ML20245B964
| ML20245B964 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 04/29/1987 |
| From: | Wessman R Office of Nuclear Reactor Regulation |
| To: | Miraglia F, Murley T, Sniezek J NRC |
| Shared Package | |
| ML20244D847 | List:
|
| References | |
| FOIA-88-198 NUDOCS 8904260408 | |
| Download: ML20245B964 (251) | |
Text
{{#Wiki_filter:U n .u . i i.'.ui.i .J : . :J. . . :4.sl.T. ~.m . ..:.s'. . m. .Ie. ? .%Gis.. h u, t.b . ':. .D.L ' [p, Men 4 UNITED STATES y g NUCLEAR REGULATORY COMMISSION 5 j WASHINGTON, D. C. 20555
~s.,...../ APR 2 9 jg Docket No. 50-293 MEMORANDUM FOR: T. Murley* J. Partlow # R. Capra J. Sniezek* F. Congel
- W. Butler F. Miraglia* W. Russell t J. Stolz R. Starostecki* S. Black
- E. Adensam.
S. Yarga* B. Boger L. Rubenstein D. Crutchfield* G. Lainas B. J. Youngblood-L. Shao* 'F. Schroeder C. Rossit G. Holahan THRU: ictor Nerses Acting Director
, roject Directorate I-3 Division of Reactor Projects-I/II FROM: Richard Wessman, Project Manager Project Directorate I-3 Division of Reactor Projects-I/II
SUBJECT:
DAILY HIGHLIGHT PILGRIM RESTART SCHEDULE SLIP On April 27, 1987, Boston Edison Company's Senior Vice President, Nuclear (Ralph Bird) stated that the Pilgrim restart date has slipped from the previously published restart readiness date of June 1987 to August 1987 Mr. Bird's statement was part of testimony before the Massachusetts State Legislature's Joint Committee on the Investigation and Study of the Pilgrim Nuclear Generating Facility. Mr. Bird further stated that the Pilgrim detailed schedule was currently being evaluated and that Boston Edison expected to reload fuel in the reactor (currently defueled) in June 1987. Pilgrim shut down April 11, 1986 due to equipment difficulties and has remained shutdown since that time to resolve equipment, operational and management issues. . b Richard H. Wessman, Project Manager Project Directorate I-3 . Division of Reactor Projects - I/II q i 8904260408 890419 FOIA [d i PDR PDR l JOHNSONBB-198
~
- w. D:V r c,.;:.. w 2 7 : ' c.w: cu ,:
.::: .; .y a. . s 3:: ;L -
L. . w. n./ ' :n .. i t MAY 1987 OVERVIEW REPORT PLANT NAME: Pilgrim-1 PROJECT MANAGER: R. H. Wessman i MONTHLY OPERATIONAL OVERVIEW FOR APRIL 1987 The plant continues to be down for an extended outage for refueling, resolution of equipment and operational difficulties, management issues and emergency planning concerns. Restart not expected until Fall 1987. Monthly Licensing Action Sumary: TRIPS THIS MONTH 0 TRIPS LAST 12 MONTHS 0 NUMBER OF REPORTABLE EVENTS 0 , LICENSING ACTIONS Opened in May 1987 - 3 Closed in May 1987 - 0
*0utstanding 05/31/87 - 40
- Reflects a correction in April total to reflect total open TACS. .
j SIGNIFICANT ISSUES AND EVENTS Significant Activities During May 1987
- Management meeting held on May 7, 1987 at the site to discuss the 1986 SALP report.
- On May 8,1987, the licensee announced changes to the current Nuclear Organization structure. The new organization structure includes a new Division of Planning and Restart which will plan and coordinate outage and startup activities. Mr. D. L. Gillispie was named the Director of Planning and restart. Mr. Gillispie will report directly to the Senior Vice President-Nuclear.
- PM and BECo Licensing representatives met in Bethesda on May 21, 1987 to review licensing actions.
- Comissinner Carr visited Pilgrim May 22, 1987.
- Licensee reported May 6, 1987 (10 CFR 50.72) that seismic qualification of hydraulic control units may be in question due to defects in hold-down bolt installation. A Part 21 report was verbally made on May 29, 1987.
} The Events Assessment Branch is following generic aspects. m.
' T ,i G a d E 2 2 Z ~?" O n X 0 h . c h " O h 4 ' c s b VO '1i M E .
. 5.
- W.':...: O r?
s
-Significant Meetings Scheduled During June lo87 (NRR, Region, IE, etc.)
None Scheduled Items for Management Attention
- Utility is still evaluating restart schedule; informal estimate from BEco is for restart about mid-September. This appears to be optomistic.
- PM preparing response to Gov. Dukakis of Massachusetts regarding Senator Golden's 2.206 Petition.
- Number of licensing actions to be completed before restart may require extra resource commitments.
Status of Items Previously Identified for Management Attention Much of staff work on response to 2.206 Petition of July 15, 1986 from Senator Golden is complete. We expect to have a product for senior management by mid-June and issue response by the end of June. Next Scheduled Refueling Outage February 1989 cc: V. Nerses R. Wessman PD 1-3 R/F M. Rushbrook l
, a n. . :.cawan. . w : m-: . . ~ .t. .a. w. a ; , . . .:. . . : .. E. u .
. a :/ '. . * -e
~..- ,
~
e
, No.: JII 7 SAFETY EVALUATION PILGRIM NUCLEAR..PCNER STATION Rev. no. 0 .
l PDC Pol System Calc. . Initiator: Dest: Grou): No.: Name: No.: Date:
"T% h ALASKE NE.b 5S&mc. 86-52C FM , 6"/3/ N P%TELT<aA/
Description of Proposed change, test or experiment: 1"M S TA LL 4 To C M OC 5EP DIESEL WIRE Pa M P SAFETY EVALUATION CONCLUSIMS: The proposed change, test or experiment:
~
- 1. (4 Does Not ( ) Does increase the probability of occurren'ce or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR. *
- 2. Od Does Not ( ) Dois create the possibility for accident or malfunction of a different type than any evaluated previously in the FSAR.
- 3. (4 Does Not ( ) Does reduce the margin of safety as defined in the basis
- for any technical specification.
RASIS FOR SAFETY EVAlt1ATION CDICLLISICDIS: SEE ATTACHFB S 6J FE -r'S Change Gange . 00 Recommended ( ) Not Recommended SE Performed by N w @ d Date 5/2.de7 Exhibit 3.07-A Sheet 1 of 3 ,_ _ ISSUED FOR l CONSTRUCTION 3.07-13 Rev. 4 f V
..G w . .m a.aaa;. u . . , ,e . .a ,w.a,r.. , . . . . a.; u., _w,. ; y , ; .
. , ; , .g.; t. , . ,_ , , . _ ,
~
1AFETY EVAlt!ATION 7 2-[5 PILCRIM NUCLEAR PCMER STATION g Rev. No. O A. APPROVAL * (K) This proposed change does not involve a change in* the Technical l Specifications. - l (N This proposed change, test or experiment does ( ) does Mt (4 l involve an unreviewed safety question as define ( in 10CFR, Part 1 50.59(a)(2). 09 This proposed change involves a change to the FSAR per 10CFR
)
l 50.71(e) and is reportable under 10CFR50.59(b). i ( ) Comments: The safety evaluation basis and conclusion is: ,
, (4) Approv .
f( Not Approved {
/. < eau ch/m /e n)) 6-+ n i Discipline Group Leader /Date ting cipline GFoup Leader /Date
{ B. R M EN APPROVAL . ( ) Comments: Nw$&$OGroup 6Leaddr/Date L b () l C. ORC R M EN . j
)
( ) This proposed change involves an unrevleved safety guestion and
- a request for authorization of this change must be flied with the Directorate of Licensing, NRC prior to implementation. l C YThis proposed change does not involve an unreviewed safety guestion. _
ORC Chairman N4 h / Date /o //o/F 7 ORC Neeting Nu r 78[ cc: Exhibit 3.07 A
. Sheet 2 of 3 -
_1,,=.__- ISSUED f610 3.07-14 CONSTRUCTION *= . l b
,~
Si&1MDX]tuM2Eih%%5MES.iss?. NH W i2.: O G l% %: ~- % W &.atW 6: + . 4 Safety Evaluat1= No. e
"-~~i55UED FOR PDC 86-52C, Rev. O Sheet , of .m a I CONSTRUCTION 1
A. DESCRIPTION OF PROPOSED CHANGE, TEST OR EXPERIMENT The proposed modifications are as follows:
- 1. Installation of a SEP (Safety Enhancement Program) diesel driven fire pump (P-179), which takes suction from the city water main and discharges to the Fire Protection System main yard loop. The main yard loop _
will discharge into the Residual Heat Removal System Header crosstie installed by PDC 86-52B. The SEP diesel fire pump is rated for' 750 gpm at 1254(j 7 psi and will be equipped with a 150 hp engine, 165 gallon fuel tank, relief valves, instrumentation, engine cooling system, engine exhaust system, engine starting system, and pump controller. The pump will take suction from the city water main through > approximately 20 feet of 6" diameter carbon steel and ductile iron piping. A gate valve will be provided for isolation and a removable strainer will be provided for startup. The discharge of the fire pump will tie into the Fire Protection system main header, 12"-KC33, between valves 12-D-7 and 12'D-8. Connecting piping will be approximately 180 feet of 6" diameter carbon steel and-ductile iron piping. The piping will include a check valve to preclude backflow at the pump discharge, a gate valve for manual isolation at the pump discharge, and a gate valve for manual isolation at the connection to the main header. _
- 2. Installation of a diesel fuel oil transfer system in order to operate the SEP diesel fire pump for more than eight hours (capacity of the diesel engine day tank).
The system will consist of a hydroturbine directly connected to a positive displacement (PD) fuel oil transfer pump (P-180). The hydroturbine takes suction from the SEP diesel fire pump discharge and returns to the fire pump suction, it uses approximately 40 gpm. gen er r -- The PD pump takes suction from the underground blac ,** in PDC diese15 fuel oil storage tanks (installed [ /g.g[g
, yg 56A) and discharges to the SEP diesel fire pump fuel oil day tank (T-162) at a rate of approximately 5 gpm.
The fuel tank will have an overflow line back to the underground tanks. The PD pump and hydroturbine l
C~i.~DWMMbi%.:si:'~55!.752:asms
. :X C M C.WJ%W'"' #:,iGV, L 5 - '
)
. I l
~
= .- '
Safety Evaluation No.ch d
" S5-52C, Rest . o ISSUED FOR Sheet 4- of jgEff?
[C >NSTRUCTION suction and discharge lines include manual isolation and control valves. The PD pump and hydroturbine will be located in the SEP diesel fire pump enclosure.
- 3. Installation of an enclosure for the SEP diesel fire pump. The SEP diesel fire pump and its enclosure will be installed near the main gu~ard house. To accommodate this, the abandoned construction water fire pump, its valves, enclosure and foundation will be removed except for the west and south walls of the existing enclosure which support existing electrical equipment connected to other buildings. These two walls will remain intact until removed under another modification. Existing piping and electric feeds will remain for use as needed. .
The enclosure will be equipped with a fire protection
~
system. The system will be of the wet pipe type with j valves, controls an sprinklers supplied from the fire protection system. ! The enclosure, a 20' x 24' prefabricated metal structure, will contain SEP diesel fire pump P-179 and accessories, diese.1 fuel oil transfer pump P-180, and all above ground piping. The enclosure will have four thermostatically controlled space heaters and a manually controlled ventilation system to maintain a temperature range of 60:F to 100cF. The enclosure will also contain lighting and access doors for maintenance. Electric power to the diesel fire pump room lighting and space heaters is provided from MCC B 40. This MCC - is supplied with normal station power, with an automatic transfer to the Blackout Diesel Generator. However, the system will operate without the space heaters operating. B. PURPOSE OF THE CHANGE The purpose of this modification is to provide a redundant water source to the existing diesel driven fire pump (P140), to the RER system for containment spray and RPV injection I during extended station blackout and severe accident scenarios (loss of core cooling and containment spray) beyond the current plant design bases. The equipment installed by this modification, with the exception of the enclosure's lighting and space heaters, operates w-_______-_____-. _ _ _ _ _ - _
- c. . . . - war. wmrz.ww.;..m.:.uw . a:: r.a r,c.
=- L m ::;w:; a : -
'Eua.. .: '
Safety Evaluation No. Al@ PDC 86-52C, Rev. O Sheet [ of M 8 independently from normal plant power, the emergency diesel generators and the blackout diesel generator. Rep,rgp/ g[egro / . Mr i5 +rw 4 C. SYSTEMS, SUBSYSTEM, COMPONENTS AFFECTED j@g g fp gfca -
- 1. Fire Protection System , 7 The SEP diesel fire pump discharge line is connected to @#1 the Fire Protection System main header, 12"-KC-33, between valves 12-D-7 and 12-D-8.
- 2. City Water System The SEP diesel fire pump suction line is connected to the city water system. Capability is provided to isolate the' city water from the rest of the plant c
services so as to prqvgpe__ = v4 m ..; apab414+y .to the
,1 . fire pump.
- 3. Blackout Diesel Gener t N
Fuel oil transf er pu:mp--..suctic= 7 ;- g i; ....nedted to 4 the 2" blackout diesel generator fuel oil tank ru ply lines. ehdoal du # The SEP diesel fire pump enclosure li;;hting will be% powered from the Blackout Diesel Generator subsequent of the failure of normal station power (Ref. PDC 86-56). gc M7 Auxiliary Electrical System . .'tv.8 tin , iL Nacks Electric power to the new diesel fire pump room / / g g q'g n e r M r- lighting and maintenance loads will b provided from o n orr4 gggg __ MCCBj,whichissuppliedfrombusB-5 D. SAFETY FUNCTIONS OF AFFECTED/ COMPONENTS C 64/gp
- 1. Fire Protection System The Fire Protection System does not perform any safety related function. The Fire Protection System is important to safety ecause it mitigates the occurrence of equipment. damage from a fire that could damage or significantly degra the capability of one safe shutdown system in th plant.
g,Q kgr het u tr
% . &#(6wMj d4/
<. f w ct W 7
p .wr%2WM";,;m w.u.2 u . . :.2 n.mSL em ' ' X ' w ! ; ; % J:W ' ~~ .:2; W .t,:c.~;;g - - - Safety Evaluation No.el M PDC 86-52C, Rev. O Sheet L of $ 8
- 2. City Water System The city water system performs no safety function. It does, however, act as a backup water supply for the Fire-Protection Water System.
- 3. Black 0'r; Jiesel Generator The Blhak0ut Diesel Generator will function as a backup to the existing Emergency Diesel Generators. The Blackout Diesel Generator is not required to perform any safety-related function.
- 4. Auxiliary Electrical System ThesourceohpowerforMCCB40isLCB-5. Bus B-5 is
. supplied from Bus A-4, uhich is a non-safety related I - ar ::.c . - -
~
electrical bus. ' ISSUED Foi\ o E. EFFECT ON SAFETY FUNCTION
- 1. Fire Protection System 4- CONSTRUC" TION The ' Fire Protection System does not perform any safety related function. However the new 6" line, the normally closed gate valves, check valve, the fire ,,
pump, and the associated vent and drain lines form a part of the Fire Protection System pressure boundary and are required to maintain that boundary as is any other Fire Protection System component. -
)
t
- 2. City Water System The City Water System has no safety function to perform.
- 3. Blackout Diesel Generator The Blackout Diesel Generator is not required to perform any safety-related function.
- 4. Auxiliary Electrical System The affected portions of the Auxiliary Electrical System perform no safety related functions, g) g rZ t% p ed is O&& &lp b !
,p a swh pa n- 100m g,e p , 5 ? hsyTK we d 4 _ s,w y , L eyL a '~er a m~.
ft X ':1:X S C'7; D i.id u a.'n s.L~XLC 1 W .4 , % 7 ._ F TU hhOf.~ ' L & ' a 1.L s. Safety Evaluation No. 240 PDC 86-52C, Rev. O Sheet '7 of Mg F. ANALYSIS OF EFFECT ON SAFETY FUNCTIONS Fire Protection System In order to ensure the Fire Protection System pressure boundary is maintained, the 6" line, the normally closed gate valves, check valve, the fire pum , and the associated vent and drain lines are designate ire Protection). G.
SUMMARY
84I N'r d ! Mf CH/0 t No safety related system is affected by this modification. - This modification does not involve a potentially unreviewed safety question. ..
.- .This modification does not require a Technical Specification change.
The demolition of the existing construction fire water pump building and erection of the new SEP Fire water pump enclosure will not affect any safety related system or safety function.
- . = _ __ _.
ISSUED FOR CONSTRUCTION _ l
- a. ; . :, c:. . . ..:~ac.M'n. ::&. .t. u. n... :..ca. . .-.uw.. -v : . ~ ; b . 4:. , ; c .
PILGRIM STATION g FSAR REVIEW SHEET g
References:
Safety Evaluation: 159 Rev. No.: O Date: Support a change fo reo viole q SEP ddse f f;pi gy List FSAR test, diagrams, and Indices affected by this change and corresr>onding FSAR revision. , Affected FSAR Revision to affected FSAR Section is shown on: sec 6.1on Pre 11sinary Final j 10, 8 Attachment 1 l l E ure. 10. 5'- I F; yare Io.8-l M E S Sheet l. ( Mt8 Sheet 3 - - -
--=----- ,.
\
Attache,nt A ISSUED FOR .. I Attachment 5 _ CONSTRUCTION Attachment 5 {
, )
PRELIMINARY FSAR REY 1510N (to be completed at time of Safety Evaluation preparation). Prepared by: O b S /Date: bReviewedby* Date: C/// B7 Approved byk A'M /Date: 6[/[91 FINAL FSAR REVISION (Prepared following operational turnover of related systems structures of components for use at PNPS). i Prep? fed by: /Date: Reviewed by: /Date: 1 Attach completed FSAR Change Request Form (Refer to NOP). I Exhibit 3.07-A Sheet 3 of 3 4 3.07-15 Rev. 4 ! I 6 2 __ _ _ _ _ _ . _ _ _ _ _ = _ -- _ _ l
?imi:HaiG72r:Pamar s.::D n.e.WGa: ..:w :4> . & ~': = ~ 'L~~J,.
. ~ ~.tJ:n; - ~ ~ , u :2: t J... !
l c - 1 si / ni,,-
/ '
Attachment 1 Safety Evaluation No. Q M ! PDC 86-52C, Rev. O Sheet / 0f 2.- , RECOMMENDED FSAR CHANGES Section 10.8.3 of the FSAR shall be updated due to this modification (PDC-86-52C). FSAR page 10.8-1 has been marked with suggested updates and included in this attachment for your review. The following FSAR figures (drawings) will be revised as part of the Plant Design Change Package (PDC 86-52C), but are not included herein. DWg I. D. FSAR FIGURE TITLE M-218 Sheet 1 10.8-1 P&ID Fire Protection System M-218 Sheet 3 10.8-1 P&ID Fire Protection System __=:=_ ISSUED FOR - CONSTRUCTION : l l i _ .__ _ _ . _ . _ _ . _ . _ . _ . _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _
7 2WJI.12.MriJR WRJ7 c.Ch Azale.L OR E ,c . C ' " " ' 1.; . ;5 %G . ' / hG
)
, _lm, /, -
~
Atter.hmenB f 1 1
- .. ~
. PNPS-TSAR -
Safety Evaluation No. O 6*l I 10.8 TIR5 PROTECTION SYSTEM i I 10.8.1 Fower Ge'neration objective
.. o
,,,The power generation objective of the Fire Protection System is to provide adequate fire protection capability in all areas of the r- - _:_ _ ._ ,
___.. station. _- 10.8.2 Power Generation Design Basis , The Tire Protection System is designed to furnish water, halon, carbon dioxide, and/or dry chemicals as necessary for fire extinguishment in the station. The Fire Protectiop._jystem is ,,,_,_ designed to provide the following { I UED FO,R j A reliable supply of fresh water for
- 1. ,
A reliable system for delivery of wa er *a UCTION
. m+eMid !E
_ 2. locations - i
- 3. Automatic fire detection in selected areas q
- 4. Fire extinguishment by fixed equipment activated either automatically or manually for areas with a high fire risk
- 5. Manually operated fire extinguishing equipment for use by l operating personnel at selected points throughout the s station i 10.8.3 Description .
