Liquid Samples The laboratory must be capable of handling up to 10 ml of undiluted reactor coolant/supp ession pool vater vith activity levels to 3.0 curies per ml. The laboratory vill. be required upto perform the following analyses within the range and accuracy indicated.
Radioisotopic Analysis
- a. 'amma-Ray Spectroscopy Identify and quantify with accuracy of + 20 all ~
isotopes which have gamma-ray peaks in the spectrum from 50 to 5000 keV with a net peak area of greater than 5",o of .he total spectrum counts within a + 5 times full width at half maximum (FTHM) band about the centroid. of the peak. The spectrometer system must be capable of analyzing samples vith total concent",ation of gamma-ray
,emitting isotopes as. lov as 0.01 microcuries per ml..
- b. BetaActi.vity Gross beta and quanti"ative determination of Sr-89 Sr-90 (up to 10 days permitted for complet'on of this analysis) .
Co Alpha Activity Gross alpha count and relative alpha activi+ies by alpha spectroscopy.
Uranium and Plutonium Identify and perform semiquantitative analyses for these elements.
- 2. Cond ucti vity Range:. 0.1 to 10,000 micromhos pe
+ 20,o cm.'ccuracy:
- 3. pH Range: 1, to. 13.
Accuracy:. + 0..3'H units, Rev. 27, 10/81
SSES-PSALM TABLE 18.1-6 ( a e 2of2)
Chloride Hange: greater than 50 ppb Accuracy: + 10%
+ 50 if if ppb greasier less than 500 ppb than 500 ppb
- 5. Boron Range: O.l to 10,000 ppm Accuracy: + 50%
+ 20%
if if less than 1 ppm greater than 1 ppm and less than 100 porn
+ 5 %'f g"eater than 100 ppm B. Gas Samples Gas samples vill be obtained from the following sources:
drywell, vetvell, secondary containment, and dissolved. gases from liquid samples. The laboratory must be able to handle 15 ml gas samples vi h activity levels up to 0.06 cu=ies per ml., The laboratory vill be required to perfo m the following analyses vithin the ange and accuracy indicated:
Padioisotopic analysis by gamma-ray spectroscopy.
See Section A.l. a for, requirements.
2%i Eleme ntal Analysis Identify and'uantify by volume 7~ the following: hydrogen, oxygen, nitrogen and krypton (spike added to dissolved gas samples) . The analysis sensitivity should be sufficient to detect any of these constituents at the 0.1%"by volume level. At the, 0.1% level the analysis should be accurate to
+ 20%. At concentrations above 0.5Ã the analysis should be accurate to vithin + 55.
C Particulate and Iodine Cartridge Samples'he laboratory must be able to hand1e and perform gamma-ray spectroscopic analysis on oarticulat, silver zeolite, and charcoal filter cartridges. The maximum activity anticipated for any of these cartridges is O.l curies. The analysis should be able to identify and quantify with an accuracy of + 50% all isotopes which have gamma-ray peaks in the spectrum from 100 to 4000 keV with a net peak area of greater than 5S of the total spectrum counts within a + 5 times PWHM band about the centroid of he peak.
Rev. 27, 10/81
SSES-FSAR TABLE 18.1-7 DOSE RATES FROM PASS AND TRANSPORT CASKSC<ii Thickness of Shielding in Source Lead Inches 0 Dose Rate ft 3 ftin mph 8 ft, Liquid Sampler<<~ 5300 310 55 Gas samplerc~) 8800 220 110 Small Volume Cask 1600 (0.1 ml sa.mple)
Large Volume Cask 5 1/2 260 (10 ml sam pie)
Gas Cask- 1 L/8 5500 (10.7 cc. sampLe)
C13 Based on source term 1 hou" folloving shutdovn C23 Dose rates from the sample panels vill be present for only a fev minutes vhil the sample is floving.
Rev. 27, 10/81
SSES-PSAR TABLE 18. 1-8 TRAINING CRITERIA FOR MITIGATING CORE DAMAGE A program is to be developed to insur that all operating personnel are trained in the use of installed plant systems to control or mitigate an accident in which the core is severely damaged. The training program should include the following topics.
A. Incore Instrumentation Use of fixed or movable incore detectors to det "mine extent of core damage and geometry changes.
- 2. Methods for calling up (p inting) incore data from -he plant. computer.
t B Vital Instrumentation Instrumentation response in an accident environment; failure sequence (time to failure, method of failure);
indication reliability (ac.ual vs. indicated level) .
- 2. Alternative methods of measuring floss, pressures levels, and temperatures.
a., 'etermination of reactor pressure vessel level all le vel transmit ters fail. if
- b. Determination of other reactor coolant system parameters has failed.
if the primary method of measurement primary Chemistry Expected chemistry resul. s with severe core damage; consequences of transferring small quantities o. 1iquid outside containment; importance of using leak tight systems.
20 Expected isotopic breakdown fo core damage; foz clad damage.
3 ~ Corrosion effects of extended immersion in primary eater; time to failure.
D Radia tion Monitor~in
- l. Response of Process and Area Monitors to severe damages; behavior of detectors shen saturated; method for detecting radiation readings by direct measurement at detector output (overzanged detector): xpec ed Rev. 27, 10/81
SSES-FS AB TABLE 18.1-8 ( a e 2of2) accuracy of detectors at differenct locate,ons; use of detectors to determine exten of core damage.
- 2. methods of. determining dose rate inside containment from measurements taken outside containment.
E ~ Gas Generation Methods of hydrogen generation during an accident; other sources of gas (Xe, Kr); techniques for venting or disposal of non-condensibles.
2 Hydrogen flammability and explosive limit; sources of oxygen in containment or reactor coolant system.
Rev. 27"," 10/81
SS ES-PS AR TABLE 18. 1-9 NITIGATXNG CORE DAMAGE COURSE OUTLINE
- 1. Causes and Thresholds of Core Damaqe A. Poser Transients B. Normal Operatinq Conditions C. Core Oncovery
- 2. Recoqnition of Core Damaqe A.. By instruments Read in the Control Room By Chemical Analysis C BY Containment Conditions
- 3. Procedures Related to Nitiqatinq Core Damaqe.
Rev. 32, 12/82
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-I 1OSOA AC 2 30 KIDS HO
~ ~
151156 CR2840 NAINT (11) IY210
-IF 1 226 AC 2 30 KIDS NO ISIISA (I1) I F210 E 11. I 205A -IF026A AC G 96 Ks IA (15) ~ ~
~ ~
ISIISA CR2040 ISIIIA (KLRC 18216 102IS HE 13. I-205A -IFOIIA AI B,C 22 $ 13A NO CR2940 SRH BUS A I 13. I-204A -IF02IA AC G se 0 I A) (15) ~ ~
15128A ISIIIA CR2040 (KLRC 10216 IS210 OUS A NE 13. I-204A -IFOI IA AI a.c 22 $ 13A) ND CR2940 SRH BUS A E Ii. I-226A -IFOOTA AC LFRH 26 Ksit) ~ ~
I 5 IOTA CRIaio SRA 10219 BUS A E IS. 2-2466 -IFOSSA PSV
-ISIOSA PSY
-IFI03A RH E-153 SH 21 E I I <0 ~ t COO I 15113A CR2840 SRH 10231
~~
E IS. I-2460 -IF0550 PSV H/A'15)
-151060 PSV
-IF1030 ~ RH f-153 SH Si El 1-60 CSOI 151130 CR2940 SRH 10241
-I F091 PSV
~
I~ . I-203A -IF004A RH I El 1-68 ~
CSOI 1510(A CR2940 (KLRO 10210
- 10. I-2030 -IF004C RN 19 f1 l-80 ~ ~
151040 10231 II. I-390 EII-IFOI68 AC E-153 SH 95 Ell-66 (KSIO) 25 CSOI 151168 CR2940 (KLRC Ia226 HAIN IIS'YDC BUS 0 f 12. X-138 -I FO I 50 AC G6160862 16 (Ksl0) NO
~ ~
Cool 1511501 CR2940 SRA 10220
- F0500 AC I 3 HO
)6 151500 CR2940 NA}NT
- F1220 AC I 30 (Dos) HO ISISOO f I-I2 NN )8
-:BN 12.