The Fire Protection System, Piping and Instrumentation Diagram is shown on Figure 10.8-1, sheets 1 and 2. l 1 10.8.3.1 Tire Water System ! The site fire water supply is taken from two 250,000 gal lined - carbon steel water tanks which are devoted exclusively to fire protection. The Fire Water system may also use water from a city water main. The water supply is delivered by either an electric motor-driven purnp (rated at 2,000 gal / min) or a diesel engine driven pump (rated at 2,000 gal / min). The diesel engine driven purep is used for standby and emergency use on loss of ac power. A small jockey pump (rated at
, 50 gal / min) $s provided to maintain a constant pressure fter the water system. If the system pressure drops substantially. the motor-driven t
fire pump will start automatically, and if pressure coratinues to dron, the diesel-driven pump will also start automatically, a n'e pumps feed outdoor fire hydrants, interior hose stations, Sprinkler Systems, and Deluge Systems for the station. w 5
--~
A Safety Enhancement Pro ram diesel engine driven fire pump (rated at , J 750 gpm), which can be manua ly aligned and started, and which takes suction I from the city water main is connected to the Fire Protection System. This / pump is installed as an additional backup source to feed the Residual Heat l Removal System during an extended station blackout through a removable piping)' crosstie. 10,8-1
# l
______s__ _____m_ .___m_ f
q = ::.; u a:a x m s. m u:a. x.c c.. ..w a , , : = . ; c .g,x,,;; . ,__. a a_. . , , , Safety Evaluation //t [? No.: 2 t< 9 i SAFETY EVALUATION WRK SHEET Rev. No. O I A. Systee Structure Component Failure and Consequence Analyses. Systee Structure e - nent Failure Modes Effects of Failure dunnents
,. Ge %%% See A+&achel %e-Svshen
< 1
- 2. I ISSU 5U. FOR .
CONSTRUCTION General Reference Material Review FSAR CALCULATIONS REGULATORY SECTION PNPS TECNNICAL SPECS.'. DESIGN SPECS PROCEDURES GUIDES STANDARDS CDDES fo. 8 3,12, 8 Spec . M- 3co lo cng so Ffare lo s I Spee . M-30l Ausr est. I A 21.58 - g l Spee , y _3o 3 bryenkx. h See c . M- 50 4-Arsnelix C B. For the proposed hardware change, identify the failure modes that.are likely for the components consistent with FSAR assumptions. For each failure mode, show the consequences to the system. structures or related _ components. Especially show how the failure (s) affects the assigned safety basis (FSAR Text for each system) or plant safety functions FSAR '. Chapter 14 and Appendix G) Prepared by db N- d Date U NOTE: It is a requiresient to include this work sheet with the Safety Evaluation. . Exhibit 3.07-C { 1 3.07-18 Rev. 4 4 _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ f
y;E % R % :L: M e n s .c = w:.:.tv.w. %u: ~. V .: a W.- ~" V :s r .:L a J:5..: L~ n .a ? j p' j l .s i Safety Evaluation No. d d PDC 86-52C, Rev. 0 , Sheet M of M
- 2. L l l
l i i SAFETY EVALUATION WORK SHEET l l
- 1. System / Structure / Component: Fire protection system l l
Failure Mode: Structural failure of the 6" piping interface with the Fire Protection System Effects of Failure: Degradation of the cecability of the Fire Protection 's t e m to perform its function. l Comments: The 6" piping interface with the Fire Protection System is designated FP-Q. l fb --- . % _ ISSUED FORD { CONSTRUCTION . _ l i 1 i 1 l _m_____a___._____.____.___ __t _ __ .
.x ,.u m .c; a x;:...:. 2. .a . . ...,u: w -- .:- , ,_ ;
.,~ J .
.. l NED Proposed Change Safety Evaluation '
1 No.: .2 3 f ,
# mr 6
,._a,u., adCLEAR POWEk 3in,iun ;
l l Rev. No. 0 - PDC PCM System Calc. Initiator: Dent: Group: No.: Name: No.: Date: 5/12/87 Fuel Oil TMHauske NED FS&MC 86-52D Transfer 4 System Description of Proposed change, test or experiment: Fuel Oil Transfer System for niesel Fire Pumn P-140 SAFETY EVALUATION CONCLUSIONS: l The proposed change, test or experiment: 1-
~
l 1. (x) Does Not ( ) Does increase the probability of occurrence or l consequences of an accident or malfunction of equipment laportant to safety previously evaluated in the FSAR. '
- 2. (x) Does Not ( ) Does cr.eate the possibility for accident or malfunction of a different type than any evaluated previously in the FSAR.
l 3. (X) Does Not ( ) Does reduce the ma'rgin of safety as defined in the basis l for any technical specification. BASIS FOR SAFETY EVALUATION CONCLUSIONS: See Attached Sheets i4 pues) Change Change (>d Recommended ( ) Not Recommended SE Performed by Date 5 !/ P_ &7 Exhibit 3.07-A Sheet I of 3 l l l 3.07-13 Rev. 4
us ca. w, :.w. '.
- .;as , . = u. . e . . , :
.:a , ,: . a. ' ~ .c ~ w -
. v--- Y s
- Safety Evaluation
'. - No.: ;i b $
SAFETY EVALUATION PILGRIM NOCLEAR POMER STATION NU EEb A. APPROVAL . K This proposed change does not involve a change in the* Technical Specifications. K This proposed change, test or experiment does ( ) does not C4 involve an unreviewed safety question as defined in 10CFR, Part 50.59(a)(2). X This proposed change involves a change to the FSAR per 10CFR 50.7)(e) and is reportable under 10CFR50.59(b). Comments: The safety evaluation basis and conclusion is: 4() Approved () Not Approved el d'bel8i Distipline Group Leader /Date .. Supporting Discipline Group Leader /Date
. 3. REVIEN APPROVAL Comments:
.26) % J v
LNW" S&SA Group Leader /Date C. ORC REVIEN . C ) This proposed change involves an unreviewed safety guestion and a request for authorization of this change must be filed with the Directorate of Licensing. NRC prior to laplementation. (j) This proposed change does not involve an unreviewed safety question. CRC Chatrean i Date #!MM2 ORC Meeting Number O Exhibit 3.07-A Sheet 2 of 3 - l Rev. 4 3.07-14 )
m =:n. . . . . w - ..;.,, .
...r. ;
. . . .> u.; . a =- ,.xy- -z : 7--- yc;p u.
Safety Evaluad on No. 2 3.f ; 7 DEST 3 oF (, N xtiotion of Chanet - Provide a backup (non-electrit.; s ower dependent) diesel oil transfer* system to the diesel fire pump P-140. This modification consists of a hydroturbine driven by existing diesel fire pump P-140. The hydroturbine in turn drives the backup diesel transfer pump which takes suction from the emergency diesel generator fuel oil storage tanks (bypassing existing diesel transfer pump P-141A) and discharges to diesel fire pump (P-140) day tank T-123. This modification involves the following activities: >
- 1. Installation of a hydroturbine, fuel oil transfer pump, the'r foundation, and piping and valves in the intake structure.
- 2. Connecting the hydroturbine inlet and outlet to the diesel fire pump ~
(P-140) discharge and suction lines respectively.
- 3. Connecting the backup diesel fuel oil transfer pump inlet and outlet to the discharge of the exis. ting diesel oil transfer pump P-141A and to the day tank fill piping'
- 4. Installation of a bypass line and valves connecting the suction and discharge lines of the diesel oil transfer pump P-141A.
B. Purcose of the Chance The purpose of this modification is to provide a redundant (non-electric power dependent) diesel fuel oil transfer pump for the diesel fire pump P-140. This redundant pump will allow extended operation of the diesel fire pump as a water source for the RHR system during extended station blackout and severe accident scenarios beyond design basis. C. Systems. Subsystems. Comoonents Affecttd _ This modification affects the Diesel Oil Storage and Transfer System in the following manner:
- The 2X-HG-38 suction pipe of diesel oil transfer pump P-141 A is affected by this modification. 1"-HG-38 line bypassing pump P-141A is connected to the 2X"-HG-38 line.
- The IX"-HG-38 discharge pipe of the diesel oil transfer pump P-141 A is affected by this modification. A IX"-HG-38 line bypassing pump P-141A, is connected to the IX"-HG-38 line.
- The IX" inlet pipe to fire pump day tank is affected by this modi fication. The suction and discharge lines of the backup diesel oil l
transfer pump are connected to this pipe. Page 1 of 4
aw .5.n 3.
~'
; '- -- - T" Safety Evaluation No. 21 M' S H E E t' & 0 5 & 1 C. Systems. Subsystems. Comoonents Affected W _nt) l 1 _ .
)
. n . ., i.r i.~ication Iffects the Fire Protectii -
. 1.1 ...e ..olowing manner: ,
- The 12"-KB-33 suction and discharge lines of diesel fire pump P-140 are affected by this modification. The 1X"-KB-33 (the hydroturbine inlet and outlet) lines are connected to these pipes.
This modification affects the Intake Structure in the following manner:
- The civil structure of the diesel fire pump (P-140) and the diesel day tank (T-123) rooms of this building are affected by this modification due to the installation of component supports and equipment foundations.
D. Safety Function of Affected System /Comoonents Diesel Oil Storage and Transfer System:
- The safety function of this system is to provide diesel fuel oil to the emergency diesel generators which provide a single failure proof source of on-site AC power adeq0 ate for safe shutdown on the reactor following ,
. abnormal operational transients and postulated accidents (Ref. FSAR Section 8.5).
Fire Protection System:
- The Fire Protection system does not perform any safety related function. Fire Protection System is important to safety because it mitigates the occurrence of equipment damage from a fire that could damage or significantly degrade the capability of one safe shutdown system in the plant.
Intake Structure:
- The Intake structure has the safety function of protecting the Class I salt service water system from natural phenomena (Ref. FSAR Section .
12.2). E. Effects on Safety Function Diesel Oil Storage and Transfer System:
- Operation of the new pump P-181 will make the existing diesel oil transfer pump P-141A inoperable.
- The structural failure of any new oil piping will degrade the capability of the existing diesel oil transfer pump to supply oil to the emergency diesel generators.
Page 2 rf 4 h
2.a . .n .u..c .u.m .- -
. . . .+ . . . ~ - -. .a , ,
Safety Evaluation No. 2il f SHEST 5 c F 6 e-fety Function (coq.': E. Ege; . - Fire Prote6 tion System: s
- The structural failure of the hydroturbine inlet or outlet will degrade i the capability of the diesel driven fire pump to perform its function'.
Intake Structure:
- The structural integrity of the Intake Structure is not degraded by the addition of a pump foundation or holes for anchor bolts.
) '
F. Analysis of Effects on Safety Functions Diesel Oil Storage and Transfer System:
- While the new fuel oil transfer pump P-181 is in operation, transfer pump P-141A will be inoperable since its suction valve (104A) will be closed. Transfer pump P-141B can be manually aligned and manually controlled to fill day tank T-124A (Ref. FSAR Section 8.5.2.7).
However, pump P-181 wil.1'only be used during station blackout, therefore, pump P-141A will not have power to start when its suction valve is isolated. Prior to testing pump P-181, day tank T-124A should be filled to ensure that pump P-141A does not automatically start while its suction isolation valve (104A) is closed.
- The following design features of this modification mitigate the effect of the structural failure of new components on the safet facsel Oil Storage and Transfer System ( a.a. i,1. bebc)y .
rm sf,fy function of , le The design change has Class I piping up to and inclusive of the globe (normally locked closed) valve which forms the boundary for the seismic portion of the Diesel Oil Storage and Transfer System. 2* All non-safety related items which have the potential of degrading the integrity of Class I items are analyzed for seismic forces of a safe _ shutdown earthquake (SSE) and are analyzed and installed as 0 Fire Protection System: l 4 %.s/de?
- The hydroturbine's suction and discharge lines were designed e -Q to minimize the possibility of structural failure degrading the capability ,
of the diesel driven fire pump to perform its function. In addition i the inlet and outlet lines will include two normally closed S.(f wa, sk/s7 ' isolation valves to isolaio the hydroturbine from the Fire Protection System during normal operation. These valves will only be opened during extended station blackouts, whe the hydroturbine is needed to transfer diesel oil from the emergency diesel oil storage. tank to the diesel driven fire pump's day tank.
' irunim-l rm r/,/g ;
Page 3 of 4 A - { irc pr ek cbos 6.P[c' Im 4 5/sb 7
w w a w tmu.-e z a:w . . c , '.. .:.. a m .: . : ..: "
~ ;c v - .,
Safety Evaluation No. 2iM'
'.' 5dEE7 6 oF 6
- r. Analysis of Urt,-'- n SAf tty Functions (cont.) . ,
m a The hy'droturbine will increase the availability of the dier.. u iven fire pump for extended periods of operation. The new fuel oil transfer pump P-181 will function as a backup to the existing pump P-141A for normal operation. Intake Structure:
- This modification was designed to assure that there is no structural degradation of the intake structure.
G. Summarv This modification does not call for the safety equipment of the system to operate at higher pressures, temperature or more severe conditions than the existing levels, hence, the modification does not increase the ; probability of malfunction of equipment important to safety. This modification is designed such that the failure of any non-safety related equipment will not. degrade the capability of safety related equipment to perform its function. This modification will increase the margin of safety for diesel driven fire pump availability for extended periods of operation, , This modification does not require a Technical Specification change. This modification does not involve an unreviewed safety question. Page 4 of 4
,,_ ; 7m,;- 7, mz ,,, ,y,, .,,,,.,.,7 , ,,, ;. ,7, 3. . y
. , , ,3, , , , . _
Safety Evaluation, No. 41/8 gev. o Sheet j of ,fa SAFETY EVALUATION
- PILGRIM NUCLEAR POWER STATION PDC PCN System Calc.
Initiator: Sept: Group: No.: Name: No.: Date: ) W. Riggs Engr. Mech. 86-53 Backup M-650 1 Nitrogen Supply i i Description of Proposed change, test or experiment: See Attachment I ! l SAFETY EVALUATION CONCLUSIONS: The proposed change, test or experiment: ,
- 1. (X ) Does Not ( ) Does increase the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR.
. 2. (X ) Does Not ( ) Does create the possibility for accident or malfunction of a different type than any evaluated previously in the FSAR.
- 3. (X ) Does Not ( ) Does reduce the margin of safety as defined in the basis for any technical specification.
BASf5 FOR SAFETY EVALUATION CONCLUSIONS: See Attachment II Change Change (X ) Recommended ( ) Not Recommended SE Performed by DateM!987 U Uf 77
- " = b 4 b i t ,1 FI T S (Chant 1 e( 1) Dov 1
- ,7 , J Attachment I 1 Safety Evaluation, Rev. 0 'E M* 41/8 I Sheet 4 of .to A.
DE5rRIPTION OF PROPOSED CHANGE. TEST OR EXPERIMENT . The desion chance covere<Lby._1hf._proppie.d modifications 4 5 as follows: 1 A.1 Addition of a liquid nitrogen /vaporfler trailer (X-168), and associated piping and valves. The trailer will be normally stored in the northwest corner of the plant site in a storage /laydown area. When needed, the trailer will be moved (via an on-site truck cab) outside the west gate of the condensate storage tank controlled access area. It is designed to deliver at two different pressures to the nitrogen system via two flex hoses, which will be stored on the trailer. Nitrogen at 120 psig will be available from the nitrogen trailer to match the existing drywell instrumentation supply pressure. Two inch supply piping and a globe valve will be added to tie the liquid nitrogen / vaporizer trailer into the existing nitrogen system. A check valve will be added to prevent nitrogen flow into the existing liquid nitrogen storage tank. Nitrogen at 70 psig will be available from the trailer - to match the existing drywell supply pressure. This will be accompitshed by connecting to the existing fill connection located on the north wall of the reactor building. A TEW/radrvt* M* V4ws wa4. ss repre As FAera THs 7994ew rWAM.. A.2 Addition of two banks of ten cylinders each, a cylinder rack f and manifold (X-169), associated piping and valves. The cylinders will be arranged to automatically maintain the nitrogen supply to drywell instrumentation once the existing nitrogen supply is not available, until the new liquid nitrogen / vaporizer trailer is available. The cylinders will deliver nitrogen gas at 110 psig (controlled by a vendor supplied pressure regulator), through 2 inch piping which will tie into the existing drywell instrument supply header. The new piping will contain a check valve, a gate valve and a relief valve. A differential pressure indication switch with annunciator will be connected between the cylinder supply and the existing supply to provide control room indication of switchover to the cylinders. The existing manual gate valve ,5 34-l, , @a (31-H0-162) will be changed from normally open to normally locked closed to isolate the instrument air supply to drywell instrumentation. Existing drywell instrument supply isolation valve A0-4356 will be modified from fail closed to fail open to maintain an open nitrogen path to drywell instrumentation following a loss of power. A.3 Modification of two existing Seismic Category I supports of the Auxi-11ary Nitrogen purge Supply (H9-1-11-19 and H91-12-1) as specified in Teledyne Engineering Services Document No. 6520-2.
~
t __= _ _____ - -__ --
n n: ~L .. ' O ' N
~ -
p + .6tu w ro.v m .--:.: 1 :; x. n:... ia X as' - ,
- .. .C: . : '. & + /5 F.Y~
l- . Attachment i Safety Evaluation, Rev. 0 (E. Alo.4//g Sheet 3 of Jo
- 8. PURPOSE OF THE CHANGE -
The safety enhancement program requires that various plant trodifications be implemented to insure the availability of those systems heeded during a station blackout. This modification provides two new non-safety related redundant nitrogen sources which will be available during a station blackout. The new nitrogen sources backup the existing containment inerting system, and the existing drywell instrument supply. The existing nitrogen storage facility will be normally aligned to supply drywell instrumentation during all modes of operation and the instrument air supply will be isolated from the drywell instrument supply header. e l A-_--_--- .-----.---------.-----------a-------__..a--a--a--- - , _ - - - - --__--. - - - . - - - - . - - - - - - - - - - - - - - _ _ . _ _ -----u - --_.a - . - . - - - - _ - -
- u. . . . . .~ .u. .~ ..n +.c,- m . a. . . ,. . :a . : .
. . a.. x .
. - w:~ c. ,
Attachment 11 ; Safety Evaluation, Rev. O S.E.M../H6 { Sheet 4 of .te j
- 0. BASIS FOR SAFETY EVALUATION CONCLUSIONS ,.
0.1 Systems Subsystems. Comoonents Affected ,
~
0.1.1 Inertina and Devwell Testina 0.1.1.1 This modification adds a check valve and a 2 inch tee in the existing cryogenic nitrogen supply header common to the drywell instrument supply and.drywell inerting supply. The 2 inch tee connection branch includes a normally closed globe valve and a normally capped pipe end. 0.1.1.2 This modification adds a pressure control valve to the existing pipe which connects tne drywell instrument header with the existing drywell inerting piping which is normally isolated by an existing closed valve. The pressure control valve reduces the 125 psig (maximum) supply from the existing cryogenic storage tank to 70 psig which is the normal drywell inerting system supply pressure. 0.1.1.3 This modification adds a nitrogen cylinder station connected to the existing drywell instrument supply = header. The system is initiated when the normal nitrogen supply line pressure drops below the bottle s.tation supply pressure. pyg D.1.1.4 This modification locks closed valve FH0-162/in the instrument air supply header to drywell instrumer.tation. this decreases the load on the existing instrument air system. Drywell instrumentation supply will be normally taken from the existing nitrogen system and supplemented by the new nitrogen cylinders and liquid nitrogen /vaporizertrailer. 0.1.1.5 This modification adds a 11guld nitrogen / vaporizer trailer on site which will be connected to the existing drywell inerting supply piping mounted on the outside of the reactor building wall and to the drywell instrumentation supply header via newly installed 2 inch piping when the existing liquid nitrogen storage tank is not aveilable. 0.1.1.6 This modiffr.ation enanges the air operator on drywell - instrument supply isolation valve A0-4356 from fail closed to fail oper.. This ensures an open path to the drywell during a station blackout event . Check valve 31-CX-167 located downstream of this valve performs the cor.tainrient isolation function. 0.1.1.7 This change modifies two hangers in the seismic portion of the drywell iner ting heade.r. This is a result of a nw analysis performed on this pipe line.
N : T K p ; E L T m m M.Jaa 'w nn,' %:T :cui;;.,n;Ol5 W m Rc2;3 5;.w; ' Attachment 11 F Safety Evaluation, Rev. 0- .TE. O, . Alle Sheet (of .to D.I.2 Reactor Buildina . This modification involves drilling holes in safety q
. related reinforced concrete of the reactor building.
D.I.3 control Room Annunciator This modification involves adding an additional alarm ing : to the existing annunciator to provide control room indication upon switchover to the nitrogen cylinders. 0.2 Safety Functions of Affected Systems /Comoonents
- 0. 2. I' Inertina and Drywell Testina D.2.1.1 The inerting and Drywell Testing System is part of the Containment Atmos >heric Contreii System (CACS), which provides the capa)llity to purge containment so that the containment design pressure is not exceeded following a design basis accident.
D.2.1.2 The Inerting and Drywell Testing System provides the capability to pressurize containment to dilute and maintain the hydrogen concentration below the lower flammability limit following a design basis accident. D.2.1.3 The Inerting and Drywell Testing System provides the capability to close and thereby isolate the Torus and Drywell Purge and Makeup penetrations and the Orywell compressed air header penetration to satisfy the containment isolation function following a design basis accident. D.2.1.4 The modified supports are required to maintain the pressure boundary integrity of the drywell inerting system during normal and accident conditions. D.2.2 ggytorBulbiq.q The reactor building is part of the secondary containment system. The secondary containment system, in conjunction with other engineered safeguards limits the release of radioactive materials to the environs, so that offsite dose; from a postulated design basis accident will la - e.aintained below the values of 10CFR100. D.2.3 Control Roq_m ar.nunciator i The control room annunciator has no safoty function.
7 u- - - _ ., .; m m . , , - a.n c- .m. m._ w . a. , ., - . ; u.. _ _ 2 , .m . i
~
Attachment 11 Safety Evaluation, Rev. O SS.S afl$ Sheet 6 of Ao D.3 Effect on Safety Function - D.3.1 Inertina and Drywell Testino D.3.1.1 - The backup nitrogen supplies provided by this modification do not reduce the Inerting and Drywell Testing System's ability to purge containment. D.3.1.3 This modification does not affect the containment isolation functions of the Inerting and Drywell Testing System. D.3.1.4 The modified supports have been reanalyzed and redesigned to accommodate thermal and seismic loads. This modification therefore enhances the systems ability to perform it's safety functions. D.3.2 Reactor Buildina The safety function of the reactor building is not adversely affected by the additional penetration in the reactor building wall. D.3.3 Control Room Annunciator The control room annunciator is not adversely affected by the addition of a new alarn input. D.4 Analysis of Effects on Safety Functions O.4.1 Inertina and Drvwell Testina - D.4.1.1 This modification does not alter the Inerting and Drywell Testing System's purging function becture the liquid nitrogen / vaporizer trailer is normally isolated and is provided to supplement the existing nitrogen supply, when needed. D.4.1.2 This modification does not alter the Inerting and Drywell stem's containment oxyra.a dilution function Testing because S{e liquid nitrogen / vaporizer trailer is normally t., isolated and is I,rovided to supplement the extr, ting nitroger, supply, when needed. ; 0.4.1.3 This modification does not alter the Inerting and Drywell Testing System's containment isolation function because the liquid nitrogen / vaporizer trailer and nitroger, cylinders are connected upstream of the existing containment isolation valves. Containment isolation valve function / operation is not altered by this modification. , i D.4.1.4 The hanger modifications ensure the systems ability to withstand seismic and thermal loads.
m . . m w.x.w i.. . . n.. ..w.:w.,;;a. x. u:a.a . : . , , ;rx , . . ,,.; v w ac . gg ;-- s . ......m .. f Attachment 11 Safety Evaluation, Rev. 0 SE .de all !6 Sheet 7 of .ao 0.4.2 Reactor Buildina ~ (A a r. a ec..ere a C/ c. 8 6 50 - A ) The reactor building penetration added by this modification will be made utilizing existing engineering and' construction precedures. Engineering has evaluated the size and location of the core drill necessary and the penetration seal required to maintain the structural and pressure boundary integrity of secondary containment. The addition of the 4 inch diameter coredrill in the reactor building wall, followed by the p',stulated failure of the 2 inch nitrogen cylinder station supply line during a seismic event has been reviewed. The additional vent area created has been judged to have a negligible affect on the. capabilit icar.draq'42.n to do secondary containment.
% s Mawks 54resnes ygg iew A sacemane urainassarr zs,bADed"ddo7A/NsD 0.4.3 Control koIm TnYu $ fa D '" " /d#W* #
- A ,
The control room annunciator is not adversely affected by this modification, as such no analysis is needed. E.
SUMMARY
E.1 The safety functions of the Inerting and Drywell Testing System . are not adversely affected by the additional nitrogen backup supplies and associated piping, valves and instruments. E.2 The safety functions of the Reactor Building are not adversely affected by the addition of the reactor building penetration. E.3 The safety functions of the primary containment are not adversely affected by the additional nitrogen backup supplie: and associated piping, valves, and instruments.
)
l 1 L 4 l i
c x .. a s.. x .u ,, : :, n . c ..
. . . . ~ . . +,:. : .niw . '..r .. x:n : ~ ~ . 3:i. ... .:.:.wc":- .m73wim > ..
l Safety Evaluation, No. alllB Rev. 8 Sheet g of JLe SAFETY EVALUATION PILGRIM NUCLEAR POWER STATION A. APPROVAL (X) This proposed change does not involve a change in the Technical Specifications. (p This proposed change, test or experiment does ( ) does not (X) involve an unreviewed safety question as defined in 10CFR, Part 50.59(a)(2). (X) This proposed change involves a change to the FSAR per 10CFR 50.71(e) and is reportable under 10CFR 50.59(b). ( ) Coments: The safety evaluation basis and conclusion is: (4) Approved () Not Approved 60Vl ANB] Discipline Group Leader /Date Supporting Discipline Group Leader /Date B. REVIEW / APPROVAL (7) Coments: N Ykw Q})proup6eader/Datd A0 9' C. ORC REVIEW ( ) This proposed change involves an unreviewed safety question and a request for authoril:ation of this change must be filed with the Directorate of Licensing, NRC prior to 1. implementation. ( ) This proposed change does not involve an unreviewed safety question. ORC Chairman - Date __ ORC Mteting Number _ _ _ _ CC*
.s Fvhih4 '. 7. A (9haat 2 of 3) Rev. 3.
7.s .- .: ,::. . .aw a :a. r , .1 i;. ;.e c....;.9 : s:.: n -, .. o ,. . ; z ..;~
. u, v x. <, :;. ;.; :s ; :-- 7.w p v; - .-- -
Safety Evaluation, No. ollf o 4.o Sheet 9 of sta PILGRIM STATIQ3 , FSAR REVIEW SHEET
References:
~
Support a change List FSAR test, diagrams, and indices affected by this change and corresponding FSAR revision. Affected FSAR Revision to affected FSAR Section is shown on: Section Preliminary Final Fio.10.11-1 (P4ID) Attachment 1 / Fio. 5.4-1 (PATP) Attachment 2 ! Sec. O. IL 3. l Attachment 3 sed. N ol Attachment 4 Attachment 5 Attachment 6 PRELIMINARY FSAR REVISION (to be completed at time of Safety Evaluation preparation). Prepared by: +' /Date: Y 87Reviewedby:M d N /Date: Y987 Approved by: / /Date: 4'7M FINAL FSAR REVISION (Prepared following operational turnover of related systems structures of components for use at PNPS). (1) Prepared by: /Date: _ Reviewed by: /Date: (1) Attach comp 7eted FSAR Change Request Form (Refer to NDP).
.__m___. _ _.
~
" ' '** ? **** #""?+
.4 S Den i
,, s _ - ,. . . . , . . , - . ,, , ,, s. ,
i l CH 10 Ch Jne Safety Evaluation No.: a1 ll A SAFETY EVALUATION WORK SHEET Rev. No. 4_ A. System Structure Component Failure and Consequence Analyses.
.Srf. I of3 ..
System Structure Conconent Failure Modes Effects of Failure Cosaments j, bg c'~fD, LsL. ~TErtR Cernd MPE CRACbOG "NE ,YEMP l ',.072)a-valve. FonLS ~Tb 8iGO8D N. A V'Al v't ~ t8LS CL.osep PResu et ?EG. Fats S nrierPArssunt2E YY/AIfr IDWascentar. VALVE 5 whLL Watks' p:-e vt*DY ovewasssam enria G Two rxMsv4L CacfdA-VALVES PgoVtDEP
- a. AL Wehuous Fau. ova sam a ais.us ssmerso y een bLD6s DRStheaED Fiiut.
MISSLE IMPtNGE#its!T~ \ (CoMT~ludEO) General Reference Material Review
~
FSAg CALCULATIONS RESULATORY SECTION PNPS TECHNICAL $PECS. DESIGN SPECS PROCEDURES SUIDES STANDARDS CODES AfM A gg, jo // 3.7, A u &rcHTE L CALC . *1~2 8 dib oTd' , 66c.62l+5",A1'CS.A M-6ro -/ 6E SIL **.q'd,Z p;&g,J./ 4, GE EMc otoCAlat/to msv. o s ec . +.9 . A. Ma,s.4-/ Mir./D.//~l S. For the proposed hardware change, identify the failure andes that are likely for the components consistent with FSAR assumptises. For eer,h failure mode, show the consequences to the systaa, structures or related components. Especially show how the failure (s) affects the asstyw) sefety basis (FSAR Text for each systse) or plant safety functices f5AI Crwipter 14 and Appendix S). Prepared by _. tate / 7 Yf fIt is a requirement - WOTE: to include this work sheet with the St.fety Evaluation. I Exhibit 3.07-C Rev. 2 ___ __= _________- _ _ _ _ _ _ - l
n y .=.-::X 7;g ,,2 m.;i=.a.a a.. : : : ;~ ; a ;.. , % : ::iz,r:
;;i ,': ~ > ,x.: . . : c .: : '
lh::,1 , .: . SN ll of k Safety Evaluation No.: .71/8 SAFETY EVALUATION WORK SHEET Rev. No, d CoGTIcoAi nod ($ hah 3)
- A. System Structure Component Failure and Consequence Analyses. ,
System Structure -Component Failure modes Effects of Failure Comments N2 0t'llAfD EiR $ AGbsdRL W ov5MPfassed5-no. kEP' VAAME l FALs Pt Pt M6 oram 1E4K 7bM Als Ala Avn' lam.E. O LeML /Awc.AT'/Al6 l
"*"rs Ano Bler-ss unz PAcA: AGE TRAsuik JAIDitM 770A)
,A 4 Ecn~rl.Er,
@RKDUAiD* FAST *kl. Wb ~
AvraM4 rte SultfcAd/KA gneum. AdrincrMFAlf" 45"&rrLES ' fpNa-rard/4## A5 Al.1.DxA ve i</ro Ain isi4cr p eutrat-mM AoM mnrgo MMoc. fCoCTIM05.p) /60 Fr'A o Alor A I gMgK,j %$$ seneral Reference MateriahReview g
~
AAID EMPECTED kota AMTL CALCUt.ATIONS REtulATORY prAs' g 4 g/J/ F5AR SECTION PNPS TECHNICAL SPECS. DESIGN SPECS PROCEDURES GUIDES STANDARDS CODES D. For the proposed harthmare change, identify the failure modes that are likely for the components consistent with FSAR assumptions. For sech failure mode, show the consequences to the system, structures or related components. Especially snow how the failure (s) affects the auigned safety basis (FSAR Test for each system) or Cant safety functions F5AR Chapter 14 and Appendix G). Prepared by / M Date , 7O j // I ~ NOTE: It'is a requirement to include this work sheet with the Safety Eval uatir,n. l Exhibit 3,07-C Rev. 2 l l i
- ____m_______.________._______
3 m7., , . ;s .;---- -- - y.c :, , s , , ., - , .. ,,, _ ac u,- - - i 6N IAefGo Safety Evaluation No.: o2 / / 8 SAFETY ivALUAT10N WORK SHEET Rev. No. O
- 04) r*w v.4 nosi (5 H.3f.3}
- A. Systern Structure Component Failure and Consequence Analyses. .
System Structure Component Failure Modes Effects of Failure Comment s 6t)PPo / T*S bJ A% Krs P PF_ Us)SCPPetTED $JWo.<0 /9A6 3 F% w. g$nsxicALL ' JA3 /6AIED )TD, fXs es.uDE R4/4W16-General Reference Material Revit y CALCULATIONS REGULATORY FSAR SECTION PNPS TECHNICAL SPECS. DESIGN SPECS PROCEDURES GUIDES STANDAR05 COD
- 8. For the proposed hardware change, identify the failure modes that are likely for the components consistent with FSAR assumptions. For each l failure mode, show the consequences to the system, structures or related i components. Especially show how the failure (s) affects the assigned I safety basis (FSAR Text for each system) or plant safety functions FSAR !
Chapter 14 arAl Appendix G). Prepared by fU , Date (( h M -
)
It is a requirement to include this work sheet with the Safety l NOTE: l Evaluation. l Exhibit 3.07-C Rev. 2 l l I I
"e , . ; V . ., u e, ; Gi s A .s : % . u t . .J T &. .: C k :.': ,.sk.c- ~s & : ,:.
- L .' o "l~. 'i. L T.. h .' . . % V *a r n. -. l
.6N lb oh alb FNPS-FSAR Sg". Ale /6s4 o i dil8 l 10CTR50.44(g) requires that the Containment Purge and
- m. Repressurization Systems (CFRS) conform to the general requirements of 10CFR50, Appendix A. Criteria 41, 42, an 43.
- The CACS in conjunction with the SGTS are the systems which Pilgrim .
J Station utilizes for primary containment atmospheric control as j required by 10CTR50.44. See Figure 5.4-1. ] The Containment Combustible Gas Control System is used primarily for purging (i.e., inerting) with N: or can .be used for containment { venting if nitrogen is not available. Exhaust from both the Torus I and Drywell can be routed co the main stack via the redundant trains of the SGTS. Makeup of nitrogen (or air) is supplied via the 1 in redundant solenoid valve trains. See Figure 5.4-1. gg }] g l Oxygen concentration is controlled below flammability limits (5 volume percent) by a feed and bleed method (purge method as - 6efined in 10CFR50.44). The time required before initiation of purge (vent) of the primary containment is controlled by repressurication techniques consisting of nitrogen (or air) addition to the primary containment. This repressurization will be initiated within 8 hr following a postulated LOCA. Calculations show that the primary
~
containment oxygen concentration will not reach 4 volume percent (Tech. Spec. limit) after a postulated LOCA based on nitrogen (or air) addition and restoring and controlling primary containment pressure within the range of 22 to 28 psig by venting. p To meet the above requirements, 16 solenoid valves are arranged to e- provide redundant paths to and from tne drywell and torus for Na
~
makeup /repressurization and' venting. Na makeup /repressurization is provided by:
- 1. Connecting, to hose connect. ions outside containment, a portable nitrogen supply via truck with vaporiser or using the existing (non-seismic) nitrogen storage tank with vaporizer (requires opening a manual block valve located outside containment in the yard area Reactor Building north wall) (primary emergency make-up) l
- 2. Alternatively, provide a compressed air supply from service air connections outside the primary and secondary containment or from portable (gasolins driven) air compressors located on site (secondary emargency make-up) l The solenoid valves are designed to remain closed against maximum containment pressure, to vent containment so that the maximum containment pressure will not be exceeded, and to provide a nitrogen flow sufficient to maintain the hydtogen concentration inside -
containment beltu the flammability limits. 5.4-2 Revision 5 - July 1985
7::_;=w .m:. v;;m ::ama u m: .. ...+ . . ; w. u;.; .s.: e .= m.w:.wm r. . a,a .. 3
^
3NI4ofslb 6E A3o.4//B Rev.o a
$$$5 A.1 4 r S A liquid nitrogen / vaporizer trailer (X-168), and associated piping and valves. The trailer will be normally stored in the northwest corner of the plant site in a storage /laydown area. When needed, the trailer will be moved (via an on-site truck cab) outside the west gate of the condensate storage tank controlled access area. It is designed to deliver at two different pressures to the nitrogen systes via two flex hoses, which will be stored on the trailer.
Nitrogen at 120 psig will be available from the nitrogen , trailer to match the existing drywell instrumentation sup>1y i pressure. Two inch supply piping and a globe valve will >e l added to tie the liquid nitrogen / vaporizer trailer into the i existing nitrogen system. A check valve will be added to prevent nitrogen flow into the existing liquid nitrogen storage tank. Nitrogen at 70 psig will be available from the trailer to match the existing drywell supply pressure. This will be accomplished by connecting to the existing fill connection located on the north wall of the reactor building. A.2 -Two banks of ten cylinders each, a cylinder rack and manifold (X-169), associated piping and valves. The cylinders will be arranged to automatically maintain the nitrogen supply to drywell instrumentation once the existing nitrogen supply is not available, until the new liquid nitrogen / vaporizer trailer is available. The cylinders will deliver nitrogen gas at 110 psig (controlled by a vendor supplied pressure regulator), through 2 inch piping which will tie into the existing drywell instrument supply header. The new piping will contain a check valve, a gate valve and a relief valve. A differential pressure indication switch with annunciator will be connected between the cylinder supply and the existing supply to provide control room indication of switchover to the cylinders. The existing manual gate valve ,s 4t-1, ' l' (3?, HO-162) will be changed from normally open to normally locked closed to isolate the instruent air supply to drywell instrumentation. Existing drywell instrument eupply isalation valve A0-4356 will be modified from fail closed to fail open to ] 1 maintain an open nitrogen path tc dcyttell instrumentation following a loss of power. _ j 1
cr. , .a . ;;.:..:. .. .: .e,.. ..a .u. .. . .; x = ~ . n. : g ' : W ~ ~ e. ' ." zw .. ;': . ., ' SH ll of eta l 3L h%dyBAmJ e k PNPS-PSAR ( l j
.c 10.11 INSTRUMENT AND SERVICE AIR SYSTEMS
( 10.11.1 Power Generation Cbjective The power generation ob]ective of the Instrument and Service Air Systems is to provide the station with a continuous supply of oil-free compressed air. This air is directed to station-instrumentation and general station services. 10.11.2 Power Generation Design Basis
- 1. The Instrument Air System is designed to supply clean. dry air to station instrumentation and controls at 70 to 100 psig with a design dowpoint of -40*F at 100 psig.