F128 A)
PSV N:u);) oil:ill fH)l carol mkloamll III. LC II/II
PAGE 4 OF 9-IASLE 10.1-10 (CONT IHUEO)
YALYE AUTONATIC ELECTRICAL GE ELEH 6 AUTO OPEH VALVE SIRIUS LOCATION HAND SSITCH POSER SOURCES (3)
PAID. E OR BASIS PENA IR. ACTUA- AC'IUATION $ CHEUATIC ACIUA1ING DH 1$ 0 INDICATION VALVE OTHER
$ 16118 NE I NO. VALVE Mo. T ION SIGNALS (2) 0 I ACRAH RELAY RE$ El SOURCE LOCAL PNL CONTROL RH No. TYPE (14) 80100 CONTROL LOCIC REIIARXG 8-1$ 1 E ~ X-2050 -IF0290 AC 0 E-153 Ql 12 Ell<0 (KSIS) (15) 2$ C?01 aol 151260 CR1040 (KLRC 10229 2$ VDC RHR NE IFOIIO Al O.C 92 ($ 13$ ) NO 151118 CR?940 GRA NA IXI 18220 10220 SUS 8 (CONT. ) E'E 13. X-2040 -IF026$ AC 0 12 (KSIO) (15) C?OI 151260 CR2940 (KLRC 10226 IFOIIB Al S.C 02 (8138) NO 151118 CR?940 SRA HA INT 10220 10220 E 14. X-2260 -IF0018 AC Lflol 03 (604S) N/A 6201 1$ 1018 CR2940 SRA 18210 E 16. X-203D -IF0040 IUI IS El 1<6 151040 CR1040 (KLRO) 10241 IIXIH X-2030 -110040 RU E1 1<6 C?OI 1510481 CR?040 (KLRD 10226 IIAIN 8-152 X-16A E21-IFOOSA AC CAT E21-3$ (KI3A) .(16) 1$ CSOI 1520$ A CR2940 SRA IS?ll 12$ VDC CORE BUS 0
$ PRAY -I f0066 RH E?I-3$ 152066 CR2040 (PS) IY? le
-IF 031A RU E21-3$
9~
152066 Ill IC020 BUSS X-168 -IF0058 AC CST E?1>>3$ (KI?8) (16) ~ ~
151050 CR1940 SRA 1022
-IF0068 IUI E21-3$ 152068 CR?040 (PS) (11 IY226 0
-IF0310 RH E?I-3$ 152060 ~ ~
(IT I C026 8 HE IO. X-201 A -IFOISA AC 0 E?1-3$ (KIOA) i (IS) 1$ 215A CR194D SRA 10231 A HE 10. X-2010 -If015$ AC 0 E21-3$ KIOB) (16) 152150
~ ~ ~ ~
10241 8 E IS. X-201A -'IF03 I A AC LFCS 621-35 H006A N/A 152316 10216 A E 10. X-2018 -I 1 0310 AC LFCS E21-3$ 80060 N/A 152318
~ ~ ~ ~
10226 8 E 20. X-206A -IFOOI A E21-3$ 1520IA CR1040 (KLRD 10216 E 20. X-2060 -IF OSIS RH E21-35 ~ ~
1$ ?DIS 18226 N-155 21. X-ll $ 41-11 002 AI E-152 SH IS E41-60 ($ 44) VES 1$ aol 15502 CR2940 (KLRO 10231 12S YDC HPC I HAIN OUS A
\~ ~ ~
8
-ITOD3 AI ($ 34) 15503 CR2840 (KLRO 10264 BUG HAIN
-IF ISO Al L (636) ~~
15521 CR2940 HAINT (11) BUG A
-I F048 CK
~ ~
- 21. X-244 -IF OI0 Al fb E-I52 Ql 11 E41-60 (KIOA) H/A ~ ~
15510 CR2940 $ RA ID254 BUG A
-IFOT 5 Al FS 11 (KISS) H/A 1551 5 CR2940 SRA ID264 BUG 8
IASLf 10.1-10 (CONTINUED) PAGE 5 Ol I YALVf AUIOVA11C ELECTRICAL Cf EL(V I AU(0 OPEN VALVE SIRIUS LOCATIOH HAND SV I 'ICH POTER SOURCES (3)
PAID. E OR SASI 5 PERE'IR. ACIUA- ACIUAIIOH SCHLVAII C AC'IUAIINC ON ISO IUD I CAT I OH YALTE OTHER SVS'IEV HE ( I) NO. VALVE NO. llON SIGNALS (2) 0 I ACRAII RELAT RESET SOURCE LOCAL PNL CONIROI. RV NO. 'ITPE (14) VOIOR CONTROI. lOCI C REVARKS V-155 f 21. 1-2 ID -11040 25 CI0 I 15560 CR2040 (KLRO) IDAI I HPC I SAINT (CORI. ) -1104$ CK
- 23. 1-209 11042 AI E IS2 SH Il 541<0 (134) NO
~ ~ ~~
15542 CR2$ 40 SRA ID 244 8US 8 S. 1-$ 0 11004 AC I (Kl) ( IS) 15504 CR2$ 40 SRA ID244 8US 8 V-IS) 2l. 1-26 HV 15) II AI 1 (II),R E-111 SH II 02I I)I (KI3) HD 25 CI0 I I5111 E-SOAb VOV 1122$ RPS 0 (11)
NE
",R I II I COVIN I AINUS CONTROL E 25. 1 COA HT HT SV 15113 15114 15140A T,R T
V (12) 3 (1$ 4)
IKI)
CS)IOA 15113 151 I 4 IS)ISA I'12 1122C ID624 RPS RPS 8
~ ~
(13)
~
ST 15142A
~
(KI) ~ ~
15'1424 ~ ~
25: IQOA SV 15150A ~ ~
Kl)) ~ ~
15140A ~~
ST 151524 ~ ~
KI)) ~ ~
1514)A ~ ~ ~ ~ ~ ~
Hf 2l. 1-202 HY 15103 T (II).R I Kll) 15103 ~ ~ * ~ ~
(11)
HY 15104 I -,R 113) 15104 11226 RPS 8 HY 15105 I (II),R Il KI3) 15105 11228 E 25. 3-2)IA SV 151IOA T (12) Kl)) C0220A 15140A ID424 0 (13)
SV 1518)A I ~ ~
~ ~
~ ~
Kl)) ~~
15142A I~
~ ~
E 25. 3-2) IA SV 151)IA '1 KI) ~ ~
15140A ~ ~
SV 15134A '1 ~ ~
Kl) ~ ~
15142K ~ ~ ~ ~
f 25. 1-ICC ST 151400 '1 ~ ~
Kll Cell le 151440 I 1216 10414 RPS A
~~
SV 151420 '1 ~ ~
~ ~
Kll ~ ~
I Sll20
- 25. 1-1 OC SY IS)508 Kl4 15140$
SY IS)S20
~~
Kll ~ ~
I 51428 ~ ~ ~ ~ ~ ~ ~
- 25. 1 IOC ST 151140 V ".R Kll ~ ~
Is)Ale SV I51140 ~ ~
Kll ~ ~
151420 SY 15141 T -.e 143 I 5141 Irlle RPS I Nf 24. 1-25 HY 15122 V -,R Kll ~ ~
15122 IT2II HY 15123 T -,R Kl) 1512) 1122$ RPS 8 HV 15121 I kl3 ~ ~
I III I ~~
11224 HY 15124 V 8 2l 1-201A (Kll) ~~
I NE HY HV I5125 15124 15'121 T,R 8 (Kl))
15125 15124 ~~
I T21$
11226 RPS RPS 8 NV T~ 0 HV 15123 T,R )DC l4
(I3)
SY 151310
~~
'1 ~ ~
R (III) I51400 11216 IDII l
~~ ~ ~
SV 15131
~ ~
8 (Kl)) 15131 11224
~ ~
0 f 25. 1-233 ST 151100
~ ~ ~ ~
(Kll) C0220$ 15140l 11216
~~
J
PACE 6 Of 0 TASLE I0.1-10 (CON'TINUED)
PS I 0. E OR . OASIS PENElN ACTUA- ACTUAllON SCNENAT I C AC'IUATINO ON ISO INOICA1ION VALVE OTHE R SVSTEII NE (I) NO. VALVE KO. 1 I ON 0 IONA LS (2 0 IACRAR RELAV RESET SOURCE LOCAL PNL CONTROL RN NO. TVPE (14) ROTOR CONTROL LOGIC REQARXS
~ ~ ~~ ~ ~
SV ISTS20 'V ~ ~ IS1410 IVXIS (XSI)
E 2S.'-ASS SV 1$ 11SA
~ ~ ~ ~
~~
~
C02204 IS140A 11226 10614 RPS 0
~ ~
SV IS114A (XSS) C0220A 10142A 11220 IDSI4 RPS 0
PAGE 1 OF 9 1ASLE IS.I"10 (CONT)NUEO)
VALVE AUTNAtlC ELECTRICAL CE ELfN 6 AUTO OPEN VALVE STATUS LOCA)IN Sel POIER SOURCES (3)
PSIO f OR SASIS rENETR. ACTUA- ACTUAT IN SCHEXATI C ACTUATIHQ ON ISO 'HOICATIN NAHD TCH YALYE OIHER 616th NE Na.