- 2. The Service Air System is designed to provide clean air to station services at 70 to 100 peig. The Low Pressure Service Air System is designed to provide clean . air at a nominal pressure of 20 psig to station services.
10.11.3 Description 10.11.3.1 General The air systems are, in general. designed to Class II requirements. p although Class I equipment requiring air under accident conditions has Class I air accumulators and piping associated with that equipment. See Figure 10.11-1. The high pressure air supply (nominal 100 psig with allowance for drops to 90 psig) is developed by three reciprocating and two rotary screw type air compressors operating in parallel. Each compressor has an after cooler and delivers the compressed air to a bank of receivers. There are five air receivers which are connected to a l common discharge header that delivers the air to two instrument air dryers to provide high quality dry air to the various instrument air headers. There is one instrument air filter located upstream of each instrument air dryer. There are three instrument air filters mounted in parallal downstream of the instrument air dryer X-105A and one filter downstream of dryer X-105B. the downstream air filters are to ensure that no dessicant or other foreign material enters the instrument mir system. There is also a bypass around the dryors and filters which can be opened by remote manual means for dryer X-105A ar.d manual only for X-105E :.o assure a continued supply of instrument air to the essential instrument air header in the cvent, of an air dryer failure. Normally, use of the two rotary compressors vill ' maintain the air receivers at f.hv doaired pressure for system supply. The remaining cov9ressors serve as standby units. Actuation of the f.tandby units is autoratic and is indicated in the control room. The low pressure air supply (nceinal 20 psig! is developed by two centrifugal air blowers. The blowers discharge for distribution 10.11-1 Revision 4 - July 1984
v <
- n. : ..L. . w. .- . , x. . :. - .
.c- _ : :,-- , - ,v m r c. u . ~ . n;nm s
a n. af .a PNPS-PSAR gg M, Jj l $ gev 6 through a wisture separator and a mist eliminator. Blower usage is intermittent *,. No dowpoint control is provided. A normally closed pressure reducing cross-over line is provided " between the high pressure distribution header upstream of the air dryers and the low pressure distribution header. This cross-over may be used to continue low pressure service in the event of blower failure. Pressure loss in the high pressure system, sensed by several pressure switches, will cause valves in the service air headcr. the low pressure service air cross-around line, and the non-essential instrument air header to close in a cascading sequence thus leaving the essential instrument air header as the only header drawing air h W from the receivers in the event that supply pressure decreases. 1E11.3.2 Equipment Description ng# Compressors The three reciprocating air compressors are vertical, single stage, compressors. They are each rated to
~ double deliver acting reciprocating '/ min at 105 psig.
159.5 standard it The two rotary compressors are each rated to deliver 655 standard it'/ min at 102 psig. Each reciprocating compressor has a pressurized lubrication system for the power-end parts. The cylinders are non-lubricated having Teflon piston rings. They also have water cooled cylinders and have ' a displacement of 261 in'. 'All intake valves have pneumatic operators which depress the valves allowing the cylinder to unload by venting to the atmosphere each time the motor starts and each time the receiver pressure reaches the top of its operating rarwje. Each of the three reciproct,tish; compressors is belt-driven (4 belts) by a 40 hp dripproof induction motor. The compressor speed is 514 rpm. The two rotary screw type compressors are direct driven by an electric motor which provides a shaft output of 156 hp at a coepressor discharge pressure of 102 psig. The compressor speed as 3,550 rpm. Aftercoolers i The compressor aftercoolers are shell and tube counter current f coolers with air passing through the tubes and water flowing around I l the tubes. They have an integral moisture separator equipped with an i autoutic drain trap to remove condensed moisture from the cooled l i air. Cooling water is supplied by the Turbine Building Closed I I ! Ccoling Water System. 7 'D C 10.11-2 Revision 4 - July 1984
c..n. .w :.: /c s : .. . :v a ;.c. a n:: .7 ~
. .n . R. 1;.O a! a : WEE ~ OK.}
SH s 7 of .ta ' w 18 SG Me ajjp him/. O , j
$Nb5U 1
l a.e -%r banks of ten cylinders ME~a 61I5ier rack and manifold (X-169), associated piping and valves. The cylinders will be arranged to automatically maintain the nitrogen supply to drywell instrumentation once the axisting nitrogen supply is not available, until the new 11guld nitrogen / vaporizer trailer is available. The cylinders will deliver nitrogen gas at 110 psig (controlled by a vendor supplied pressure regulator), through 2 inch piping which will tie into the existing drywell instrument supply header. The new piping will contain a check valve, a gate valve and a relief valve. A differential pressure indication switch with annunciator will be connected between the cylinder supply and the existing supply to provide control room indication of switchover to the cylinders. The existing manual gate valve,s-at-1, (31-HO-162) will be changed from normally open to normally locked closed to isolate the instrument air supply to drywell instrumentation. Existing drywell instrument supply isolation valve A0-4356 will be modified from fail closed to fall open to maintain an open nitrogen path to drywsil instrumentation following a lens of power. . ese
--.___._.__m_ _ ____________m_.
7;~ ;.,.,,. ' :-' ' 7;~::Q, ,;[;L'5f . ,.,.q * -
' ' ;C;;',':
' ;g
,. ,q';.;
, 1; : u,.' ,.. ,
,,_.y,,,,
s> L Ne a110 M ** U M
%: 4 N, i,
%n Nuu I.
>"f
==c
.M esa aws.am
.g_ . . . . .... . . . . _.. .L . .. .
m n.2 - ( .- ., z z o
. __ i -
I y . .. l m i qM 1 _ MD _ Sw.
-, n - .
, .c =# i= T
===i = 84 TT6 T j
&. - .. *-M
~
~ ** ,,g qeg q wm np A g,% . .
. l g^ j[$.
mr M , w sy p t . . ...
* /5 6 a / m ...".J $/
I) l m3r< *M
, f* *
* ;: : _ " & ~ -
'" ' 'ggsc vf"
.3 }A "--y m
+ a .
~~
p g. + m,/ - 50 p.,s.@ g a 'f m, .. - }2 Is 8 v,. g + m
== .-
,. - .. _ . . . . .. . ~ . - . . = . ~ ~ ~ - -
, SM l9 *J .30 SE 4. .IIL4 feif o 7
't 4i
,(r 7
1x m
! lgh : : : ::s w
2H l
'l--42% Q k c !$5
"$1"e" = = s g,.
n, su - e - e- dl F@.p
> = p M @. , 9"
?,"n(4 . 2 .%%.9
.. . -- -. I ,
..- m - u.y , ;a p@-g g M k .? :
6 P S Ii.,,l] .: t, a
..s.:
e ..-2.: . a esa
.j
% > a.
g ___ <_. ________ __
,A y- - t ga%v == n a r-1;; g ** l a j g
l\ Q gjQ[d!h a.- . sw om s - af kN 5 g q c' m
$b y _ __ __
(coas As osmia ss3ww.) li 8 l i l _-- .J
u ., ._: ., , ;a: ...; ~ .:. e .a. ; . \ . > . = x .. : c Y. ' - :O: ' a i u :. %..M'3 - L, :.: , .: , ' V: e. ' , . . - -
. SN ato of .to se a.au w.o
; o;h5 L
' _ -i.
. 4 k
.p.;-
i- 1- 'l IN 4 llC l = " l% l m ] giHb ~ h ( 9' Me{lgitJ Wly B1SNI IVuM3553 NON ' s-w W*
' """*Ag n_ g j- 7 - .- W - -n.< l ea -8 m tlY*b1SNI j es ,L y y_ Adbgo 'kr Qjg,!"'[g 4, y4e n =r a..m *w: xa e -
i a a ga bb! h)!_paar
~
1 1 I !3, . #, aim.nugh.fQ
= Ij@8*i g ;
/[t-i
~
,l na d!ts i 9
~~ -
.s- .
i
- gl; C'
p+(ggg_1p4Oh
\liht Ij FA d
- F ca u[%e g[' 1, ., a
v-- = . ::.z.a. a , - .. = . _ .. x . . , , ,
' Safety Evaluatien HE.D Proposed Change No.: 2.1 y RAFETY EVALUATION PILGRIM NUCLEAR PW ER STATION l
Rev. No. PDC PCN System Calc. Initiator: Dent: Group: No.: Name: No.: Date: B1 ut J. Coughlin NED P.S. 86-56B -- 5/18/87 Add Blackout Diesel Generator Description
+o provide alternate of Proposed non-sa etychanke, related sourcetest or exper1Nnt:
of 4KV power to emernency buses A5 / A6 SAFETY EVALUATION CONCLUSIONS: The proposed change, test or experiment: ' I
- 1. M Does Not ( ) Does increase the probability of occurrence or consequences of an accident or malfunction of equipment important to .
I' safety previously evaluated in the FSAR.
- 2. M Does Not ( ) Doe's create the possibility for accident or malfunction of a different type than any evaluated previously in the FSAR.
~
- 3. QQ Does Not ( ) Does reduce the margin of safety as defined in the basis for any technical specification.
BASIS FOR SAFETY EVAL 11ATION CDNCLtfSIONS: Ser ATTACHMENT NO. 1 Change Change M Recossended ( ) Not Recommended SE Performed by ~ ten Date f//A/67 ' (/ 0 Exhibit 3.07-A Sheat 1 of 3 ISSUED FOR I 3.07-13 CONSTRUGfiON J
. ; = _. -
[
nT;Ian; ;5 r;w. : u. . . . .. .. ~. ,1. . Ac. _. .crw . x - , .. . ::r . a batety tvaluatson . SAFETY EVALUATION
- PILGRIM NUCLEAR PGiER STATION Rev. No.
A. APPROVAL (X) This proposed change does not involve a change in the Technical Specifications. ( ) This proposed change, test or experiment does ( ) does not M involve an unreviewed safety question as defined in 10CFR. Part 50.59(a)(2). OC This proposed change involves a change to the FSAR per 10CFR 50.71(e) and is reportable under 10CFR50.59(b). ( ) Comments: The safety evaluation basis and conclusion is: , 99 Approved () Not Approved
/ M/Y!87 M/4 Dis Wie Gro' up Leader /Date Supporting Discipline Group Leader /Date
- 8. PEVIEW APPROVAL
~
(d Comments: I*-- 8 $ US&SA lWoup Lead 6r/Date l C. ORC REVIEM f/ lt1 . () This proposed change involves an unreviewed safety question and a request for authorization of this change must be filed with he Directorate of Licensing, NRC prior to laplementation. (, This proposed change does not involve an unreviewed safety question. ORC Chairman Date Moo /P7 ORC Meeting Number P 7-Il f i Cc: Exhibit 3.07-A Sheet 2 of 3 ISSUED FOR ! I 3 m-" w CONSTRUCRON - - ,_ ,. ..- i
e m m m m . w.=,i .:w w.:. .. x ; u : ~ ,x . ,s. . . . . . . xn ; 5/18/87 M & ch. 4 [ , Safety Evaluation N i Sheet 1 of i l l l Basis for Safety Evaluation
Conclusions:
A. Change Description This design change installs a blackout diesel generator set as a non-safety related source of 2000 kw continuous (2200 kw standby), 3-phase, 60HZ, 4160V electrical power for a minimum of one week without re fueling. Electrical power from this blackout unit will be utilized to operate one ECCS pump and all other associated loads from one safety train required for a reactor scram without a 1.0CA, as defined in the FSAR, Table 8.5-1. The unit is mounted on a skid and housed in a pre-engineered enclosure to protect it and the associated electrical and [p mechanical support systems from the environment. The unit will be la located south of the plant) adjacent to the switchyard re &y house.,. 4n.,f f,,y [ p en N. ex.fth7 FMv4 skw The diesel generator can,be connected to the existino essential service g ' 4.16kV buses A5, A6 .
,The di~esel gefierator will be through a new two-breaker 4.16kV bus A8.
connected to one of the 4.16kV breakers of the new switchgear A8 The and the shutdown transformer will be connected to the second breaker. outgoing feed from the switchgear A8 will be connected to the existing 4.16kV breaker A600 which is in turn connected to breakers A501 and A601 of the essential buses A5 and A6. Breaker A802 which is connected to the shutdown transformer will be kept closed during normal operation to supply power through breaker A600 This (normally closed) to breakers A501 and A601 (normally open). alignment of breakers is consistent with the present arrangement which msintains shutdown transformer power available for automatic connection to the emergency buses (via automatic closing of A501 and A601) upon a unit trip, loss of the start-up transformer and failure of the emergency diesel generator. The blackout diesel gererator cutput breaker A801 will be maintained open during normal operation and will be closed to the Provisions safety related buses only during station blackout or test. are made so that the blackout diesel generator can be load tested by synchronizing to the shutdown transformer during normal plant operation. During this time breakers A802, AE00, A501 and A601 are maintained in their normal line up. When the control switch is selected to the test position and buses are synchronized, the shutdown transformerW aA hu m and . diesel generator are both svailable to supply men P +b A synchro-check relay, a synchronizing : r located at the A8 switchgear/ diesel gen Ewitch Rc p.ffh)hhstF sac ha e, e fsp}y fic . for ) load testing. l. cad testing can only be a-a Interlocks arm - m WOL;M1 ?( __ N s e sea _.) l enclosure /switchgear AB. i Eeibit 3.07-A l l
-w . ,, . , ;;; - w ... ,- v.c . , . ~ ,w rm t
5-15 87 A & e4 N I
- Safety Evaluation 24 44 '
Sheet 2 of 9 generator breaker is closed to a dead bus (i.e. the shutdown transformer 1 is de-energized) the shutdown transformer breaker A802 will be prevented { from closing until the diesel generator breaker A801 is tripped. The blackout diesel generator and the two new 4.16kV breakers will be j operated manually (except for automatic protective actions) from the main l control room or locally from the diesel generator. When the diesel l generator is used during blackout conditions, load will be controlled by i manual operation by disabling of breakers on buses A5 and A6. In general either A501 or A601 and the ECCS pump breakers on the bus being energized must be opened and prevented from closing. Control switches and indicating lights are installed on main control room panel C3 to allow start /stop of the diesel generator and trip /close of the A8 bus breakers. Engine speed control and generator voltage control is not presently provided in the control room. These controls located at switchgear A8 must be set / adjusted during unit test. Once set they will operate automatically when the unit is started. The diesel generator breaker can be closed from the main control room only when the shutdown transformer breaker is open. During a blackout the operator must trip the shutdown transformer breaker before closing the diesel generator breaker. Relaying on the blackout diesel generator and an engine speed , - interlock provide permissives to close the diesel generator breaker to bus A8. The shutdown trancfarmar breaker can be tripped and closed from the control room. The diesel generator is fully self-contained and is not dependent on any permanent plant systems, except for a 480 volt AC feed from existing non-safety related load center B4 of PNPS. This feed will be connected to a motor control center for supplying power to the maintenance loads (engine water heating, lube oil system, heating / ventilation, lighting, etc.) of the diesel generator when it is not running. If the diesel i generator is not operating, loss of power to the maintenance MCC will be alarmed (diesel generator trcuble) in the main control room. If the diesel is operating, an automatic transfer switch is provided to supply power to this motor control center from the blackout diesel generator upon loss of the station supply. I The blackout diesel generator and the new two breaker 4.16kv bus will be supplied 125vdc power from a new independent battery system (battery and charger) dedicated for this purpose. This de power supply installed in { the A8 switchgear is not connected to the existing plant de power system. ! Annunciation will be provided in the main control room for blackout I diesel generator trouble, diesel generator breaker (A801) l j 1 trip / inoperative and shutdewn transformer mma (/E2)
- l trip / inoperative. Annunciator window 25 >n paneMtgrp @ %ly s l breakers A501 and A601. Breakers A600 aM A802 WVW x #in i l these breakers to annunciate on automatic t QIge ry y
any of these four breakers. The blackoutg d .. it){ e . Exhibit 3.07-A
m ,;; ;. m a;.7 , u . ..- .y . _., , ., c . x w, , 5-15-87 Mec.k M / Safety Evaluation 2 3 44
- Sheet 3 of 9 l
breaker A801 will be annunciated on window 69 of panel C3. The window 69 at present annunciated Unit 1 emergency shutdown transformer untier voltage relay or breaker A600 open. This under voltage signal will be combined with signals connected to annunciator window 47, " emerge *ncy shutdown transformer mist, alarm". After these modifications, window 25 will annunciate " Shutdown Trans Bkr Trip or Inop." This will indicate the power supply from shutdown transformer to the emergency service buses A5 or A6 is affected. Window 47 will annunciate " emergency shutdown transf. misc alarm". This will indicate trouble with the shutdown transformer and under voltage on breaker A600. Window 69 will annunciate " Blackout Diesel Gen Trouble". This will indicate trouble on the blackout diesel generator and breaker A801. l Protective relaying is provided to prevent damage to the diesel generator and the shutdown transformer. Voltage regulation and engine governor control are provided to maintain acceptable engine speed and generator voltage and frequency during diesel generator operation. Overvoltage, underfrequency and overfrequency relaying are provided to prevent the diesel generator from suppling unacceptable power to the 23kv electrical system or the emergency service buses. A two element differential current relay used to protect the shutdown transformer up to breaker A600 will be replaced by a 3 element relay. The relay mounted in control room panel C5 will allow the' addition of the blackout diesel generator to the existing differential protection scheme. B. Purcose of the Chance A diesel generator will be installed at PNPS to provide a backup source of power to safety related buses A5 and A6. This backup power supply is being installed to reduce the probability of a station blackout which could lead to core damage and/or containment failure. In addition, it is expected that installation of this unit will allow PNPS to retain its 4 hour coping requirement under the proposed NRC blackout rule instead of being forced to upgrade to an 8 hour coping requirement. 7 \ ISSUED FOR l i CONSTRUCTION _=_ ._.,_ y
.e_.
.a s. , ..a.a .n : 1 2 ,, . ~ .w
. . f. ~ :. , . c. r:,. ~ , . , : :- .-
5-15-87 Mach.*/ Safety Evaluation 214( Sheet 4 of 9 C. systems. subsystems. comoonents Arrected This modification involves the addition of non-safety related components. The following systems are affected by the addition of the blackout diesel generator set.
- 1) The secondary off-site ac power source (shutdown transformer).
- 2) Emergency service buses A5 and A6.
- 3) Main control room panels C3 and C5.
- 4) 480V load center B4.
- 1. Secondary ac power source (shutdown transformer):
The secondary ac power source (shutdown transformer) is discussed in Section 8.3 of FSAR. The secondary ac power source (shutdown transformer) is a non-safety related system and provides a source of power to the emergency service buses A5 and A6 of the Auxiliary Power Distribution System. The new blackout diesel generator will be connected between the secondary side of the shutdown transformer and emergency buses A5 and A6. A new two breaker 4.16KV bus will be installed to which the new diesel generator, the shutdown transformer and the existing 4.16kV breaker A600 will be connected. The differential current relay installed to protect the shutdown transformer will be modified from a 2 element unit to a 3 element unit to allow the addition of the diesel generator in the differential protection scheme.
- 2. Imergency Service Buses A5 and A6: , ,
The 4.16kV emergency service buses are described in Section 8.4, Auxiliary Power Distribution System of the FSAR. The emergency buses AS and A6 will be supplied power from the new diesel generator under station blackout conditions. There is no l physical change to the emergency buses except that existing contacts from the differential relay discussed in A.1 above in the control schemes of circuit breakers A501, A600 and A601 will now include input from the blackout diesel generator circuit. The differential relay is non-safety related and will operate for faults when power is supplied to the emergency buses from the shutdown transformer or the blackout diesel generator which are both non-safety related power sources. This relaying will not change the reliability of the --
-H N , d er -
emergency service buses A5 and A6 when N =-J., , emergency on-site sources are suppling power. pg
- 3. Main Control Room Panels (C3 and C5):
CONSTRUCTION # The main control room panel (C3) annunci Ror,rrtiie wdified to-provide annunciation for blackout diesel generator trouble and new switchgear breaker trouble. The existing control switch for the 23kv line circuit switcher will be moved to a new location on the I Exhibit 3.07-A L_ _-__ - _
. w ~ xm.a . .. wr . .o.a. > :.. :~- :n : .:n ,. w w. r .. n --- =:. w s v 5-15-87 A &ck O f Safety Evaluation yl'f /
Sheet 5 ofet same panel. Control switches and indicating lights for the blackout diesel generator and the two new 4.16kV circuit breakers will be installed on panel C3. .
- 4. 480V Load Center B4:
The 480V distribution system is described in section 8.4, Auxiliary Power Distribution System, of the FSAR. The 480V ac power feed to the blackout diesel generator maintenance loads from the PNPS auxiliary distribution system will be tapped from existing load center B4. Load center bus B4 is a normal service bus and is not safety related. Safety Functions of Affected Systems and Comoonents Q
- 1. Secondary ac power source (shutdown transformer) .
The secondary ac power source provides a non-safety related source of power to the emergency service portion of the auxiliary power distribution system. When the unit is tripped (i.e. unit ac power is not available), the preferred off-site ac power (start-up transformer) is not available and the emergency diesel generators do not start, the secondary ac power supply provides power to the emergency service buses.
- 2. Emergency Service Buses A5 and A6 The emergency service portion of the auxiliary power distribution system under design basis conditions, distributes ac power required l
to safely shutdown the reactor, maintain the shutdown condition, and operate all auxiliaries necessary for station safety. The control of the 4.16kv circuit breakers A501, A600, and A601, which control the power from the shutdown transformer, are not safety related. However, the controls for these breakers are interlocked to prevent interconnection with either the unit ac power source, the preferred ofi'-site te power source or the emergency diesel generators. These interlocks backed by procedural restrictions protect against failures associated with paralleling nonsynchronized sources. The existing controls of these circuit breakers are not changed except for the modification to the shutdown transformer differential relaying wh = .. .,. .:~._ 6 J= % initiat4en. ISSUED FOR ;
- 3. Control Room Panels C3 and C5
=1 Control room panels C3 and C5 will immace e-orm :;aw related functions of the electrical distribution system. The changes incorporated in this PDC however, do not interface with these safety related functions.
Exhibit 3.07-A
- s. . u :a.. .. .
. -2: . .~ ..a.: , .: :m . v. . , < n x , w. , - :.: = . .
7 5-15-87 A &oeh W l
'~
Safety Evaluation 2144 Sheet 6 of 9
- 4. 480V Load Center B4 1 l
480V bus 84 has no safety function. E, Effect on Safety Functions ! 1
- 1. Secondary (off site) ac power source-I The new blackout diesel generator is a backup to the secondary I j
off-site power source and is manually started. The diesel generator ' breaker A801 is normally open, the new breaker in. the secondary of the shutdown transformer (A802) is normally closed and the present alignment of breakers A600, A501 and A601 are not modified by this change. The shutdown transformer's ability to supply emergency buses A5 and A6 under design basis conditions therefore will not be affected during normal operation. During load testing of the blackout diesel generator it will operate in parallel with the shutdown transformer. Lineup of the switchgear to supply power from the shutdown transformer to the 4.16kv emergency service buses will remain the same during load testing of the blackout diesel generator as during normal operation. Therefore, power from the shutdown transformer (or the blackout
- diesel generator) is available to the emergency service buses A5 and A6, if required.
The 4.16kV breakers A501 and A601 interface the shutdown transformer and the blackout diesel generator to buses A5 and A6. These breakers will remain open at all times when buses A5/A6 are supplied power from the unit auxiliary transformer, the start-up transformer or the emergency diesel generator, thus isolating the blackout diesel generator from the emergency buses. The blackout diesel generator will supply power to the emergency essential 4.16kV buses A5 and A6 only when all other sources of ac power are lost (i.e., under station blackout conditions). The single failure analysis for the secor.dary ac power source discussed in section 8.3.3.2 of FSAR is not affected by this change I since the blackout diesel generator is wired via existing shutdown transformer cabling from the turbine building south wall into the safety related switchgear and sufficient separation is maintained between the shutdown transformer /blae' W generater ene rnh and the emergency diesel generator controlsISSUED FOR
- 2. Emergency Service Buses A5 and A6 CONSTRUCTION The failure of an emergency diesel gn = = = W er; M f.ag L emergency service buses A5 or A6 after a loss of power would result
( in automatically connecting the affected bus to the secondary off-site ac power source after a time delay of approximately 12 i seconds from loss of power. The secondary power source may also be manually connected to the emergency service 4.16kV buses to reduce Exhibit 3.07-A l
& a 4.a...u A vlu % .x; + . a v, u. . J. u. . . , :d ..M .u.s s EG 2 ~ . W D
- 5-15-87 A&ck O f Safety Evaluation 2 s vg l
Sheet 7 of9 the duration of emergency diesel generator operation whenever both unit power source and the preferred power source are unavailable. The operation of the emergency buses is not affected by the addition i: of the blackout diesel generator except for the modificati'ons to the shutdown transformer differential relay. The existing shutdown transformer differential relay (187-5) is non-safety related and supervises the zone from the primary of the + shutdown transformer to the emergency service bus breakers A501 and I A601. The existing two element differential relay will be replaced by a three element differential relay to include protection of the l blackout diesel generator set. Controls of the 4.16kv circuit j breakers A501, A600, and A601 are not safety related. The existing relay and new relay are both non-safety related. A failure of the relay will not cause a malfunction of the system (such as breaker closing when not required) nor trip any of the safety related breakers of the emergency service buses controlling the power from 4 j the unit ac, preferred off-site ac and emergency diesel generators. l
- 3. Control Room Panels C3 and C5 Although the changes made to Q panels C3 and C5 by this modification
- are non safety related, all work to these panels is performed utilizing Q-materials seismically installed to procedures qualified for safety related work. Therefore there will be no impact on safety related functions of the panels.
- 4. DC system
~
The controls of the blackout diesel generator and the new 4.16kV breakers will be supplied 125V DC power from a new independent battery system dedicated for this purpose. This 125V DC control power supply is not connected into the existing de system. Therefore, the existing station de system is not affected by this modification.
- 5. Fuel Oil The fuel oil for this diesel generator set is obtained from dedicated tanks and therefore will nampse- w .;4ib.. y dies #
generator fuel supplies. {} QQ { Separation Criteria
- 6. CONSTRUCTION Although the modification incorporate 6 by W i > rs; i s rJr.:T&TET.Ty- '
related, separation criteria has been maintained between the blackout diesel generator and all ether sources of AC power except the ;hutdown transformer. The cabling for the blackout diesel generator controls are routed separately from the controls for all other power sources. The blackout diesel generator control switches l on the main control room panelboard C3 are mounted on the vertical of the panel (with the shutdown transformer controls) and controls for all other AC power supplies are on the benchboard section of the panel. 1 i Exhibit 3.07-A
p,5;; m. .;;. ;: n my w;m.: x .: _ .;r- .~ ; m., <
- a. ...x : . : <.s l l
5-15-87 4+k44. F/ Safety Evaluation 2.8 48
. Sheet 8 ofq Power cabling from the blackout diesel generator (switchgear A8) is routed to the emergency switchgear via the routing established for the shutdown transformer as discussed in the FSAR, Section 8.3.
Therefore, independence o maintained between the blackout diesel generator and the emergency diesels as between the shutdown transformer and the emergency diesel generators.
- 7. II/I Interfaces )
l All conduit and cable installed by this design change is non-safety j related, however, all conduit located within safety related areas will be supported in accordance with seismic II/I criteria. In addition all drilling of seismic 1 structures are performed as Q activities. F, Analysis of Effect on Safety Functions
- 1. Secondary ac power system The seco'ndary ac power system, the shutdown transformer and the '
controls of 4.16kV circuit breaker A600 are not safety related. The new blackout diesel generator and associated 4.16kV switchgear are also not safety related. The new 4.16kV bus enables either the ) diesel generator or the shutdown transformer to supply power to the emergency service 4.16kV buses as required. By this modification, an additional 4.16kV circuit breaker is introduced into the 4.16kV i power supply circuit between the secondary of the shutdown transfonner and 4.16kV breaker A600. This circuit breaker controls , the power from the shutdown transformer to the new 4.16kV bus. The addition of a new 4.16kv (A802) breaker in the secondary supply of the shutdown transfonner does not adversely affect the availability of the secondary ac power except under conditions of malfunction (tripping) of the breaker and shutdown transformer overcurrent relay. In such an event, the blackout diesel generator will be available to supply power to breaker A600 through breaker A801. This new 4.16kv metal clad switchgear will be designed, built, rated and tested in accordance with recognized industry codes and ANSI I
. standards. Thus the overall availability of the secondary power system is not reduced by this modification. , ,
J
- 2. Emergency service buses A5 and AG The emergency service buses A5 and A6 will be affteted by tMs
, {d ication as these buses can be supplied power from the new cut diesel generator. The addition of the blackout diesel generator to the existing scheme will not impact the abilf t of I the shutdown transformer to supply power to the amergency b ses during nomal operation, since the blackout diesel generator will hot normally be operating. The blackcut diesel generator will tie 4 added scheme. to the shutdown transfonner differential current rela The new differential relay scheme provides trip signal to 4.16kv breakers A600 A501, A601 A801. A802 and the 23ky line circuit switcher.
ISSUED FOR CONSTRUCTION l
~; W : m m . z ,; m y ww . ,-- _ ,,
q Safety Evaluation 2; y f ' Shset 1 ofit 4 The new differential relay scheme provides the same. level of j 4 protection for buses A5 and A6 for faults in the secondary ' L ac power system. The controls for all of these breakers are non-safety related. There are no changes to the safety rejated
" portions of the emergency service buses. The ratingrof the safety related breakers are adequate after the addition of the blackout diesel generator set. The safety functions of these buses are not affected by this modification.
- 3. Control Room Panels C3 and C5 The main control room panels C3 and C5 are safety related panels.
However', the control switches and indicating lights for the blackout diesel generator and associated 4.16kv switchgear installed on panel C3 are not safety related. The . differential relay mounted on panel C5, is also not safety related. Even though the changes incorporated by this PDC are not safety related, all modifications to panels C3 and C5 will be implemented to Q criteria to prevent II/I failures.
- 4. 480 Volt Load Center B4
'This system does not perform a safety function. The maintenance - loads of the new diesel generator will be supplied power from this load center when the diesel generator is not operating. The additional loading on the buses is within the capacity of the load center.
g Summary Based o' the preceding evaluation, the addition of the blackout diesel gener dc+ does not increase the probability of occurrence or consequences of m accident or malfunction of equipment important to safety as previoufly evaluated in the FSAR. This modification does not create the possibility for accident or malfunction of a different type than any evaluatt d previously in the FSAR, nor does it reduce the margin of safety as defit ed in the basis for technical specification. Therefore, this modification does not result in an unreviewed safety question per IOCFR50.59. l ISSUED FOR CONSTRUCTION _.__.___L-.-__-_-----
, s -i a. ... a ., -
. . .. , ; '~ . . < . ~ ' :. . . . , i
--~-- 1 PfLGRIM STATIM ,3 g ,, 2 ; y FSAR REVIEN SHEET l
References:
Safety Evaluation: Rev. No.: 0 Date: 5/18/87 ; Support a change l List FSAR test, diagrams, and indices affected by this charige and corresponding FSAR revision. 4 1 Affected FSAR Revision to affected FSAR Section is shown on: Section Preliminary Final 1.6 "I Att;;tsr,t i 8.0 Attachment 2 8.1 4tta;tsat 1 8.3 Att;;is nt ? l l 8.10 (N M , Att;;IAn 6 3 Att;;is at ; l PRELIMINARY FSAR REVISION (to be completed at time of Safety Evaluation preparation). Prepared by: N.b /Date: S/fB/OReviewedby:d /Date: I/Mf7
/Date: M2 ,,
Approved'by: , / FINAL FSAR REVISION (Prepared following operational turnover of related systems structures of components for use at PNPS). Prepared by: /Date: Reviewed by: /Date: 1 4 Attach completed FSAR Change Request Form (Refer to HOP). 4 Exhibit 3.07-A Sheet 3 of 3 ISSUED FOR ' CONSTRUCTION I I 3.07-15 Rev. 4 l
777 ;5;W 1.: q w.; .J = . a . - ;;; . . .. . m' .. . x ,is >,: , Safety Evaluation 21'14 Attachment 3. Page 1 of 11 RECOMMENDED FSAR CHANGES The FSAR needs to be updated as a result of the addition of the bladkout diesel generator (PDC-86-56B). This attachment identifies the pages requiring revision and a new section for the blackout diesel generator. The following pages have been marked with suggested changes for incorporation. FSAR Section FSAR Pace No. Att. 1 Sheet No. 1.6 1.6-8 2 8.0 (T of C) 8-ii & 8-iii 3&4 8.1 8.1-1 to 8.1-2 5&6 8.3 8.3-2 7 8.10 (New) To be assigned 8 to 11 lig.1.g: In addition to the above, the following drawings will be revised as part of the Plant Design 'hange package but the associated FSAR figures are not included herein.
. Dwa. ID FSAR Fioure Title El 8.2-1 Single-line Diagram Station E6 8.2-2 Single-line Metering and Relay Diagram E7 8.4-1 Single-line Metering and Relay Diagram E9 8.4-2 Single-line Metering and Relay Diagram E13 8.6-1 DC System ISSUED FOR CONSTRUCTION i
.;a m - , , .
I .. . . . . . . . . . . . . .
. . . . . . - - . ]
Safe %y Evaluation 21# AtBachmeng)
' Sheet 2 of it pgp3 73g
'E* N:
l g will pass through the condensate deep bed demineralisers. ThehEp;9 purified flowwillthencombineintotwoparallelstreams,eachwith})"h* three stages of low pressure feedwater heating. Feedwater will then *uj boosted in pressure by the reactor feem pmps. There will be no be regulating valvas in the piping between the conde. sate pumps and the bE* reactor feed pumps. The flow from the three u ntrifugal, motor S l driven reactor feedwater pumps vill be combined into two parallel streams, each with two stages of high pressure feedwster heating. r8.o*g.g {*c The feedwater will then flow to the reactor. Control valves will be TSx located in the piping between the high pressure feedwater neaters and 5 ,5 ~ the reactor for regulation of feedwater flow. a as E E' 1.6.1.5 Electrical Power System 'jx* I o Y S .* { The station main generator . feeds electrical power at 24 kV via) an s isolated phase bus into the main transformer which steps the voltage E743 k o up to 345 kV. A 345 kV power connection is made between the main Eg*s'.as 345 kV x E h:' trans former and the 345 kV switchyard ring bus. The
/ i$
5 switchyard ring bus is connected to the startup transformer. Power
.7," " E from the station is transmitted to the Boston Edison system at 345 kV :5 af." $ .
l g/through transmission lines which are interconnected with the major 'E'i5 x o 8 .* E
- ':liT E
power systems in the New England area. I=$$ b{ station trans former auxiliary powertoduring connected the 345startup will be from kV switchyard the startup ring bus. In .5 5*5Q ve n, , k addition, power for station emergency shutdown will be from two to f Sg'standbydieselgeneratorsorfrya24kVlinewhichisconnected - j the New England grid by an independent 115 kV line. f
'1.6.1.6 Radioactive Waste Systems The Radioactive Waste systems are designed to control the release of station produced radioactive material to within the limits specified in 10CTR20 and to meet the design objectives of 10CTR50, Appendix I.
This i s done by various methods such as collection, drisuning, filtration, holdup for decay, and dilution. The methods employed ,i47 for 1, 2 the controlled release of these conte s. upon the state of the material liqund, sol I, 1.. 1.. 1 u guid madwaste syste* CO NSTRUCTION .h, The 1.iquid Radioactive Waste Control Systee-coneetsv-Wrets; These wastes are and disposes of all radioactive liquid wastes. collected in sumps at various locations throughout the station and then transferred to the appropriate collection tanks in the Radwaste Building for treatment, storage, dilution, and disposal as necessary. Wastes are processed on a batch basis. Processed liquid wastes may be returned to the condensate system or discharged to the environs through the circulating water discharge canal. The liquid wastes in the discharge canal are diluted with condenser effluent circulating water to achieve a permissible concentration at the site boundary. Equipment is selected, arranged, and shielded to permit operation, For inspection, and maintenance with minimum personnel exposure. 1.6-8 0
p . .m.s.. : a. . m .__ s, , , . . .~.e., m m._ . . . . . 2 .
.u..x. . .. . . .s..