'6.
'ALYE NO. TIN 6 I GNALS 2 0 I ACRAH RE(AT RESEt SOUR LOCAL PNL CNtROL IOI NO. TYPE (14) NOIOR COHIROL LUCIO SENARKS N-1ST . Nf X-143 NT 15166 Al (6$ 4I NO 15 CS0 I 1516$ f-30AC NNW 18231 RPS A COITNT HY 1516$ 1516$ E-30AC NOHAI 10214 RPS 8 ATNOS CON1ROL (CNT.)
H-IS I NE 21. X-)2$ HT lileeAI Al E-159 SH 6 811;131. 859) NO 15 C209 Geol 16104A I E-SOAB HN I'1236 I't 219 RPS A
~ ~ ~~ ~ ~
LIOUID HY 1610162 6 (K60) I6104A2 RPS 8
~ ~ ~ ~ ~ ~ ~ ~
RA01AStE Hf 21. ~ X-11A HY ISIISAI 4 (X5$ ) Ie I I 6 A I RPS A
~ ~ ~
COHIROL HY 16116A2 9 (X60) ~
161 l662 11226 RPS 8 H-le) HE 1-$ 58 HY 1$ 19162 Al T f-216 SH II $ 21-131 ( K44) NO 26
~ ~
CSSI 1419 1 6 CR2$ 40 HAINT
~~
IY234 I0634 RPS A (11) ( I 4)
REACTOR HY 1419282 ~~
23 ( K$ 3) 141928 11246 RPS 8
~~
SLOS
~~
HY 1$ 19IAI .Y II K44) ~ ~ ~ ~
19)9 I A 1123& 10634 RPS A 8
CHILLEO ~~
HT 1419281 ~ ~
1 13 K43) 14192$ 11246 RPS
'SATER ~ ~
I-54 HY 1914182 Y II X$ 3) ~ ~ ~~ 141418 11246 1123S 10634 RPS RPS 8
A
~ ~
HY 1$ 1$ 2A2 Y 23 644I 1$ 1$ 2A
~ ~
X-53 HT Ielelel Y II ~ I ~ ~
I 41418 IY246 10634 RPS 8
~~ ~~ 23 K44) 14142A 11136 RPS A HY 1$ 192AI ~ ~
~ ~ ~~
X-SSS HY 1419182 '1 - 11 643) ~ ~ I $ 1 SIR IY246 I0434 RPS 8
~ ~ ~ ~ RPS
~ ~
HY 14192A2 Y 23 K94) 191 92A 11136 A
' ~ ~
HT 1$ 19181 ~~
1 II Kes 1$ )SIB 11246 11236 RPS RPS 8
A
~ ~ 14192A
~ ~
1-54 HY 141$ 2AI HY 14)SLA2 Y
Y 23 11 K$ 4)
K$ 4)
~ ~ t ~
141$ 'IA
~ ~
11236 10834 RPS A HY 1414242
~~
2 K43 ~~ 141428 11246 RPS 8
~ ~
~~
X-55 HY Ie)SIAI II 8$ 4 g ~
14141 K IT236 IO634 RPS A 8
~~ '0 HT 1$ 1$ 2$ 1 T 2 K43 ~~
le)etc IT246 RPS Aov, 3li 7/41 E
~ -M
SS ES-FSAR TABLE 18. 1-10/Pa<re 8 ~~g)
REGIA RKS
~>> Essential or non-essential classification basis codes are described in Table 18.1-11.
(2) Automatic actuation siqnal codes are described in Table
- 18. 1-12 (3) Rhere the control power source is left blank, the control power source is the same as the valve motor power source.
(4) E32-1F001B automatic actuation signal is dependent upon action of QSTV~s, time,, RPV pressure. The valve is normally closed and interlocked when RPV pressure is greater than 35 psiq., The valve cannot be opened unless the inboard t1SIV is closed. Xnformation presented is representative of that for main steam lines B, C and D.
(5) Automatic signal code UA prevents operation of condenser low vacuum bypass.
(6) Reactor recirculation'ystem sample line valves B31-1F019 and 1F020 receive hiqh radiation signals for isolation but since the line does not provide an open path from the containment to the environs, the radiation isolation signal may be considered a diverse signal in accordance with Standard Review Plan 6.2.4. This judgement is based on our definition of an open path as a direct, untreated path to the outside environment.
Hand Switch Sos. are from the PSlD rather than referenced Schematic Diagram.
(8) Automati" actuation signals for E11-1F015A and B:. codes UB and Z are isolation siqnals; codes G and T are initiation signals.
Automatic actuation signals for E11-17050A and B, and 1F122A, B= code Z is an isolation signal; no initiation siqnals.
(10) Either valve opening '(or closing) vill energize a common open (close) status light HS-11314 controls both valves.
Typical for HV-11345 and HV-11346.
Closes on "LOCA" signal but can be reopened af ter 60 minutes. Valves can be administratively reopened if the high drywell pressure is due to plant heat up or loss of drywe11 cooler.
(12) Closes on "LOCA" signal but can be reopened after 10 min utes
SSES-PSAR T~Bgg 18 1-10 /Page 9 gf 2).
(13), Power source 17226
~ is for control. 1D614 or 1D624 is for status indication lights.
~ ~
(14) Switch types:
E-3 OAC Cutler Hammer two button operater momentary with NI = mechanical interlock'utler E-30AB Hammer, momentary CR2940 GE: KI. = keylock
'Q = k ey removable in open position. ~
RC = k ey re mo vab Le in close position RN'= key removable in normal position SRN = spring return to Normal SRA = spring return to Auto 5AINT = maintained contacts PB = pushbutton (15) Initiation reset will automatically handswitch is in open position.
reopen valve if valve (16) Initiation reset will not automatically reopen valve.
t17) Pneumatic actuated valve.
(18) Power sources 1Y236 and 1Y246 are AC control power, and 1D6 3 4 is DC'ontrol power ..
Rev. 31, 7/82
SSES-FSAR TABLE 18.1-11 E SSENTIALQNON-ESS ENTT AL PENE RATION CL ASSIF'ICATION BA S ZS Closed Cooling 'Water Non-essential since used during normal operation only fo- reacto= recirculation pump cooling, reactor water cleanup and oth r system components.
Not. required for design basis accident situation.
(2) Containment Intstrument Gas - Essential "o suppor= safe y eauip ment.
(3) Inst"ument Gas Non-essen ial support to non-safety elated equipment-, and for testing of saf ty rela+ed equipmen (4) Hain Steam Line 'and ZSZV Leakag~ Con+rol System Non
~
essential for shutdown.
Feedwater Line Not essen+ial. for shutdown but desirabl=
for makeup water to vessel. Po"=ion between reac"or vessel and outermost con ainmen+ isolation valve is essen" ial =or HPCI, and RCIC injection.
Reactor Core I'soltion Cooling "=ssential for coro cooling following isolation from turbin condenser and feedwater makeup Beac+or Rater Cleanup No+ essen".ial. du ing or immediately following an accident., maybe important in long +erm recovery operations.
Reac" or Rater Sampling - Not essen" ial for safe shu" down.
Post-accident samples will b ".aken utilizing the pos+-
accident sampling system developers in response to item II.B. 3.
S'tandby Liquid Control - Essen ial as backup to CRD system.
(10) Pesidual Heat Femoval (RHB) Head Spray Not essential for safe shutdown.