Safety Evaluations l Attachment 2L %dV( i PNPS-FSAR Sheet 3 of 11 Section Title Page 8.5 STANDBY AC POWER S0URCE.................................<.a.5-1 8.5.1 Safety objective ...................................... 8.5-1 , 8.5.2 Safety Design Basis.................................... 8.5-1 ! 8.5.3 Description............................................ 8.5-2 Safety Evaluation...................................... 8.5-5 8.5.4 8.5.5 Inspection and Testing ................................ 8.5-7 8.5.6 Proposed Nuclear Safety Requirements for Initial Plant Operation................................ 8.5-8 8.5.7 Current Operational Nuclear Safety Requirements. . . . . . . . 8.5.10 8.6 125/250 V DC POWER SYSTEMS .............................. 8.6-1 8.6.1 Safety Objective ...................................... 8.6-1 8.6.2 Safety Design Basis.................................... 8.6-1 8.6.3 Description............................................ 8.6-1 8.6.4 Safety Evaluation........................................ 8 6-3 8.6.5 Inspection and Testing ................................ 8.6-6 8.6.6 Proposed Nuclear Safety Requirements for Initial Plant Operation................................ 8.6-6 8.6.7 Current Operational Nuclear Safety Requirements. . . . . . . . 8.6-8 8.7 24 V DC POWER SYSTEM .................................... 8.7-1 8.7.1 Power Generation objective ............................ 8.7-1 8.7.2 Power Generation Design Basis.......................... 8.7-1 ) 8.". 3 Description..............,............................. 8.7-1 8.7.4 Inspection and Testing ................................ 8.7-2 8.8 120 V AC POWER SYSTEMS .................................. 8.8-1 8.8.1 Power Generation objective _ _ .. .e.5-1 8.8.2 Power Generation Design Basj i.... 4. .g.14.E n. ..g h. @ .8-1 8.8.3 Description................. ......J J U.GM . . I . V. IN. 8-1 8.8.4 Inspection and Ter'.ing . . . . . ,. . 3 8.9 CABLE INSTALLATION CRITERIA...C. '~
- Tr~fi+i?~.~~....;..rr:1ht 1i~
8.9.1 General Design Criteria................................ 8.9-1 8.9.2 Specific System Wiring Criteria........................ 8.9-1 8.9.3 Physical Separation and Protection Design Criteria . ... 8.9-3 8.9.4 Installation Evaluation................................ 8.9-5 8.9.5 Cable Protection and Process Instrumentation i
~
Location Criteria...................................... 8.9-5
-vw - n _
( 8.10 BLACKOUT AC POWER SOURCE . . . . . . . . . . . . . . . 8.10
/
8.10.1 Power Generation Objective . . . . . . . . . . . . . 8.1C 8.10.2 Power Generation Desian Basis .......... 8.10 8.10 8.10.3 Description............................. 8.10.4 Inspection and Tes ti no . . . . . . . . . . . . . . . . . 8.10 l
' l
" '~ s- j 8-ii i 1
m _ _ _ _ _ _ _ _ _ _ _ . _ _ _
- , . ,.. u.....
. w e.- ~ .. ,....; .- . - a . . .- ..... . -c -
x - . v;= c ., w w.mr.m m a PNPS-FSAR Safety Evaluation 2/4/ Attachment 2. - Sheet 4 of 11 SECTIoN 8 1 LIST of TABLES
- 1 Table Title 8.2-1 List of Hajor Electrical Equipment - Unit and Preferred AC Power Sources
/
8.4-1 List of Major Loads - Auxiliary Power S,ystem 8.4-2 List of Major Electrical Equipment - Auxiliary Power System 8.4-3 Periodic Tests - Class IE Power Systems l S.5-1 Diesel Generator Emergency Loads Standby AC Power System S.5-2 Standby Diesel Generator System - Typical Timing and Sequential Loading of Diesel Generators 8.5-3 Standby AC Power Source Equipment List S.6-1 List of Major Electrical Equipment - t j 125/250 V DC Power Systems 8.9-1 Cable Separation Requirements in Missile Zones y_ y 8.10.1 List of Major Loads-Blackout AC Power Source 8.10.2 Blackout AC Power Source Equipment List ,
- j ISSUED FOR COh STRUCTION J l l
1 e
, x . . . c.m w . _c. wa -- n > .-
. ;; ; x m n .- q Safety Evaluation 2 l'((
FNPSoPSAR Attachment 1 Sheet 5 of 11 SECTION 8 ELECTRICAL POWER SYSTEMS j 8.1 SUNMARY DESCRIPTION The station electrical power systems provide a diversity of dependable power sources which are physically separated. The station
- electrical power systems consist of unit and preferred ae power systems, the Secondary AC Power Systenfluxiliary Power Distribution b !
System, standby AC Power system, 125/250 V DC Power System, 24 V DC g Power System, and the 120 V AC Power System. See ANSI /IEEE E Standard 260-1978 for electrical abbreviations used in Section 8. The one line diagram, Figure 8.1-1 illustrates the powet flow and connection from the main generator to the Transmission System and y Station service system. Electrical diagram symbols are emplained on % Figure 8.1-2. E ' The unit ac power source provides ac power to all station auxiliaries 88, and is the normal station ac power source when the main generator is g ; r opereting. The station preferred (offsite) ac power source provides ac power to all station auxiliaries and is normally in use when the k unit ac power source is unavailable. See Section 8.2. M The. secondary (offsite) ac power source provides ac power to essential station auxiliaries. It is used to supply essential 'l' station auxiliary loads only when the main generator is shut down and there is an extended outage of the. preferred ac power source. See p
,e section 8.3.
S The station Auxiliary Power Distribution system distributes all ac ,, power necessary for startup, operation, or shutdown of station loads. i g All portions of this distribution system receive ac power from the , g unit ac power source or the preferred ac power source. The emergency service portions of this distri5"**-- -"-*- "- _--- --- '-- l power from the standby ac power sou !ce ori gt i gpayiarM Spower [ source. See section 8.4. lJ VCJ TV N
- J The standby ac power source gr h k h hJJ a-bmnsy-,ewice
[nk h b , {p q generators as the onsite sources of = ;-- - - 4 portions of the station auxiliary Power Distribution System. Each (ge onsite source provides ac power to safely shut down the rea cor, maintain the safe shutdown condition, and operate all m iliaries PE'j necessary for station safety. See section 8.5. l l l The station 125/250 V DC Power System provides two independent onsite sources of de power for startup, operation, shutdown, and all loads essential to r.tation safety. See Section 8.6. ; The station 24 V DC Power System, provides a reliable onsite source of power to some radiation monitoring instrumentation. See section 8.7. 8.1-1 i
- n. .o .
- s '~. ,
- e ,a. . .
Safety Evaluation 2/ V4
- Attachment 2 pg,43g Sheet 6 of 11 The station 120 V AC Power System provides a versatile distribution instruments, and system to supply ac power to the station computer, and conventional control devices requiring uninterruptable power See section 8.8. l instrumentation, and monitoring systems. j t
~7 ' M .
The Blackout AC Power Source provides an independent diesel generator as the onsite source of AC power to:.the emergency service portions of the Auxiliary Power Distribution System in the unlikely event of a loss of ' preferred and secondary offsite power sources combined with a complete failure of the. Standby AC Power System. See Section 8.10.
-_ A
)90D/ T/OUAb AGAA?$
ykDD &/3 O ISSUED pon .. CONSTRUCTION l l l 8.1-2 l
c . j, :.w r ,, c m
,, __ , _. , . . , . , ,a.,s.
2lg SafetyEvaluatih . Attachment 3 PNPS-FSAR Sheet 7 of 11 The source for the 23 kV line is the Manomet Substation of the Commonwealth Electric Company. This substation has sufficient " capacity to supply the area load and, in addition, that lead * , required by the shutdown transformer to supply the two 4,160 V b emergency buses at Pilgrim Station. The 23 kV line is conducted d@g } overhead with one 336 KC mil ACSR cable per phase. 1 The 23 kV line and the shutdown transformer are normally energized with power supplied from the Manomet Substatiend g Breaker 3152-60M3FIjIerated normally closed and breakers?152-501 m hind 152-601 are operated normally open. 1 Refer to Figure _8.4-1 g for breaker arrangement. The shutdown transformer and diese13 cc generators supply power to the two 4,160 V emergency service
,g buses (A or B). The A bus is separated from the B bus by a
- concrete floor. The 4,160 V cables from the shutdown transformer y are run in an underground duct tank into the Turbine Building,
= and then in rigid steel conduit to the south end of emergency j service buses A and B. Each diesel generator's power leads are f .c run in separate conduit and cable tray to the north and ef each(53 bus (A or B). Since the le ^ m-6 " 4 h"= f - any =4teJ E$
direction, maximum possible se u paratio $La e :neDd. FOR M,** i
- 2. Single Tailure Analysis E
>3Eg2H L. The shutdown transformer and
- CONSTRUCTION V: 5
;- - -- - - - -- - - " f r tb h O 2ee each emergency service bus are electrically independent as far as ji O .3
)*8m . I D 'E [j possiblewithoutcompromisingtheindependenceofthesafeguardAqg*ECgL and B buses. For example,. the de supply to bus A controls are 9o l MS "fYE from battery A, hence both shutdown transformer and diesel Ew 5 # generator supply breakers have a common control power source.kEuE E E The control power source for Bus B is independent from Bus A;/b
- 2 L E E *y s however, control devices for each bus are separated from eachl e 2 "' I
*3TE other in the control board and their cables are routed separately EEi E E U "! to their respective switchgear buses. 3
-*gy -
- - - :- fy _E.E g l cox
~ ~d The centrel: fer the 22 '" tr:n:-1::ier line feedin- the ch"td *$5
*3 E o j tren:fer er ere len ted -id"ey St"eer the c entrol.a Jor eji*f
*t &c(
- die::2 1:ne r;t er en the rentr:1 hud./ Ine controls for trIe *' " f shutdown transformer supply breaker to one 4,160 V bus are not $e 75sgU Eoe separated from the controls for the diesel generator supply d- 3 * ,
E 5 m $ ci breaker to the same bus. The controls for the A and B buses are' uMb
e f o S separated, however, to satisfy the design intent for separation 'E2 $
%*u5% of the two emergency service buses. DC control power is 8e"'
Riff 5 i individually supplied to each switchgear bus, maintaining imf e o % m .c I separate routing from separate batteries. Each diesel generatorlg g ',
'f8eE* is individually supplied with de control power, maintaining boo u"ES3 separate routing from separate batteries. The shutdown E*E EEeE transformer protective relays receive de power from 125 V de "b*
y E @} m *"5E' panel C which may be supplied from either battery. o'E ,E " I I t h * "cm l " A single failure in the controls or interlocks in the standby ac u ," M E .i E power (diesel generator) supply breakers, or in the secondary ac yo'
" E 'E % T power (shutdown transformer) supplybreakers,wouldnotpermitfEy*
l)j2E8 simultaneous electrical interconnection of both diesel generatorsj s o .* 8.3-2
n: x,2; 7.- , . . . 3:m ; ,, ,, ~ w ;r _ , , , ,a ., , ,r. & , m c. , 5 15-87 Safety Evaluation 2J # l Attachment 3 l Page 8 of 11 8.10 BLACK 0UT AC POWER SO'JRCE 8.10.1 Power Generation Objective The Blackout AC Power Source consists of an independent diesel generator which provides a non-safety related source of onsite AC power l to the 4.16kV emergency service buses through a two breaker 4.16kV bus in the event of a station blackout. l 8.10.2 Power Generation Desian Basis
- 1. The blackout diesel generator has the capacity to carry the load Nociated with one of three ECCS pumps on either emergency service bus A5 or A6 and all associated loads on that train required for loss of offsite power without a LOCA. Some loads i associated with buses AS or A6 (See Table 8.10-1 for ratings) are one salt service water pump (SSW), one reactor building closed cooling water pump (RBCCW) and various electrically operated valves. Operator action is required to limit blackout diesel generator loading and automatic start of the 4kV-ECCS pumps.
- 2. The Blackout AC Power Source is completely self-contained, not relying on any permanent plant system for operation. The blackout diesel generator is mounted on a skid and housed in a pre-engineered enclosure for protection from the environment. The unit is capable of providing rated power (See Table 8.10-2) for a minimum period of one week without refueling.
- 3. Maintenance loads for the blackout diesel generator are provided by a 480V feed from the station during normal operations.
However, upon loss of power, the unit is capable of "blackstart", automatically picking up the maintenance loads. The Blackout AC Power Source is a manual start system either from the main control room or locally from the diesel enclosure.
- 4. The Blackout AC Power Source will operate in parallel with the shutdown transformer during load testing of the blackout diesel.
8.10.3 Description The Blackout AC Power Source is connected to PNPS through a two breaker 4.16kV bus A8 with the blackout diesel generator connected to the first breaker, A801, and the shutdown transformer secondary connected to the second breaker, A802. The 4.16kV bus A8 is connected by cable to breaker A600 of the erhergency service buses. Power frora the secondary AC Power Source (shutdown transformer or blackout diesel generator) to the 4.16kV emergency service buses is controlled hv hie na~ *^ " -O L
""d ^' '
ISSUED FOR CONSTRUCTl@ 1 _ . , . . . ~ _ = . _
. wa. .. x .. .; . a: a, a:w : . ,m.,. ..
~ ~ x:cu:nu. '
5-15-87 Safety Evaluation 2/ Li/ Attachment 2 Page 9 of 11 The controls of circuit breakers A801 and A802 are interlocked to prevent interconnection of the diesel generator with the shatdown transformer er. cept for testing of the diesel generator. The blackout diesel generator is interlocked, as is the shutdown l transformer, to prevent operation in parallel with the unit AC , power source, the preferred offsite power source. or the standby AC :ower source. f The diesel generator and the 4.16kV breakers of bus A8 are controlled manually either from the main control room or locally within the diesel generator enclosure. The diesel generator has sufficient fuel capacity to supply ratedload for a minimum of one week. Protective relaying is provided to prevent damage to the diesel generator and the shutdown transformer. Relaying is also provided to prevent the diesel generator from supplying unacceptable power (voltage and frequency) to the emergency 4.16kV buses A5/A6, as applicable and to the 24kV system during load testing of the diesel generator. The diesel generator on starting will pickup its auxiliary load automatically under station blackout conditions. An independent 125V DC system is provided to supply control power to the blackout diesel generator unit and associated 4.16kV switchgear A8.
~
8.10.4 Insoection and Testina Inspection and testing by the supplier and initial system tests were conducted to assure that all components are operational within their design ratir.g. The diesel generator will be tested at regular intervals for its ability to start and pickup load by operating it in parallel with the secondary offsite AC power source (shutdown transformer). The diesel generator will be tested, during unit shutdown, for connection to the 4.16kV emergency service buses and acceptance of the required loads. ISSUED FOR CONSTRUCTION 4
(,e . . :,=.a. . n .. .. .,x. . -a.- . . .~.a.- ., - a w . .u. =s --
, . = :wx wc ;
5-15-87 l Safety Evaluation 2/ 4/ { Attachment 3 J
, Page 10 of 11 1 TABLE 8.10-1 LIST OF MAJOR LOADS BLACK 0UT AC POWER SOURCE ,
I Loads on 4160V switchgear buses (blackout service) Bus A8 ECCS Pump 800 HP ^ Salt Service Water Pump 100 HP RBCCW Pump 60 HP Bus A8 supplies power to these loads through emergency service buses A5 or A6. The loads are manually connected as required.
~[
ISSUED FOR CONSTRUCTION 1 of 1 ; h
- m. ~ = =c . .:u , ,. . . . . . -- - . .a .- , .. c.. ... -
l, . 5-15-87
' Safety Evaluation 21Y4 Attachment 2.
I PNPS-FSAR TABLE 8.10 2 . BLACKOUT AC POWER SOURCE EQUIPMENT LIST l DIESEL ENGINE , Rated Speed 1,200 rpm Continuous Rated Capacity 2,000 kW l Standby Rated Capacity 2,200 kW Fuel Consumption at Rated Capacity .498 lb/kW-hr GENERATOR Continuous Rated Capacity 2,000 kW l Power Factor 0.8
~
Frequency 60 Hz i Voltage 4,160 V ! Phase (Connection) 3 (wye) SWITCHGEAR 4160 V (Blackout Service), 350 MVA,1200A Breakers, Switchgear A8 FUEL OIL STORAGE Day Tank 275 gallon Main Storage 40,000 gallon (2 tanks 9 20,000 gallons each) ISSUED FOR COfjSTRUCTION
c w. ' ,: p . . u .u , * .
. .w ., .. . - -
.c - :
Safety Evaluation ; No.: 2/44 >
., SAFETY EVALUATION E SHEET A. Systes Structure Component Failure and Consequence Analyses.
Systes structure fa=aonent Failure Modes Effects of Failure Coinints 1. SEE ATTACHMENT No. 1
}
2.
- 3. .
d i General Reference Material Review . FSAR ' CALCULATIONS REGULATORY SECTION PNPS TEGINICAL SPECS. DESIGN SPECS PROCEDURES ELIDES STAllDARDS CDDES
- 8. For the proposed hardware change, identify the failure modes that are likely for the components consistent with FSAR assumptions. For each failure mode, show the consequences to the systes, structures or related components. Especially show how the failure (s) affects the assigned safety basis (FSAR Text for each system) or plant safety functions FSAR Chapter 14 and Appendix G).
Prepared by # A Date (//9/67 i e a NOTE: It is a requirement to include this work sheet with the Safety ~' Evaluation. Exhibit 3.0?-C L CONSTRUCTION 3.07-18 Rev. 4 1
w wiu,.s mm w. u;:. ..a x , . - > . .w.= ., , a. . :> - u.- :. , '-
- v. = x-== .mw-Safety Evaluation No. 2124 Sheet i ot 4 5,J E TY ?. .'.J:'; >h PILGRIM NUCLEAT'"FU KR 5TATION .
PDC PCN System Calc. Initiator: : Dept: Group: No.: Nane: No.: Date: L.R. Namer NED S&SA 86-73 Automatic 3/2/87 D. Gerlits Depressurization System Description of Proposed change, test or experiment: Addition of high drywell pressure and reactor low pressure bypass timer, and manual ADS inhibit switch to ADS logic. Addition of reactor low pressure bypass contacts to RHR and core spray logic. SAFETY EVALUATION CONCLUSIONS: The proposed change, test or experiment:
- 1. (X) Does Not ( ) Does increa'se the probability of occurrence or consequences of an accident or malfunction of equipment fuportant to safety previously evaluated in the FSAR.
- 2. (X) Does Not ( ) Does create the possibility for accident or malfunction of a different type than any evaluated previously in the FSAR.
- 3. (X) Does Not ( ) Does reduce the margin of safety as defined in the basis for any technical specification.
BASIS FOR SAFETY EVALUATION CONCLUSIONS: See attachment 1 t l I Change Change . l (X) Reconnended ( ) Not Recommended . 4f.. Namer /.Ger11tsQ.kw/A-WKAle SE Performed by L Date
, /F7 !
l Exhibit 3.07-A Rev. 3 l l Sheet 1 of 3 l
m.xm=.1:<. x . c w u.c . - . .uu n . w. . . .: c: .: - ,. w. . . n :. . -~- . a Safety Evaluation No. 2.l24 Sheet 2 ot4 S AFETY EV.i t:! A' M PILGRIM NUCLEAR POWcn a i ,u t v., A. APPROVAL ( ) This'broposed change does not involve a change in the Technical Specifications. . (X) This proposed change, test or experiment does ( ) does not (X) involve an unreviewed safety question as defined in 10CFR, Part 50.59(a)(2). (X) This proposed change involves a change to the FSAR per 10CFR'50.71(e) and is reportable under 10CFR 50.59(b). ( ) Commnts: The safety evaluation basis and conclusion is: ! (X) Approved () NofApproved k U l- 4 l, lo + Discipline Group Leader /Datr Supporting Discipline Group Leader /Date B. REVIEW /APPRO/ AL (64 Corcents: N%
$!lE ,
e45&SA Grot (p) Leader /8atel C. ORC REVIEW ( ) This proposed change' involves an unreviewed safety question and a request for authorization of this change must be filed with the Directorate of Licensing, NRC prior to implementation. ( 4 This proposed chan d es et involve an unreviewed safety question. ORC Chairman 0, - Date f!/E!87 ORC Meeting Number 7-49 . cc: Exhibit 3.07-A Rev. 3 (Sheet 2 of 3) ,
)
i
m.xmsn.:.m=- c . : . - .:,... m :~ ., , . w, . .. . w. . . : .. ; , .; . . n . . Safety Evaluation No. 2/24 Sheet 3 of g; 1 1
- - PILGRIM STANON
- F5AR RE/IEw 5MEtl
References:
i Support a change List FSAR test, diagrams, and indices affected by this change and i corresponding FSAR revision. Affected FSAR Revision to affected FSAR Section is shown on: Section Preliminary Final N/A a.tt;;h ;r.; 1 4.4.5 Attachment 2 7 V 7.4.3.3 Attachment 3 Figures 7.4-6,7,9,11; Attichment 4 4
- 7.3.6 Attech.;r.t 5 Att;;.t;r.t i PRELIMINARY FSAR REVISION (to be coupleted at time of Safety Evaluation preparation).