RH> Containment/Suppression Pool Spray Essential fo pressure con+rol.
(12) RHP. Shutdown Cooling Essential to achieve cold shutdown.
(13) BHR Steam Condensing Recirc./Test Return Line No" essen+'al since not a safety function. Used during ho+
standby and pump tes+s.
(14) RHR Pump Minimum Flow Recirculation Essential for pro+ect pumps, for safety function.
Rev. 27, 10/81
SSES-PSAR TABI;E 18. 1-11 (oaee 2of2)
(15) RHR heat Exchanger Relief Valve Discharge Line Ess ntial to protect HX from overpressuriza ion for use in safety function.
(16) RHR Suppression Pool Suction Essential for vessel injection and pool cooling safety func ions.
(17) Core Spray Injection - Essen .ial safety function.
(18) Core Spray Pump Test R etu" n Lines !ton-essential. Used only during testing of pumps.
(19), Core Spray Pumps t'tin. Plow Bypass - Essentail to protec+
pumps for safety function t
(20) Core Spray Suppression Pool Suction Essential fo vessel injection safety function.
{2t) High Pressure Coolent Injection (HPCI) Turbine Steam Supply and Exhaust,-Essential to drive HPCI pump for vessel injection safetv function.
(22) HPCI'ump Min. Recirc. Essential to. protect pump for safety function.
(23) H2CZ Suppression Pool Suction "ssential for vessel injection safety function. Backup to Condensa.e S"o=age Tank supply..
(24) Containment Atmospheric Pu"ge .'ton-essential vent pa" h to Standbv Gas Treatment System. Backup to four hyd"ogen recombiners.
(25) Con+ainment Atmoshere Sampling - Essential. Not "eaui"ed for shutdown, but would be necessary or post-accid n" assessment.
(26) Suppression Pool Rater Filtration Pot essential.- Users only for periodic. cleanup of pool water.
(27) Liauid Radwaste Collection Non-essen.ial for safe shutd own.
(28) Reactor Bldg. Chilled Fa er Von-essential supply ".o recirculation pump motor coolers, drvwell coolers.
Rev. 27, 10/81
SSES-FSAR TABLE 18.1-12 ACTUATION/ISOLATION SIGNAL CODES 5 CORRESPONDING ACTUATING SWITCHES
- ISOLATION FUNCTIONS: OTHER CODES FOR INFORMATION ONLY.
A~ Reactor Vessel low water level 3 B21-N024A or B B* Reactor Vessel low water level 2 B21-N026A or B C* Main Steam line high radiation D12-K603A (typ. of 4)
D* Main Steam line high flow B21-N006A (any one of foui) B21-N007A (typical for MSL "A") B21-N008A B21-N009A E* Main Steam line leak/high temp B21-N600A (typ. of 4)
(either) B21-N603A (typ. of 4)
TSH-10100A (typ. of 4)
FA+ High drywell, pressure Reactor/ Ell<<N010A or C RCIC Steam line low pressure E51-N019A or C (both required)
FB* High drywell pressure Reactor/ Ell-N010A or HPCI Steam line low pressure or C C'41-NOOlA (both required)
G Reactor Vessel low level 1RHR, Core B21-N031A or C Spray (level 2 RCIC, HPCI), or Drywell high pressure E11-N011A or C E11A-S18A for Ell-F016A, F021A, F028A (typ. for B)
I Reactor Vessel low water level 2 B21-N031A-D (one of two twice)
J* RWCU line break/high flow G33-N044A RWCU high flow differential G33-N603A (either)
Rev. 31, 7/82
SSES-FSAR TABLE 18.1-12 (Pa e 2 of 3)
K* RCIC Steam line leak/high temp:
Equip room area high temp E51-N600B Equip room vent air high A temp E51-N601B Emer area cooler high temp E51-N602B Pipe routing area high temp E51-N603B$ D Pipe routing area high b, temp, after E51-N604B, D time delay (any one of 5), Bypass (typ. for HV-F007)
Switch B21B-S3BI RCIC steam line break/high A P E51-N018 (for HV>>F007)
(N017 for HV-F008)
Reactor/RCIC Steam line low pressure E51-N019B 5 D (typ. for HV-F007)
Turbine exhaust diaphragm high E51<<N012B 5 D pressure (typ. for HV-F007)
L* HPCI Steam line leak/high temp:
Equip room area high temp E41-N600A Equip room'vent air high A temp ~
E41-N601A Emer area cooler high temp E41-N602A Pipe routing area high temp E51-N603A$ C Pipe routing area high A temp, E51-N604A, C after time delay (any one of 5) (typ. for HV-F002)
B ass Switch B21B-S4A HPCI Steam line break high A P E41-N004 (for HV-FO 02) I (N005 for HV-F003)
HPCI Steam supply low pressure E41-N001A & C Turbine exhaust diaphragm high pressure E41-N0012A 5 C (typ. for HV-F002)
LFCS CS pump discharge low flow E21-N006A (with CS pump running)
LFHP HPCI pump discharge low flow E41-N006 (with HPCI pump running)
LFRC RCIC pump discharge low flow E51-N002 (with RCIC Pump running)
LFRH RHR pump discharge low flow Ell-N021A (with RHR Pump running)
' Rev. 31, 7/82
SSES-FSAR TABLE 18.1-12 (Pa e 3 of 3)
RHR Shutdown cooling and Head Spray line break:
Equip area ambient high temp Ell-N600A or C Equip area vent air high L temp Ell-N601A or C B ass switch B21B-S6A Cooling line high A P Ell-N019A (any of the above)
(all typical for .
HV-F009 and -F022)
P* Main Steam line low pressure B21-N015A Run Mode Onl B assed on (typ. of 4)
Start, & Hot Stdb , Refuel or.
Shutdown modes R+ High-high radiation in SGTS exhaust vent D12-K617A, B T'eactor Vessel low pressure B21-N021A-D (permissive, one of two twice)
UA Main condenser vacuum low-Bypassed when B21-N056A Condenser Bypass switch is in "BYPASS" and B21-N020A turbine stop valves not open, or Reactor (typ. of. 4)
Vessel pressure above Low Pressure Interlock setpoint UB Reactor Vessel low pressure B31-N018A W* RWCU leak detection ambient high G33-N600A, C, E temperature vent air high temp G33-N602A, C, E (any one of six)
Z* Reactor Vessel low water level 3 B21-N024A, B Drywell high pressure C72-N002A$ B (typ. for outboard )
valves)
Y* Reactor Vessel low water level 2 B21-N026A$ B Drywell high pressure C72-N002A, B (typ. for outboard )
valves)
Rev. 31, 7/82
SS ES- PS AR TABLE 18 1-13 M~ET OROLOGICAL 'I NPORM ATION HRC and Emergency Meteorological Information CR'SC EOF Response Organiza-tions Basic Met. Data X X X {NBC)
(e.g., 1.97 Parameters)
Pull Met. Data X . X (1.23 Parameters)
DCM {for Dose X X Projections)
Class A Model (to X X Plume Exposure EPZ)
Class B Model or X X Class A Model (to Ingestion EPZ) f;18.1 581$
Rev. 27, 10/81
SSES-FSAR TABLE 18.1"14 CALCULATION OF COOLANT BETA DOSE RAD/HR AT 1 HR DECAY 2 3 Col. 1" F (')
B MeV/s( ) B MeV/s per per HWt at MVZ, T= 1 hr Coolant One Hour Released Rad/Hr I-131 1.78E14 8.8&E13 6.28E4 I-132 7'.26E14 3.63E14 2.57E5 I-133 8.15E14, 4.08E14 2.89E5 I-134 1.06E15 5.30E14 3.75E5 I-135 6.67E14 3.33E14 2.36E5 Br>>&3 2.,04E13 1.02E13 7.22E3 Br-84 T.72E13 3.86E13 2.73E4 Xe-133 2.06E14 2.06E14 1.46E5 Xe-135 1.16E14 1.16E14 8.18E4 Xe-138 5-.85E13 5.85E13 4.14E4 Kr-85 2.55E12. 2.55E12 1.80E3 Kr-85 4.95E13 4.95E13 3.50E4 Kr-88 1;33E14 1.33E14. 9.40E4 Kr-&7' 3. 71E14 3.71E14 2.63E5
- 9. 71E15 l. 92E6 (a From CINDER run 8/18/80, SNUMB 2100T, Core Inventory 3 yr burn (b) F = 1.0 for noble gases and 0.5 for halogens, 3293 MWt*3600 s/h*1 . 6E-6 erg/MeV 7 p77 lp 10 (2.68E& g)*(100 erg/g/Rad)
Rev.. 27, 10/SI
TABLE 18. 1-15 CONDUCTIVITY OF PURE WATER UNDER IRRADIATION IRRADIATED CELL UNIRRADIATED CELL Temp. (O.l cm Balsbaugh) (O.l cm Beckman) oF S/cm S/cm No Irradiation 63.5 0.10 0.11 No. Flow 0.14 0.11 1.3x10 4Rad/hr Flow 65. 0 0.11 0. 12.
No Flow 1.4 5'low 6.6xlO 4Rad/hr Flow No F,low Flow 65.5
'W W 65.8
- 0. 13 2.2
- 0. 18 0.
0.20 14 1.3x10 Rad/hr No Flow 2.2, 5'.6x10 5
Rad/hr Flow 66.0 0.31 0.37 No Flow 2.1 6.6xlO 5 Rad/hr Flow 64.5 0.65 0.64 No Flow 1.8 to 1.4+
9.8x10 Rad/hr Flow 65.5 0.66 No Flow Still dropping slowly after 5 min Very steady value Rev. 27, 10/Sl
TABLE 18.1-16 CONDUCTIVITY OF 10 m Cl SOLUTION NaC1 Unirradiated Cell Irradiated Cell (O.l cm Beckman) (0.1 cm Balsbaugh) s/cm s/cm Flow 27.7 32.1 No Irradiation No Flow 27.7 32.0 Flow 27.4 32.7 Flow 27.9 33.6 9.8x10 5 Rads/hr No Flow 28;0 25.1 Flow 28.0 34.0 No Irradiation Flow 28.0 33.4 No Flow 28.1 32.2 Rev. 27, 10/81
SSES TABLE 18.1-17 INFORMATION RE UIRED FOR CONTROL ROOM HABITABILITYEVALUATION Data Reference*
- 1. Control Room Mode of 0 eration, (i.e., pressurization and filter recirculation 6.4.2, 6.4.3 for radiological accident isolation or chlorine release).
- 2. Control Room Characteristics:
- a. air volume control room; 6.4.2
- b. control room emergency zone (control room critical files, kitchen, 6.4.2 washroom, computer room, etc.);
- c. control room ventilation system schematic with normal and emergency air Figs. 9.4-1 6 9.4-2 flow rates;
- d. infiltration leakage rate;. 6.4.2
- e. HEPA filter and charcoal adsorber efficiencies; 6.4.3, 6.5.1
- f. closest distance between containment and air intake; Fig. 6.4-2
- g. layout of control room air, intakes, containment building, and Fig. 6.4-2 chlorine or other chemical storage facility with dimensions;
- h. . control room shielding including radiation streaming from penetrations 12.3.2 doors, ducts, stairways, etc.;
automatic isolation capability-damper closing time, damper leakage Table 6.4-1 and area; chlorine detectors or toxic gas (local or remote); 9.4.1
- k. self-contained breathing apparatus availability (number); E.Plan, App. D
- 1. bottled air supply (hours supply); E.Plan, App. D
TABLE 18. l'-17 (Continued)
Data Reference*
l
- m. emergency food and potable water supply (how many days .and how many E.Plan, App. D people);
- n. control room personnel capacity (normal and emergency); and, 6.4.1
- o. potassium iodide drug supply. E.Plan, App. D .
- 3. On-site stora e of chlorine and other hazardous chemicals:
- a. total "amount and size of container; and, Fig. 6.4-2
- b. closest distance from control room air intake. Fig. 6.4-2
- 4. Off-site~manufacturin , stora e or trans ortation facilities of hazardous chemicals:
- a. identify facilities within a five-mile radius; Table 2.1-17
- b. distance .from control room; Table 2.1-17
- c. quantity-of hazardous chemicals in one container; and, 55 gallon drums
- d. frequency of hazardous chemical transportation traffic (truck, PLA-694 rail, and barge).
- 5. Technical' ecifications:
- a. chlorine detection system; and, TS 3/4.3.7.8
- b. control room emergency filtration system including the capability TS 3/4.7.2 to maintain the control. room pressurization at 1/8 inch water gage, verification of isolation by test signals and damper closure time and filter testing requirements.
+All references are in the FSAR unless otherwise noted.
SSES-FSAR TABLE 18.1-18 POSSIBLE ICC DETECTION DEVICES Name of Device Reference Number Source Range Monitor 1 Intermediate Range Monitor 2 Local Power Range Monitor 3 Traveling Incore Probe 4 Gamma-Neutron Reaction Detector 5 Gamma Attenuation 6 Gamma Void, Meter 7 Neutron Modulation Void Meter 8 Core Reactivity Detector 9 Fuel Plenum Tracer 10 Primary System Activity Meter 11 Incore Thermocouples 12 Heater Junction Thermocouples 13 Gamma Thermometers 14 Control Rod Drive Thermocouples 15 Sight Glass 16 Cerenkov"Light Detector 17 Wave Guide 18 Vessel Weight 19 Vessel Vibrations 20 Floats 21 Conductivity Probe 22 Capacitance Probe 23 Sonic Reflection 24 Loose Parts Monitor 25 Microwave Probe 26 Mass Balance 27 Differential Expansion Integal Anem ometer 28 Delta-P Bubbler 29 Self-Powered Neutron Detector 30 Resistance Temperature Detectors 31 Steam Dome Thermocouples 32 Liquid Level and Void Fraction Dete ctor 33
200 100
- 40 Permissible Regian 20 10 8
6 Nitric Acid I
~ Sodium Hydraxide 4
a 6
I
~ 8
~ I
.6
~ 2
~ 08
~ 06
~ 04 Impossible Regleh
~ 02
.01 6 10 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 AND 2 FINAL SAFETY ANALYSIS REPORT I
-SPECIFIC CONDUCTANCE AND pH OF AQUEOUS SOLUTIONS AT 25 C FIIlURE 18 . l-ll
OTHER CLOSE ON HIGH DIFF. PRESSURE ISOLATION SIGNALS
/
/ / / TIME DELAY RELAY CONTACT
//
/
TIME DELAY RELAY AUTO ISOLATION RELAY L
L -aFUSE 8 D m I M
A r C X
I rr ll M
H g8 r rk f 2 m
' M < U I
~
o O
r O
INCHES RELATIVE TO INCHES ABOVE VESSEL 0 160 STEAN INSTRUMENT 0 HAIN STEAN LINES 140 DRYER 658,5" 120 RXPRESSXXXXPSIG WATER LEVEL 100
'HUTDOWN RNG XXX IN 80 MATER LEVEL TREND UPSET RANGE XXX IN 60 AR RAN GE 40 AR RAh GE AR RAN GE 20 STEAN WIDE RANGE XXXX IN 0
SEPARATOR 527,5"
-20
-40 CORE SPRAY INLET 484,5"
-60
-80 FLOWS-CO C RHR LP A XXXXX GPM D -100 C
th W r
~ 5h 'll C25 -140 RCIC XXX GPN m p CO MQ ~~m 15 10 5 160 UPPER SHROUD- 377.5" L
Wh O
rZr
<Um t"INUTES TOP OF ACT IVF FUEL 366.3" Bsoo CO m-- 2 o CO th O
R
NORMALWATER LEVEL 0
%2 0 CQRB UNCQVBRY BEGINS PFAK CLAOOING TEMPERATURE AT MEI.TING POINT 10 lU 12 0 ~14' ug CC 0o 18 t
TOP OF ACTIVE FUEL O
a 18 1
20'OP OF JET PUMPS
~ 28 BOTTOM OF ACTIVE FUEL
'40 0 30 40 50 80 70 80 90 100 110 TIME AFTER SCRAM. IMINUTES).
SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 AND 2 FINAL SAFETY ANALYSIS REPORT R
DOWNCOMER 'PATER LEVEL HISTORY FIGuRE >8 ~ >->4
MELTING POINT PEAlC CLADDINGTEMPERATURES 3000'LA'DOING I T ENTIRE CORE 0 AVERAGE POWER BUNDLE Cl LOWEST POWER BUNDLE 3200 C
C W
I 0
z 1B00 5
1600 800.'