A- ' Prepared by: L.R. Namer / Date: h f /}O Reviewed by: A /Date: #2p/r7 D. Gerlits // ' Approved by: SA /Date: Y/1t/r7 _ h AfLN FINAL FSAR REVISION (Prepared following operational turnover of related systems structures of components for use at PNPS). (1) Prepared by: /Date: Reviewed by: /Date: ; e , (1) Attach completed FSAR Change Request Form (Refer to NOP). I Exhibit 3.07-A Rev. 2 (Sheet 3 of 3) l m,--___-_--._ - . - . - - . - . . - - - - - - .
c-ra xnm.:ra . =.r. . =. ::::..m: z a. . . . . . u - u w e.m - n -
- Safety Evaluation No. 2lH
-
- Sheet 4 ot e;
FETY EV ALUATION T.~ L. a r A. Systen/St .ure/ Component Failure and Consequence Analyses. ,
I Sysh m/ ' Failure Effects of Structure / Component Modes Failure Coments
- 1. High Drywell Pressure Timer Contact No ADS Channel Redundant Bypass Timer Logic Closure Failure Initiation Channels plus Signal Manual Capability Available ADS Inhibit Switch Switch Contact ADS continuous Redundant channel
- 2. channel inhibit plus manual ADS Open In Normal Position initiation
.available Timer Contact No RHR or Core Redundant channels
- 3. Low Reactor Pressure plus manual capa-Bypass Timer Logic Closure Failure Spray initia-tion signal bility available General Reference Material Revie CALCULATIONS / REGULATORY FSAR SECTION PNPS TECHNICAL SPECS. DESIGN SPECS / PROCEDURES GUIDES / STANDARDS 1
Sheet 49 GE Document NUREG 0737 II. K.3.18 4.4.5 (Table 3.2.B) 24A1719. Rev. 0 NE DO-24951 NUREG 0700 7.4.3.3 Sheet 73b . . 7.4.3.4 Sheet 73 (Sec. 3.2) GE Report EAS 154-1286 (Analysis) Sheet 73a i 7.4.3.5 Sheet 109 (Sec. 3.5.E) 505-C200.0 _ 570-C100.1 670-C100.2 B. For the proposed hardware change, identify the failure modes that are likely for n e e cponents consistent with FSAR assumptions. For eacn failure mode, show the consequences to the system, structures or related { components. Especially show how the failure (s) affects the assigned l safety basis (FSAR Text for each system) or plant safety functions FSAR Chapter 14 and ppendix G). 1 d Date b) kh Prepared by L.R. Name . D. Gerlits . . NOTE: It is a requirement to include this work sheet with the Safety Evaluation. Exhibit 3.07-C Rev. 2 1
m.ammx.a . : w w . w .... ~ . - . x ..a n. .. w - w. a= --
= x c.m.u. n . .w . , , x- 3 ~
Safety Evaluation Attachment 1 No. 1814
- Sheet 1 of 10 l S.M a
- _.iety Evaluation Conclusions 1
A. Description of Change: ] 1 The proposed change to the Auto Depressurization System contains ! these new features: -j a) allowing initiation of automatic vessel depressurization on sustained low RPV water level for 11 min. (nominal setting of the new i bypass timers), This supplements present logic by expanding automatic blowdown capability to include events that produce low water level in the RPV without producing high drywell pressure. b) modifies logic to permit continuing vessel blowdown if it has started, even if the low pressure pump running permissive signal is lost after SRVs have opened, thus reducing the need for operator . action. c) allowing ADS to be disabled with the new " ADS Inhibit" switches. This'provides the plant operator the capability to implement the Emergency Operating Procedures in a convenient manner, freeing him from having to repeatedly depress the reset pushbuttons before the
~
timers time out. Detailed Description The existing Automatic Depressurization System logic design requires concurrent trip signals for high drywell pressure and low reactor water level to initiate ADS. The high drywell pressure signal is sealed into the initiation sequence and does not reset even if the high drywell pressure subsequently clears. When both high drywell pressure and low-low water level trip signals occur, the 120 sec. timer starts. This timer will automatically reset if the low water level trip signal clears before:the timer times out. The_ timer can also be manually reset to delay initiation of ADS, allowing the operator to bypass the automatic blowdown if the conditions are correcting themselves or if the signals are spurious. The logic also - requires that a low pressure pump be running to assure that makeup water will be delivered after the vessel is depressurized. If the low water level signal clears before the ADS solenoids are energized, all the timers will be reset automatically. Using the reset pushbuttons before the ADS solenoids have energized will also reset all the timers. The proposed modification will use a new timer to bypass the , requirement for the high drywell pressure trip signal if the low reactor water level trip signal does not clear before the'new timer times out. This new timer will also bypass the low-low RPV pressure permissive of the RHR and Core Spray pump start after the same time delay. This feature will increase the automatic initiation j capability of the ADS, allowing auto initiation, if required, for , events such as breaks external to the drywell or a stuck open SRV. 4 4 l
m.a w a.a '
. ..w :a . :2 =. c . :s :. a . ..a . : i .~, e ':x = -
x .w - n.n - Safety Evaluation Attachment 1 No. 2l24 Sheet 2 of 10 A new 11 minute timer iner ... into the logic system aesign is initiated on the same l a~ 'ater level trip signal currently,used for ADS. An alarm alerts the operator that the bypass timer has been ' activated. After the set time delay, the relay contacts for the ADS high crywell signal are bypassed, and the RHR and LPCS pump start logic low RPV pressure permissive bypass circuit contacts are closed, producing the bypassed condition. This provides the signal to start the low pressure pumps and the existing 120 sec. timer. When the 120 see timer times out, the logic system confirms that the low pressure pump running signal is available and energizes the solenoid circuits on the four ADS SRVs. The logic will also permit vessel blowdown to continue even if the low pressure pump running signal is lost subsequent to energizing the ADS solenoids. The existing logic will terminate ADS if this signal is lost. The addition of the new bypass timer will not change the response of ADS during design basis accidents. When both high drywell and low low water level signals occur, the ADS logic will actuate the blowdown after a 2 minute time delay. This modification will nat change the response of ADS in the existing small break LOCA analysis. The " ADS Inhibit" switches proposed in this modification allows the capability to disable ADS. This feature, to be used only after the operator has entered the Emergency Operating Procedures, permits the operator to avoid having to reset the timers repeatedly to avert initiating ADS in situations where it is important to avoid uncontrolled injection of low pressure coolant. . One "two-position maintained contact" type manual inhibit switch will be provided for each division. Each switch will activate a white indicating light and an annunciator to alert the operator of the inhibit action. The inhibit switches do not affect ability of the SRVs to open on reactor pressure above the SRV setpoints or to open individual valves on manual signal. The inhibit switches will not be able to stop an ADS blowdown once it has begun. _ B. Purpose of the Change: Following the accident at Three Mile Island, increasing attention has been focused on reducing the need for operator action. The Auto Depressurization System was originally designed as a backup for the HPCI system and was needed for small and intermediate sized breaks inside the drywell. Item II.K.3.18 of NUREG 0737 requires that "The Au 4 matic Depressurization System (ADS) actuation logic should be mod.ified to eliminate the need for manual actuation to assure adequate core cooling. A feasibility and risk assessment was I required to determine the optimum approach. One possible scheme which was considered is ADS actuation on low reactor ve'ssel water level provided no HPCI or HPCS system flow exists and a low pressure ECC system is running. This logic would complement, not replace, the existing ADS actuation logic." The BWR Owners Group evaluation of J I i h
m a. r c- . m w ~.~ w ~ . n.2 -. -
-- :w.
. - . - v- r. = , ,
I
' ' Safety Evaluation j Attachment i No. 2124 4
" -+ 3 of *O j l
this resulted in GE report HEDE.. . . debruary 1983). The NRC has l accepted either addition of a bypass to the drywell pressure trip l plus a manual inhibit (Option 4 on NEDE-30045), or elimination of the l high drywell pressure trip plus a manual inhibit (Option 2 en l NEDE-30045) in their safety evaluation report on II.K.3.18. BECo q evaluated both and decided to implement Option 4. C. Systems, Subsystems, Components Affected: The systems affected by this change are: ADS, RHR and Core , Spray. Changes to ADS: a) Automatic start of vessel depressurization on low-low reactor vessel water level even if high drywell pressure is not high. b) If automatic depressurization has started, loss of low pressure pump running signal will not terminate vessel depressurization. (Using the reset pushbuttons will interrupt the depressurization and restart the timers. However, if the low pressure pump running rignal was lost after pushing the reset pushbuttons, ADS will'not reactivate automatically). c) Resetting the timers will delay ADS by 13 min. if the sequence was activated by only low RPV water level. If high drywell pressure started the sequence, reset action will delay ADS by 120 sec. d) The use of the " ADS Inhibit" switches will disable auto initiation logic, but use of the " ADS Inhibit" switches will not terminate a valid initiation of ADS once the actual blowdown has begun. Changes to RHR and Core Spray a) The low RPV pressure permissive (400 psig) to start the RHR And Core Spray pump will be bypassed after 11 minutes of sustained low low water level. The same timer whose contacts bypass the high drywell signal in the ADS logic will also close contacts in the RHR - and core spray pump logic. D. Safety Functions of the Affected Systems
- 1) System: ADS
'SAR Section: 6.4.2 datety function: In case the capability of the feedwater System, Reactor Core Isolation Cooling (RCIC) System, r,r.' HPCT. ^1:J:em is not sufficient to maintain the reactor .M n level, the Automatic Depressurization System functions to reduce the reactor pressure so that flow from LPCI and the Core Spray System enters the reactor in time to cool the core and limit fuel clad temperature.
g n:.r.- w .-. ~ ~.w . , w .~w. ~ .- , .: .. . . - .u.- , - . = - =m ~n-- Safety Evaluation Attachment 1 No. 2.l 2 4 Sheet a v ~ l 1 i
- 2) System: Core Spray
{ FSAR Section: 6.4.3 Safety Function: In case of low-low water level in the reactor vessel, or high pressure in the drywell, when reactor vessel pressure is low enough, the Core Spray system sprays water onto the top of the fuel in time and at a sufficient rate to cool the core and limit fuel clad temperature.
- 3) System: RHR (LPCI Mode) ,
FSAR Section: 6.4.4 Safety Function: In case of low-low water level in the reactor vessel, or high pressure in the drywell, when reactor vessel pressure is low enough, the LPCI mode of RHR pumps water into the reactor vessel in time to flood the core to limit fuel clad temperature. E. Effect on Safety Functions
. 1. Effect of the' ADS modifications:
- a. Addition of the bypass timer around the high drywell pressure signal will allow ADS to depressurize the reactor vessel for those events which have previously required manual operation. The existing logic is ,
designed for protection against excessive fuel cladding heatup upon loss of coolant, over a range of steam or liquid line breaks inside the drywell. The addition of the bypass timer would not change the system's response to breaks inside the drywell, but will broaden the spectrum of events to which ADS will automatically respond. Manual depressurization has been required for events such as: RPV isolations (including breaks outside the - drywell) with a loss of high pressure makeup systems, and a stuck open relief valve. This modification will provide additional assurance of adequate core cooling for these events which do not directly produce a high drywell signal, by eliminating the need for manual actuation to assure adequate core cooling during these events.
- b. Addition of a blowdown seal in relay will allow the ADS blowdown to continue, even if all low pressure CSCS pumps are lost, and the low pressure pump running permissive in the ADS logic is lost. This modification will'cause the blowdown to continue without low pressure makeup. Yht; modification will not change the number or type of events to which ADS responds. The modification will just continue the blowdown once it has begun.
pa m.;.r.s . - a.m.;.c w. . w , w w.a x , . . ~ =:, .>.. ,x... / m , - .rm.= mm -- Safety Evaluation Attachment 1 No. I f 14 - Sheet 5 of 10
- c. Addition of the " ADS Inhibit" switches will aklow the operator to disable the automatic actuation of ADS.
The operator has the ability to delay the initiation of ADS indefinitely under the existing logic, by pushing both timer reset pushbuttons every 120 seconds. This modification will offer the operator the existing ability to inhibit ADS actuation. It will reduce the possibility of inadvertent initiations of ADS by reducing the number of operator actions required to keep ADS inhibited.
- 2. Effect of the Core Spray and RHR Modifications These modifications will not reduce the number or types of events to which Core Spray and RHR presently respond. The addition of the bypass timer, contact around the low reactor pressure signal will allow core spray to respond, in concert with ADS, to those additional events discussed in Section E.1.a above., '
F. Analysis of Effect on Safety Functions
- 1. Analysis of the Effect of the ADS modifications
- a. Addition of high drywell pressure bypass timer:
NUREG 0737 Item II.K.3.18 requires that the ADS actuation logic be modified to eliminate the need for manual actuation to assure adequate core cooling. The BWR Owners Group responded to this requirement with NEDE 30045. NEDE 30045 states'that transients and accidents which do not directly produce a high drywell pressure - signal, and which are degraded by a loss of all high pressure injection systems, require manual depressurization of the RPV followed by injection to assure adequate core cooling. NEDE 30045 groups these events into two classes: (1) RPV isolations (including breaks outside the drywell) with loss of high pressure makeup systems, and (2) RPV isolations with a loss of high pressure makeup systems, further degraded by a stuck open reli:f valve. NEDE 30045 then refers to NEDO-24708A "Addi'tional Information Required for NRC Staff Generic Report on Boiling' Hater Reactors" Section 3.5.2.1, and says that the operator has at least 30 to 40 minutes to initiate ADS and prevent excessive fuel cladding heatup for both the classes of events listed above. This 30 to
=__-_:_ _ -- _ _
m e; m;v.w.u . ,_ a .,.s ..n.s . a m-m m, . .. 2. , = : .u ,= .w q l 0 Safety Evaluation Attachment 1 No. 2. / 2 4 3 2haet 6 of 10 40 minutes was based on starting from full power with equilibrium core exposure and complete failure of all the high pressure makeup systems. f The addition of the high drywell pressure bypass timer is J one of the two modifications proposed by NEDE-30045 that ] the NRC accepted. The analysis of the effect of adding 1 the high drywell pressure bypass timer is contained in l j
" Bypass Timer Calculation for the ADS /ECCS Modification for Pilgrim Station" EAS 154-1286 (DRF668-0003-5), dated December 16, 1986.
This report investigates the class of transients and events that do not pressurize the containment, but i eventually require ADS to depressurize the reactor vessel. The limiting case for this class of events is a Reactor Mater Cleanup line break outside primary l containment This event also assumed a loss of all high t pressure makeup systems. The analysis evaluated bypass timer settings of 16,17,18,19 and 21 minutes for their effect on peak clad temperature (PCT) during the event. The report concluded that a bypass timer setting of 16 minutes will not exceed the conservative limit of 1500*F. Another analysis was done by General Electric, titled ;
" Minimum Allowable Time Delay for ADS Initiation for ATHS !
l Events in Pilgrim" DRF-T23-637, dated December 18, 1986. ' This report recommended that the minimum time delay ' ~ between the ADS low level setpoint signal and actual depressurization is 6 minutes. : The actual time delay setting of 1182 minutes was chosen as providing equal margin from both the maximum and _ minimum recommended bypass timer settings. Electrical separation and the single failure criterion of the original design of ADS, RHR and Core Spray Systems, were satisfied by conformance to IEEE 279 (1968) and ths GE Specification 22A3034. This modification provides M ther improvement in electrical separation for RHR aiid Core Spray Systems. The added events of the above systems that are powered by bus "B" will be routed in flex conduits in the panel 932. The circuits added to the existing ADS logic that is powered by bus "B" are not required to be routed in flex conduits, since the entire ADS is a "Divison I" system.
m m n: w.= : w .a: w. m .a. w .. w. x .. :, . m =: . . a : .: ~r. . . . xw . ~. ,w. l Safety Evaluation Attachment 1 No. 212.4 Sheet 7 cf 10 i Moreover, the bus "B" is protected from faults which i l may occur in the "Divison I" panel 932 by fuses in the individual circuits, and by circuits overload protection in the power distribution panel. The electrical equipment to be installed has been ; seismically quiaified per IEEE 344. The equipment to be installed are relays (Agastat Type TR and GP), i switches (GE type CR2940), and indicating lights (GE type and ET-16). The switches and indicating lights i will be installed in panel C903 and the relays will be installed in panel C932. , Bechtel has performed the necessary analyses to show that the devices installed in the subject panels (C932 - 1 and C903) of this modification are appropriate for the seismic environment. Calculation 505-C200.0 was performed to develop in-structure response spectra (IRS) for panels C903 and C932. Calculation 670-C100.1 was developed to compare the manufactu'rer's test response spectra (TRS) for the relays to the TRS curves generated in Calculation 505-C2000.0 for panel C932. Likewise, i Calculation 670-C100.2 was developed to compare the manufacturer's TRS for the switches and lights mounted in panel C903..
- b. Addition of the blowdown seal-in relay l Analysis for this logic modification is contained in G.E. letter 698-86-144 to R. N. Swanson from R. R.
Ghosh and is summarized as follows: It can be assumed that vessel depressurization started - because all the necessary conditions to initiate ADS have been satisfied i.e., high drywell pressure low-low RPV water level, and 120 second timer runout with a low pressure pump running (such as for a LOCA inside the drywell); or new bypass timer runout, low RPV' water level and 120 seconds timer runout with a low pressure pump running (such as for a LOCA outside the containment) and no operator action to inhibit A05 blowdown. If all these conditions are satisfied, it is GE opinion that vessel depressurization 'should not be interrupted. l I i l
;..- . _ . ~. .. . c _ m . m ,. . .m m. .m m. ~ m. - .v. .. . . = ~ . -. . . . , . .
Safety Evalaation 1 E Attachment 1 No. If 24
'he+ 8 of 10 Analytical results for the limiting LOCA used.to establish the bypass timer setting show that 60-90 seconds after all the ADS SRVs have opened the vessel water level will have decreased to a point where it would collapse to approximately two thirds core height if the depressurization was halted. Rev. 4 of the Emergency Procedure Guidelines Contingency 3 instructs the operator to depressurize the vessel by. manually
' opening SRVs if water level is this low so that he can
. provide effective steam cooling of the core while trying to get any available makeup system into operation. Thus the operator would have instructions to. blowdown the RPV even if no ECCS pumps were available.
The probability is extremely unlikely that all six of the low pressure ECCS pumps would be unavailable, especially since confirmation of low pressure pump availability was necessary condition to start the process. It is also possible that the indication of loss of pump availability is an erroneous signal. One of the objectives of NUREG 0737 is to reduce operator decisions during a postulated accident scenario. It does not appear that maintaining the logic sequence that Pilgrim currently has in place, which will terminate ADS, is the preferred alternative. The modification recommended adds to the - spirit of the requirements addressed by NUREG 0737 and allows for continuation of an already started vessel depressurization sequence.