600i
~4
'12' 2 3 4 5 B 7 'B. 10 11 WATER LEVEL IFEET ABOVE BAPI SUSQUEHANNA STEAM ELECTRIC STATION
. UNITS 1 AND 2 FINALSAFETY ANALYSIS REPORT WATER LEVEL AS AN INDICATOR OF CORE OVERHEATING FIGURE 18 ~ 1-15
PEAK CLAOOING TEMPERATURES.
0 AVERAGE POWER BUNDLE 0 HIGHEST POWER BUNOLE
+< WATER
~
LOWEST POWER BUNDLE LEVEL 1'100 FEET~
Oe C
C i000 0Z' Or' 4 PRET 700 8 FEET~
0 14 20 30 40 50 60 70 80 90 100 110 120, TIME AFTER SCRAIA (minuteel SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 AND 2 FINALSAFETY ANAlYSIS REPORT CLADDING TEMPERATURE SENSITIVITY TO CORE UNCOVERY TIME FIOURE
SSES-FSAR 18.2 ~
RESPONSE TO HEQUIHENENTS-IN NUHEG 0694 NUREG-0694 supersedes NUREG 0578. The clazifications given in the Vassallo letter on Novembe'r 9, 1979 were used in the development of applicable responses.
$ 8 2.$ - SHIFT TECHNICAL ADVISOH~I- A. 1. lg Requirement superseded by NUREG 0737. Refer to Subsection 18.1.1 for response.
$ 8~2 2 - SH1PT SUPE/ VISOR ADMINISTRATIVE DUTIES QI A 1. 2) 18.$ ~.1 Statement of Heguizement Review the 'administrative duties of the shift supervisor and delegate functions that detract from or are subordinate to the management responsibility for assuring safe operation of the plant to other personnel not on duty in the control room. This reguirement shall be met before fuel load.
- 18. 2. 2. 2 Interpretation I
None required.
18.2.2.3 Statement of Response PPSI has restructured the operations organization and redefined responsibilities of shift personnel to relieve the shift supervisor of routine administrative duties.
Administrative procedure AD-QA-300, "Conduct of Operations,"
implements this policy..
The Vice President Nuclear Operations reviews and approves the Shift Supervisor's responsibilities to ensure proper" delegation of duties that detract from or are subordinate to the safe operation of the plant.
18 2-1
SSES-FSAR g8. 2. 3 ~ SHIFT MA NNI NG ~I. A. 1. 3g Requirement superseded by NUREG 0737. Refer to Subsection 18.1.3 for response.
18 2 4 IMMEDIATE UPGRADING OF OPERATOR AND SENIOR OPERATOR TRAINING AND gUALIFICATION gI. A. 2. 1)
Requirement superseded by NUREG 0737. Refer to Subsection 18.1.4 for response.
18 2 5 REVISE SCOPE AND CRITERIA FOR LICENSING
- ~ ~ EXAMINATIONS gI A. 3. 1$ ~
Requirement superseded by NUREG 0737. Refer to Subsection 18.1.6 for response.
18 2 6 EVALUATION OF ORGANIZATION AND MANAGEMENT IMPROVEMENTS
.-OF NEAR-TERM OPERATING LICENSE APPLICANTS~I
~ B 1.2~
1-8 2.6 1 Statement of Requirement The licensee organization shall comply with the findings and requirements generated in an interoffice NRC review of licensee organization and management. The review will be based on. an NRC document entitled Draft Criteria for Utility Management and Technical Competence. The first draft of this document was dated February 25, l980,. but the document is changing with use and experience in ongoing- reviews. These draft criteria address the organization, resources, training, and qualifications of plant staff, and management cboth onsite and offsite) for routine operations and the resources and activities (both onsite and offsike) for accident conditions. This requirement shall be met prior to fuel load.
- 18. 2.6.2 Interpretation None required.
18 2-2
SSES-FS AR
- 18. 2.6.3 Statement of Response A review of organization and manaqement has been completed in accordance with draft NUBEG 0731, "Guidelines for Utility Management Structure and Technical Competence." An NRC audit of the organization was conducted. March 2-6, 1981.
18 2 7 SHORT TERM ACCIDENT ANALYSIS AND PROC%DURE REVISION QI C. 1} .
Requirement superseded by NUREG 0737. Refer to Subsection 18.1.8 for response.,
18 g 8 - SHIFT RELIEF AND TURNOVER PROCEDURES gI C.2) 18..2 8.1 Statement of Requirement Revise plant procedures for shift relief and turnover to require signed checklists and logs to assure that the operating staff
{including auxiliary operators and maintenance personnel) possess adeguate knowledge of critical plant parameter status, system status, availability and alignment. This requirement shall be met prior to fuel load.
- 18. 2.8.~> I ter 2Xetation D
None required.
18.2.8.3 Statement of Response Administrative procedure AD-QA-300, <<Conduct of Operations,"
discusses operations personnel responsibilities at shift turnover. Administrative procedure AD-QA-303, "Shift Routine,"
specifically defines the shift turnover process.
18 2-3
SSES-PSAR
$8 2.9 SHIPT SUPERVISOR RESPONSIBILITIES gI. C.3$
18,2,9,1 Statement= of Requirement Issue a corporate management directive that clearly establishes the command duties of the shift supervisor and emphasizes the primary management responsibility for safe operation of the plant. Revise plant procedures to clearly define the duties, responsibilities and authority of the shift supervisor and the control room operators. This requirement shall he met prior to fuel load.
~8. 2.9. 2 Interpretation'one required.
18.2 9.3 Statement of Response The Senior Vice President Nuclear has issued a statement of policy establishing the primary responsibility of the Shift Supervisor for safe operation of the plant under all conditions and establishing authority to direct actions leading to safe operation in the Shift Supervisor. The Senior Vice President Nuclear shall re-issue this statement of policy on an annual basis.
Administrative Procedure AD-QA-300, "Conduct of Operations," sets forth the plant policy on Shift Supervisor duties.
Traininq for Shift Supervisors includes plant Administrative Procedures, and vill encompass AD-QA-300.
18,2 10 CONTROl BOOM ACCESS QI C. 81 18.2.10 1 Statement- of Reauirement Revise plant procedures to limit access to the control room to those individuals responsible for the direct operation of the plant, technical advisors, specified NRC personnel, and to establish. a clear line of authority, responsibility, and succession in the control room. This requirement shall be met prior to fuel load.
18 2-4
SSES-FSAR
- 18. 2.10.2 Inte~rretation None required.
18.2.10.3 Statement of Response Administrative procedure AD-QA-300, "Conduct of Operations,"
provides the authority and instructions,for control room access control; 18' ll PROCEDURES FOR FEEDBACK OF" OPERATING EXPERIENCE TO PLANT STAFF QI C 5+
Requirement superseded by NUREG 0737. Pefer to Subsection 18.1.12,for response.
g8 2 12 NSSS VENDOR REVIEW OF PROCEDURES gZ C 7}
$ 8,$ ,1$ ,1 Statement-of geguirement Obtain nuclear steam supply system vendor review of low-power testing procedures to further verify their adeguacy., This requirement shall be met prior to fuel load.
Obtain NSSS vendor review of power-ascension test and emergency procedures to further verify their adequacy. This requirement must be met before issuance of a full-power license.
18.2.12.2 Interoretation None required.
18.2.12 3'Statement of-Response The General Electric Company, through its site startup orqanization, has reviewed all startup tests associated with NSSS systems and will review all Emergency Operating procedures that were submitted to NRC in response to item I.C.8 (see Subsection 18.2.13) ., The startup tests encompass the low power testing and the power ascension testing phases.
18 2-5
SSES-FSAR 18 2 13 PILOT MONITORING OF SELECTED EMERGENCY PROCEDURES FOR NEAR-TERM OPERATING LICENSE APPLICANT~S I C 8$
18.2.13.1 Statement of Requirement Correct emergency procedures, as necessary, based on the NRC audit of selected plant emergency operating procedures (e.g.,
small-break LOCA, loss of feedvater, restart of engineered safety features follovinq a loss of AC power or, steam-line break) .
- 18. 2.13.2 Interpretation None required.
$ 8~.l3 3 Statement of Response Emerqency procedures based on those quidelines have been developed and are currently in trial use on the Susguehanna SES Simulator. These procedures have been reviewed by the NRC.
Final versions which incorporated NRC comments vere submitted in a letter from N. S. Curtis to B. J. Youngblood on May 15,, 1981 18 2 14 -
CONTROL ROOM DESIGN fI D. 11 Requirement superseded by NUREG 0737. Refer to Subsection 18.1.16 for response.
18 2 15- . TRAINING DURING LOQ POSER TESTING
=
gI G I) =
18 2.15.1 Statement of Requirement.
Define and commit to a special lov-pover testing program approved by NRC to be conducted at pover levels no greater than 5 percent for the purposes of providing meaningful technical information beyond that obtained in the normal startup test program and to provide supplemental training. This requirement shall be met before fuel load.
Supplement operator training by completing the special lov-power test program. Tests may be observed by other shifts or repeated 18 2-6
SS ES-FS AR on other shifts to provide training to the operators. This requirement shall be met before issuance of a full-power license.
18 2.15. 2 Interpgetation-None required.
18.2 15.3 Statement of Response The BWR Owners'roup has prepared a generic response to this requirement..'his was transmitted to D- G. Eisenhut by a letter from D. B. Waters on February 0, 1981. PPSL. concurs with this response. This generic approach outlines an extensive testing proqram desiqned to contribute to and provide for extensive training opportunities during the start-up program. The objectives of this program are to provide:
- 1. A plant that has been thoroughly tested.
2.. An operating staff that has received the maximum experience and. in-plant traininq to safely operate it.
Plant procedures that have been reviewed and revised to provide the staff with proven directions and controls.
Susquehanna's Operator Training Program has been in progress since 1977 and is completing the final phases of training at .this time. This program utilizes the Susquehanna Simulator located at the plant site and, provides the operators with extensive training prior to actual operations in the plant itself. The Simulator is also used for procedure development and check out.
The Operator Training Program that is being developed for the Preoperational and Low Power Testinq Program incorporates and builds on the extensive training already completed by the operations section. It will include the recommendation presented in the BWR Owners'roup position but goes beyond those recommendations by maximizing the use of the Susquehanna Simulator- in preparing the operators for the start-up tests to be performed.
The objective of the Operator Training Program is to provide .each operator with the maximum learning experience Curing the start-up phase. In order to achieve this objective, a comprehensive training program is being developed that utilizes the many training opportunities that are available during this period and ensures actual testing. This proqram covers the period from Preoperational/Acceptance Testing through the Power Test program 18 '-7
SSES-F SAR on Unit I. To support this amount of training the operations section which is staffed for six sections has reorganized into four sections. This reorganization provided the benefit of allowing more operators off shift to attend formal training as well as provide more operating experience for each shif t team.
Every effort is being made to keep the shifts intact and provide traininq that promotes the "Shift Team" ccncept.
The training program being developed covers the areas of activit'ies listed below but recognizes .the overlap that exists between some of the areas.
I. Preoperational/Acceptance Testing II. Cold Functional Testing III. Hot Functional Testing IV. Start-up Tests V.. Additional Testing Each area of testinq has activities that lend itself to operator training. The major ones are outlined in Table 18.2-1. The training program provides a vehicle to identify activities that have a significant benefit for training, documents this training, and ensures that all shift crews receive equal experience opportunities., The program also attempts to schedule. repeats of certain evolutions that are considered critical and cannot be routinely performed at a later. time.. The training program will identify areas of. testing/training that while not required by start-up program would have additional training benefit. This testinq/traininq could then be scheduled into the testing program as additional testing.
Finally this program will develop the basis for the Zn-Plant Drill Proqram. This comprehensive approach to testing/training more than adequately satisfies the requirements of NUREG 0737.
On June 15, 1982 (PLA-1136) PPSL submitted a station blackout Safety Analysis which demonstrates that a station blackout test unnecessarily geoparidzes the plant and the public. Therefore, no station blackout test vill be performed on Susquehanna SES Units 1 and 2. PPSL has completed additional testing per the BWROG recommendations during the Startup Test Program for Unit 1 which satisfies the intent of this requirement. As stated in Generic Letter 83-24, this additional testing along with the safety analysis vill satisfy Item I.G 1.
18 2 16 -REACTOR COOLANT SYSTEM VENTS /II B 1}
Requirement superseded by NUREG 0737. Refer to Subsection 18.1.19 for response.
18 2-8
SSES-FSAR g8. 2 17 PLANT SHIELDING /II.B 2}
Requirement superseded by NUREG 0737. Refer to Subsection 18.1.20 for response.
18 2 18 ~ POSTACCIDENT SAMPLING /II.B 3}
Requirement superseded by NUREG 0737. Refer to Subsection
- 18. 1.21 for response.
$ 8 ~ 2-19 TRAINING FOR MITIGATING CORE DAMAGE /II B 4}
Requirement superseded by NUREG 0737. Refer to Subsection 18.1.22 for response.
- 18. 2,20 RELIEF AND SAFETY VALVE TEST REQUIREMENTS /II D.1}
Requirement superseded by NUREG 0737., Refe'r to Subsection 18..1.23 for response.
18 2 2],- -- RELIEF AND SAFETY VALVE POSITION INDICATION /II D 3}
Requirement superseded by NUREG 0737. Refer to Subsection 18.1.24 for response.
$ ~82 22 -
CONTAINMENT ISOLATION DEPENDABILITY~II E.4 2}
Requirement superseded by NUREG 0737. Refer to Subsection 18.1.29 for response.
$ 8 2 23 - ADDITIONAL ACCIDENT- MONITORING INSTRUMENTATION /II F 1}
Requirement superseded by NUREG 0737. Refer to Subsection 18 1.30 for response.
18 2 24 - INADEQUATE COQE COOLING INSTRUMENTS~II F 2}
~
18- 2-9
SSES-PSAR 1
1 Requirement superseded .by NUREG 0737. Refer to Subsection 18.1.31 for response.
18,2 25 - ASSURANCF, OF PROPER ESF FUNCTIONING /II K 1 5i 18.2.25.1 Statement of Reauirement Review all valve positions, positioning requirements, positive controls and related test and maintenance procedures to assure proper ESP functioninq., This requirement shall be met by fuel load.
18 2.25'.2 Interpretation None required.
18.2.25.3 Statement of Response Operating. and surveillance procedures are currently being developed. writing the procedures to reflect ESP reguirement is a key ohlective of procedure originators. Additionally, these procedures vill receive a review (independent of the oriqinator) to provide further assurance that the procedure is technically correct and provides for accomplishment of procedural objectives (includinq maintenance of proper safety function).
18,2 26 -
SAFETY RELATED SYSTEM OPERABILITY STATUS /II. K 1 ~10 18.2.26.1. Statement. of Reauirement ~
Review and modify, as required, procedures for removing safety-related systems from service (and restoring to service) to assure operability status is known. This requirement shall be met by fuel load.
18.2 26.2 Interpretation None required.
18 2-10
SSES-PSAR 18 2.26 3 Statement of Response-Surveillance testing will be controlled by administrative procedure AD-gA-022. This procedure, which is currently being drafted, will require that surveillance implementing procedures contain a review of redundant component operability prior to removing the system to be tested from service, (if such removal is required bv the test), a review of proper system status prior to return of the tested system to service, and provide for notification to Operations of the need for system status changes.
,Administrative Procedure AD-QA-306, "System Status and Equipment Control,>> {see Subsection 18.1.13.3) establishes control of system status as an operations responsibility and vill provide the same reviews described above during normal operations and maintenance activities. Maintenance procedures will only cover activities while systems and components are removed from service, the Operations section will actually accomplish changes in system status as controlled by the described Instruction.
18 2 27 TRIP PRESSURIZER LOW-LEVEL COINCIDENT SIGNAL BISTABLES
-)II K 1 $ 7)
This requirement is not applicable to Susguehanna,SES.
18 2 28 OPERATOR TRAINING POR PROMPT MANUAL REACTOR TRIP /II.K. 1 2g)
This requirement is not applicable to Susguehanna SES.
18 2 29=- . -AUTOMATIC SAFETY GRADE ANTICIPATORY TRIP @II K 1 21}
This requirement is not applicable to Susguehanna SES.
18 2 30 AUXILIARY HEAT REMOVAL SYSTEMS OPERATING PROCEDURES
/II K $ g2) 18.2.'30~1 Statement of Requirement Describe the automatic and manual actions necessary for proper functioning of the auxiliary heat removal systems that are used
SSES-FSAR when the main feedwater system is not operable. This requirement shall be met by fuel load.
- 18. 2.30. 2 Interpretation None reguired.
- 18. 2.30.3 Statement of Response A generic response to this requirement vas provided by General Electric in NED0-24708, >>Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors,>> (August, 1979) and supplement I. A plant specific description is provided belov.
If villthe main feedvater'ystem is not operable, a zeactor scram be automatically initiated when reactor water level falls to Level 3 (540.5 inches above vessel bottom or 178.2 inches above the top of the active fuel) . The operator can then remote manually initiate the reactor core isolaticn cooling system from the main control room, or the system vill be automatically initiated vhen. reactor water level decreases to Level 2 (489.5 inches above vessel bottom or 127.,2 inches above the top of the active fuel) due to boil-off. At this point, the high pressure coolant injection system will also automatically start supplying makeup water to the vessel. These systems vill continue automatic injection until the reactor water level reaches Level 8 (581.5 inches above vessel bottom or 219.2 inches above top of the active fuel), at which time the high pressure coolant injection turbine and the reactor core isolation cooling turbine are automatically tripped.
In the nonaccident case, the reactor core isolation cooling system is utilized to furnish subsequent makeup vater tvo the reactor pressure vessel. The Reactor core isolation cooling system and'he high pressure coolant injection system vill restart automatically vhen the level falls to Level 2 (The reactor core isolation .cooling system is being modified to automatically restart, see subsection 18.1.50). No manual actions are required for these systems to restart. Reactor vessel pressure is regulated by the automatic or remote manual operation of the main steam relief valves which blov dovn to the suppression pool.
To remove decay heat, assuming that the main condenser is not available, the steam condensing mode of the residual heat removal system is initiated by the operator. This involves remote manual alignment of -the residual heat removal system valves. If the 18 2-12
SSES-PSAR steam condensinq mode is unavailable for any reason, the main steam relief valves can be manually actuated from the control room. Remote manual alignment of the residual heat removal system into the suppression pool cooling mode is then required for suppression pool heat removal. Makeup water to the vessel is still manual supplied by the reactor core isolation coolinq system under control.
For the accident case with the reactor pressure vessel at high pressure, the high pressure coolant injection system is utilized to automatically provide the required makeup flow. No manual operations are reguired since the high pressure coolant injection system will cycle on and off automatically as water level reaches Level 2 and Level 8 respectively. If the high pressure coolant
~
injection system fails under these conditions, the operator can manually depressurize the reactor vessel using the automatic depressurization system to permit the low pressure emergency core cooling systems to provide makeup coolant. Automatic depressurzation will occur if all of the following signals are present: high drywell pressure 1.69 psig, Level 3 water Level permissive, Level 1 water level (398.5 inches above vessel bottom or 36.2 inches above the top of the active fuel), pressure in at least one low pressure injection system and the run out of a 120 second timer (set at 105 seconds) which starts with the coincidence of the other, four siqnals.
18 2 31'- REACTOR LEVEL INSTRUMENTATION /IX,K,l.23)
- 18. 2.31,1 Statement of Requirement For boiling water reactors, describe all uses and types of reactor vessel level indication for both automatic and manual initiation of safety systems. Describe other instrumentation that miqht give the operator the same information on plant status. This requirement shall be met before fuel load.
18.2.31.2-Interpretation None required.
18~3$ ,3 Statement of Response-The response to this requirement was provided by General Electric in NED0-24708, Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors," (August 1979) and Supplement I.
18 2-13
SS ES-FS AR 18 2 32 CONMISSION ORDERS ON BABCOCK AND WILCOX PLANTS /II K.2}
These requirements are not applicable to Susquehanna SES.
18 2 33 REPORTING REQUIREHENTS FOR SAFETY/RELIEF VALVE FAILURES OR CHALLENGES /II. K. 3. 3g
- 18. 2.33.1- Statement of Requirement Assure that any failure of a PORV or safety valve to close vill be reported to the NRC promptly. All challenges to the PORVs or safety valves should be documented in the annual report. This requirement shall be met before issuance of a full-power license.,
18.2.33.2 Interpretation Prompt reportinq to the NRC consists of notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone with confirmation by telegraph, mailgram or facsimile transmission, folloved by a written report vithin 14 days The annual operating report has been supplanted by more detailed
~
Monthly Operating Reports.
~
Documentation required to be included in the annua1 report vill be supplied in Monthly Operating
~ ~
Reports.
18..2 33.3 Statement of Response Subsection 6.9.1.8 of the Technical Specifications requires prompt reporting with vritten follovup for failures of main steamline Safety/Relief Valves to reclose after actuation.
Subsection 6.9.1.6 of -the Technical Specifications requires documentation of all challenges to main steamline Safety/'Relief Valves to be included in the Monthly Reactor Operating Report.
18 2 34 = PROPORTIONAL-INTEGRAL DERIVATIVE CONTROLLER /II K 3 9}
This requirement is not applicable to Susquehanna SES.
18 2~35-
- ANTICIPATORY REACTOR TRIP NODIFICATION /II K 3 10$
18 2-14
SSES-FSAR This requirement is not applicable to Susquehanna SES.
18 2 36 = POWER OPERATED RELIEF VALVE FAILURE RATE @II K 3.11)
This requirement is not applicable to Susquehanna SES.
18.2.37= ANTICIPATORY REACTOR TRIP ON TURBINE TRIP~II.K.3.12}
This requirement is not applicable to Susquehanna SES.
18 ~ 2 38 - EM'ERGENCY PREPAREDNESS-SHORT TERM (III A. 1 1)
- 18. 2.38.1 Statement of /equi.rement Comply with Appendix E, "Emergency Facilities," to 10 CFR Part 50, Requlatory Guide 1.101, >>Emergency Planning for Nuclear Power Plants," and for the offsite plans, meet essential elements of NUREG-75/ill (Ref. 28) or have a favorable finding from ZEMA.
This requirement shall be met prior to fuel load.
Provide an emergency response plan in substantial compliance with NUREG-0654, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants" (which may be modified as a result of public comments solicited in early 1980) except that only a description of and completion schedule for the means for providinq prompt notification to the population (App. 3), the staffing for emergencies in addition to that already required (Table B.l), and an upgraded meteorological program (App. 2) need be provided (Ref. 10). NRC will give substantial weight findings on offsite plans in judging the adequacy against NUREG-0654.
Perform an emerqency response exercise to test the integrated capability- and a major portion of the basic elements existing within emergency preparedhess plans and organizations. This requirement shall he met before issuance of a full-power license.
- 18. 2.38.2 Interpretation PPGL is interpreting Emergency Facilities as encompassing those requirements for TSC, Interim TSC, EOF, Interim ZOF, SPDS, OSC as outlined in NUREG 0696 and TMI Action Items in 0737. Develop Site, State, County, Township and Municipality Emergency Plans usinq the Guidelines of NUREG-0654 Rev. l.. Exercise the plans to 18 2-15
SSES-FSAR ensure they are integrated and workable. Comply with meteoroloqical requirements of NUREG 06S4 Rev. 1 Appendix 2.
18,2.38.3 Statement of Response The proposed method of responding to this requirement was submitted by a letter to B. J. Youngblood from N. W. Curtis on April 2, 1981 (PLA-704) . Details on the emergency response facilities are presented in the Emergency Plan.
18 2 39- UPGRADE EMERGENCY SUPPORT FACILITIES /III A 1.2i Reguirement superseded by NUREG 0737. Refer to Subsection 18.1.67 for response.
18 2. 40 PRIMARY COOLANT SOURCES OUTSIDE CONTAINMENT /III l. lj D
Requirement superseded by NUREG 0737. Re fer to Subsection 18.1.69 for response.
18 2 41- - INPLANT RADIATION MONITORING'EEI.D. 3 3g e Requirement superseded 18..1 70 for response.
by NUREG 0737. Refer to Subsection 18~42 - CONTROL RQQQ HABITABELITY /III D 3 4g Reguirement superseded by NUREG 0737. Refer to Subsection 18.1.71 for response.
18 2-16
1