- c. Addition of the " ADS Inhibit" switches As part of the BHR Owner's group response to NUREG 0737 Item II.K.3.18 NEDE-30045 also considered Anticipated Transient Without Scram (ATHS) events in the development of the proposed modification to the ADS logic. The addition of a manual ADS inhibit switch was part of both of the approved, logic modification options.
The keylocked " ADS Inhibit" switches pro! de capability to conveniently disable the automatic logic for starting vessel depressurization. The TOPS, provide instructions to deliberately disable the ADS logic for the following two conditions only: ( 6 l
- l
n x:.i : ;an u.:w .= r;.>u.w w , .::m .x. . .o . - vi w 2.z . ~: =~m :z ;
. - I Safety Evaluation Atta:hment 1 No. 2:24 Sheet c ~ * ^
j l
- 1) There has been a failure to scram, water level is dropping or being reduced and Standby Liquid Control System has been activated to inject boron.
For a failure to scram event, EOPs provide a specific method for vessel blowdown, if required, to avoid uncontrolled low pressure cold water injection and concomitant potential reactivity excursion. In this scenario, the ADS Inhibit switches aid the operator by ; permitting him to concentrate on other variables instead of having to remember to repeatedly use the reset pushbuttons.
- 2) A liquid line break has occurred, core uncovery is a concern and it is desirable to conserve remaining inventory. The operator, who has more information available than the automatic system logic, has entered the EOPs under Contingency 1. He uses the inhibit switches to avoid ADS actuation while increasing coolant injection into vessel. The operator will manually initiate vessel blowdown, when it is required, in this scenario.
- 2. Analysis of the Effect of the RHR and Core Spray Modifications Implementation of the ADS logic modification described above requires that the pump start logic to RHR and Core Spray be changed. This aspect of the ADS logic system modification was analyzed by General Electric for the BWR Owners Group in AE-06-0184 " Modification of ECCS Pump Start Logic".
This report states that neither low RPV pressure nor high drywell pressure would be expected on a timely basis for the events that the ADS modification is intended to - cover. Thus, the timely start of the low pressure pumps could not be assumed without operator action. Furthermore, since automatic initiation of ADS requires confirmation that at least one low pressure pump is running, timely ADS actuation could not be assured without operator action. Addition of the time delay relay contacts around the
- low pressure permissive allow for the actuation of ADS as described above. The proposed logic change retai'ns all of the existing design features of the pump start. logic, and also allows the low pressure pumps to start and respond to the additional events outlined in the analysis of the ADS mod. Section F.1.a above.
vx., .- ,: ; n ma, ..m.a ;u e ,. .- : . . .,. - .u: .
~..
Safety Evaluation Attachment 1 No. Alts( Sheet 'O af 10 G. Summarv In conclusion, the addition of the proposed modifications to the existing logic meets the requirements of NUREG 0707. These changes do not increase the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR. These changes retain all of the existing logic, and do not modify the response of the systems to these accidents previously analyzed. These changes do not create the possibility for accident or malfunction of a different type than any evaludted previously in the FSAR. The additional bypass timer relays and bypass timer contacts have been integrated into the existing logic channels, maintaining the redundancy and diversity which protects these systems from single failures. These changes'do not reduce the margin of safety as defined in the basis for any technical specification. Since the modifications do not change the response of the systems to the analyzed accidents, the margin of safety as defined in the basis for any technical specification is not changed. Therefore, these modifications do not result in an unreviewed safety question. W 4 . . ) : ,
%: x se. ;. . w .. c.;; mswe . . s%: n :. m ..c 's : M ' .c ' . s. . - a = v .a . Attachment-2 5afety Evaluation . . _ s..-.-._..- _. No 2.12.4 Sheet l of f
- PNPS-FSAR vis t! " ~.f**M RIq aly For large breaks, the vessel depressurizes rapidly through the break without assistaryce. The signal for the reitef valves to open and g remain open'is 6'ased upon simultaneous signals from: (1) drywell r rep (2) reactor vessel low. low water level. (3) high asequate
> ss.u, discharge pressure on one of the LPCI or coreFurther spray pumps, and (4) 120 sec delay timer completes timing cycle. descriptions of the operation of. the automatic depressurization feature are7.4, found in Core Section 6, Core Standby Cooling Systems, and Section Standby Coollag System Control and Instrumentation.I equipment The Automatte designed as Class in Depressurization System is accordance with Appendix C. A manual depressurization of the nuclear systes can be effected in the event the main condenser The is not by steam generated available as a heat nuclear systes sf m sensible and after
- reactor shutdown.
the core decay heat is discharged to the suppression pool. The core 1s reflooded by the low pressure CSCS. The relief valves are individually operated by remote manual controls from the main control roca to control nuclear systes' pressure. . The number, set pressures, and capactties of the relief valves and safety valves are shown on Table 4.4-1. l 4.4.5 Safety Evaluation _
~
The ASME toller and Pressure Yessel Code requires that each vessel'
- designed to meet Section III beAprotected from pressure la escess of peak allowable pressure for upset 'l the vessel design pressure. .
percent of the vessel design pressure is allowed by l 9 conditions of*110 the code. The code specification for safety valves requires that. safety valve be set at or below vessel , design (1) the lowest - pressure, and (2) the highest safety valve be set to open at or ,G below 105 percent of vessel design pressure. ' The relief valves are set to open by self actuation (overpressure l safety mode) at 1,104 to 1.126 psig and the safety valves are set to operate at 1,227 tosafety 1,253 psig. This satisfies the ASME Code specifications for valves since the valves open below the 1.250 psig nuclear systera design pressure and below 1.313 psts (105 percent of nuclear system design pressure). Safety valvo ceatity is deterstned.by analyzing the pressure risestoppag
** Leccepanyl.ag the cia.i n steam f. low The closure with .the reactorassumes initially operating the reactoratis1,998 MWt.Dy'an shut down analysis hypotheticallyReference 1 describes the reasons for . choosing Indirect flus scras. of applying upset condition 11alts to this event, the conservatism and sodels and methodology used in the evaluation the event analysts,The analysts is repeated for each reload cycle and the of this event. The analysts indicates that the results are given in Append 1r Q.
design capacities of the safety valves and relief valves are capan,le of maintaining adequate margin, approximately 75 psi below (1,375 the peak pstg). The A5ME Code allowable pressure in the nuclear system g AAt Revhion 6 - July 1986
..~x x. sn .. . :.a , . ....
. n. ~ u .a u. . . .. , ::. - .-
Attachment-3 Safety Evaluation No 'Z-124 yq.c .3ag Sheet i of 't. purp . ..' 'an that thay relieve pressure by inherent mechanical (overpressure) action er by action of an electric poe matic control sys teer. The relief by mechanical action is initiated inherently by an overpressure condition in the nuclear systes. N depressurtsation by automatic action of the control systas is employed to re&ce nuclear system pressure so that the core spray and LPCI systems can inject water into the reactor vessel Atring a IDCA when the NPCIS is inoperable. The automatic control and instrumentation equipment for the automatic depressurisation mode af relief valve operation is described la this section. The control
- system, which is functionally illustrated en Figure 7.4-7, consists physically of pressure and water level sensors arranged in trip systems that control a solenoid operated pilot air
. , valve. h solenoid operated pilot valve controls the pneumatic pressure applied to a diaphrap actuator which controls the relief
.,Q.gvalve',.directly. &a accumulator is included with' the centrol
,, . , ; f ',; eM=-at . :for each relief valve to store pnematic energy for relief
. , g ' ,, valve',aperation. The accumulators are sized to provide sufficient
. : a'ir i a ., .Y.f.Damra~for!.a minisian a1'faii sgply of accumulator.
to the twenty pilot actuations
- Cables fromfollowing failure the sensors .1 of madthe ,
,'yy.g: b A. ih' t, e-ithe',' centrol .~ roes 'wbere Ethe logic arrangensats'are formed.in r..L 3.. .? .W.'.J* N;.following;.manderV;'.,the . . cab *inetsMJbe electrical equipment of control ads Logiccircuitry
- 6. . V .- '., .. A"wi~thout'adtomatic transfer. The equipmaat of ads Logic 3 ' is en
& is placedisenpowered Battery . by de e in th W *
".",Sa'tteryit' automatic' transfer to Battery 1
'" .l ' BattTorf s. .(sith ..aabrefore, 'less of any battery affects ese 120 onlysec ,gea loss of
, . . - .' timing ,cifcuit. Eleetrical elements in the control system aaergise
./ td cause opsning of .the relief valve. Each relief valve eis '. powered i ~!dc? from.* either station battery .through ..sutamatic.,.tranafor ' '
.3."7.s. .:4y . . ,wi tekei..x :. .. - .
- 9 .: . . c :P ;. ..,~.:
.& '. . . . M s ' U c..
. # .? .c. ,
v .
~~ ~~
4; ."W} Automatic Depressurisation System
?Z,~'.7,4.3.3.2 Initiating Sipals'-
,' and Logic ,.
Two initiation .sipals are used for the Automatic Depressurisation 1 system l 1. Reactor vessel low-low water level
- 2. Primary containment (drywell) high pressure W .> /f
/f 4 r ):d open.
S Afm ter + thesa @h .W = M-9 t d' A comic'rtr*1
* * = ea .
posest, th3re is a 120 see e rttime 1.. ! ; ; L a to le UJNfl delay to permit the HPCI to restore veter level before the relief valves are actuated. Reactor vessel low water level indicates that the fuel is in danger of becca.hg overheated. This low water level condition yould normally not be sustained unless the NPCIS failed. Primary containment high pressure indicates that a breach in the nuclear systes process barrier may have occurred inside the drywell. Sf.AT iC --> = g,. -. After receipt of both initiation signals, and after me 2 min degvelavra f g u ff e.- provided by timers, tie solenoid operated pilot air valve as energized provided that at least one LPCI or core spray pump is' I . 7 d-10 Revision 2 - July 1983 .
- n. w - r- - .- . .w. w , m u. . w: = . + .
w = Ta :- - - ~ .. ,
-. . . _ _ _ . _ . . _ . . . _ _ . , m- __ _
Safety Evaluation Attachment-3 No Il24 rs - Sheet 'l of P t *&Sftd '
* *N 6' running'. An in'... [ is provided in each trip system in order. to b give reassur ance that low pressure core coolant is available before w the ADS actually permits depressurization of the reactor vessel.
These pressure permissive interlocks are desiped to meet the requirements of single failure and separation. Two pressure switches on the discharge of each core spray and each LPCI pump (12 total) are connected through relays in redundant groups so that each ADS trip system is blocked from actuating unless at least one low pressure pump shoes verified discharge pressure. These pressure switch relay ' circuits are monitored continuously during normal station operation so that if any pressure switch circuit gives a false si pal of the presence of pressure in the low pressure systems. an annunciator immediately alerts the o so that the malfunction can be L'*"' ierrehted.ss.t wr*&n!".se s*perator hte the A ~= kas I t M
- tr2 uel s'a swaterfs emovf 9%t Vtshwt
=ar me 9%s b l* +,7ske n h* ett 6mp + sse nr r= =rs non sos v.
- rgiration of the solenoid operated pilot valves allows pneumatic pressure from the acemulator to act on the diaphrap actuator. The diaphrop octuator is an integral part of the relief valve and expands to hold the relief valve open. Lights in the main control roce infons the main control room operator whenever the solenoid operated pilot valve is energised, indicating that the relief valve is open or being opened. ,
a two position owitch is provided in the main control room for the remote control of each relief valve. The two positions kre "open* . and " auto". In the 'open" position the switch energises the solenoid-operated pilot valve, which allows pneumatic pressure tm .be applied to. the diaphrap actuator of the relief valve. This allows h the main control room operator to take manual action independent of the automatic system. Wropriate ambers of relief valves can be manually opened in this manner to provide a controlled nuclear system cocidown under conditions where the normal heat sink is not available. In ' auto" position, the valve is contrtiled by the ADS provided for the reactor vessel i logic. Manual reset circup low-low water level and F -v & ga - -
- hi essure initiatine l .
sipals. By manually resetting these sipal delayJtimers are n/d % h'Sh er lis, eq mes rec The operator can use the reset switc_hes to delay or,de p e.n p m N d'5 n automatas w nia, n sm relisi valves? if such delay or %r.s M D ", $ _ ycled. lag m g 2orga},,,a,e,,tuationonone ADS ' Reset" button The second vdu,ps..;/NreveRio),t.he recycles Emer, one pghe wo tr:* s
,V,ystems rs must. be reset in
/ " Reset" button resets the secondg ames ,
order for the operator to delay e m automatic activation of J f.hese valves. .
. a e ., e: ..
The logic scheme used for initiating the ADS .;ystc2 is a single trip system containing two trip system logics as shown on Figure 7.4-6. Each trip system logic can initiate automatic depressurisations when the logic in,that trip system is satisfied. Each trip systes logic includes a timer that delays the openihg of the relief valves. This [fi S f a H allows time for the urer to restore water level before the relief - valves are actuated.V The AD5 trip system is de powered. L a5r / t_ #
- b. Instrument specifications and settings used in the original plant safety analysis are listed on Table 7.4-2. The current instrument, d
7.a-11 Revision 2 - July 1983 l
.,,/,
- x ; - . a.. - . .. .. . . . , u . .,. w . ,. ..u, ;', c : . ;a ---
.n.. .'. ~n .
Attachment-3 c Evaluation clio fety/21 4
'=
,s Sheet 3 of13 settisgs are listed in the Technical specifications referenced in appendix B. The wiring for t.he trap systems is routed in separate conduits to reduce the probability that a stagle event will prevent automatic opening of a relief valve. Pump discharge pressure switches are used to sense that the core spray and 1.PCI pungs are running.
l The reactor vessel low-low water level initiation setting for the automatic depressurization system is selected to open the relief valves to depressurize the reactor vessel in time to allow adequate cooling of the fuel by the core spray and 1.PCI systems following o LOCA in which the other makeup systems, Feedwater, RCICS, HPCIs fail to maintain vessel water level. h primary containment high pressure setting is selected to be as low as possible without inducing spurious initiation of the ads. 7.4.3.3.3 Automatic Depressurisation system Initiating Instrumentation The pressure and level switches used to initiate the ads are common to each relief valve control circuitry. Reactor vessel low water level is detected by four switches that osasure differential pressure. Primary containment high pressure is estocted by four pressure switches. yng fr/ fuh rf thd f 4 Hi pyJWon f key ftThe f**yed* I*k Two timers, are used in the control circuitry for each relief valve. The delay time setting before the &Ds is actuated is chosen to be long enough so that the MiPCIS has time to start, yet not so long that the core spray system and LPCI are unable to adequately cool the fuel if the NPCIS fails to start. An alars in th main control room is annunciated every time either of the timers is timing. Resetting the l ads initiating trips - reactor vessel low-low water level and primary containment high pressure - recycles the timers. - "b \ Automatic Depressurisation systes &larms N 7.4.3.3.4
& temperature element is installed in a thernovell in the relief valve discharge piping several feet from the valve body. The ten:rsture element is connected to a multipoint recorder in the main control room to provide a means of detecti.ng relief valve leakage (uring station operation. When the temperature in any relief valve dischsr,p pipelin. ,gri nds ;tetst value, an alarm is sounded in the main control room. The alarm setting is selected far enough above normal rated power temperatures to avoid spurious alarms, yet lov enough to give early indication of relief valve leakage.
l Additionally available are individuhl valve displays (acoustic monitors) located in the control room. These displays provide a means of
- determining the status of each of the four relief valves, RV-203-3A, 3, C, and D, and also the status of the safety valves RV-203 4A and 43. The open/close indication is made possible by the installation of acoustic transducers on the discharge piping of the relief valves RV-203-31, 3, C, and D, and on the bodies of the code safety valves RV-203-4A and B. When the valves are open, indication 6
T. 4 - M &nnm') 194YIW2
. , , s. -. . : .: ,, . . , ww . -
. 2 + .. . a. .
. . . v. .
Attachment-3 Safety Evaluation Nu LI2.4 rups- m a '- 4 x /3 is provided by means of indicating ,. :. en the safety and . relief valve monitors. An audible alarm will also sound if any of the valve + open. There are 10 indicating lights for each relief valve,
" which illu' minate sequentially to give an indication of volve opening as indicated by noise and vibration induced by the steam flow through ,
the valve. - Panels, located outside the control room, are also available to remotely operate the relief valves. 7.4.3.3.5 Automatic Depressurization System Environmental Considerations The signal cables, solenoid valves, and relief valve operators are the only items of the control and instroentation equipment of the ADS that are located inside the primary containment and must remain functional in the environment resulting from a 14CA. These items are selected with capabilities that permit proper operation in the most severe envirorsment resulting from a design basis 14CA. Gammy and neutron radiation is also considered in the selection of these items.
- Other equipment, located outside the drywell, is selected in consideration of the normal and accident environments in which it muct operate. _,
7.4.3.4 Core Spray System Control and Instrumentation 7.4.3.4.1 Identification and Physical Arrangement The core , spray systems consist of two independent spray 2eops as illustrated on Figure 7.4-8. Each loop is capable of supplying sufficient cooling water to the reactor vessel to cool the oora adeguately following a design basis 14CA. The two spray loops are physically and electrically separated so that no ' single physical ; event makes both loops inoperable. Each loop includes one ac motor driven pump, appropriate valves, and the piping to route water from "' the suppression, pool to the reactor vessel. The controls and instrumentation for the core spray systems include ,the sensors, relays, wiring, and valve operating mechanians used to start, l operate, and test each system. Racept for the check valves 9A and 93 in each spray 2oop, which is inside the primary containment. the, sensors and v.31ve closing mechanisms for the core spray systems are located in the Reactor Building, Cables from the sensors are routed l to the main control roca where the control circuitry is assembled in electrical panels,. Eac,h :org, spray 90 is powere4 from a different ac bus which is capable of re'eivang standby power. The power supply for autoratic valves in each loop is from the same source as that used for the core spray pop in that loop. Control power for each of the core spray loops cones from separate oc buses. The electrical equipment jn the main control room for one core spray loop is located j in a separate cabinet from that used for the electrical equipment for the other loop. 7.4-13 Revision 5 - July 1985 l
mam,7-- : :w:' = ';.se :z .u.s. u i s ;.n v .u. : . :n . - .m ~ .. ver . .2 ; Attachment-3 Safety Evaluation
- --- - .-- ._ - _. .. g ,.
Shea
.~;.v;.-7pJt 7.4.3.,4.2 Core Spray System Initiating Signals and Logic The control scheme for the core spray systas is illustrated on Figure 7.4-9. Trip settings used in the initial plant safety analysis are given on Table 7.4-3. The current instrument settings are listed in the fechnical Specifications referenced in Appendia R. l The overall operation of a system following the receipt et an initiating signal is as follows:
- 1. Test bypass valves are closed and interlocked to prevent
*Pening
- 2. If the preferred offsite ac source is available. the oore spray pumps in both spray loops start lamediately
- 3. If the preferred source is not avellable, the core spray pop in both spray loops starts 1/3 see after standby power becomes available,to the p o p
.- 4. When reactor wessel pressure drops to a preselected Islue, valves open in the psme discharge lines allowing water to be sprayed over the oore Two initiating functions are used for the oore spray systes: reactor c. .
i vesselDow water level ooireident with reactor low ~ pr. essure and I
~
primary containment (drywell) high pressure. Either ialtiaties signal can start the systems. , , l w .s,a ' Rector vessel amt water level indicates that the enro7 I' s' is danger of being overheated due to the loss of ecolmat. , ,Drywell high pressure indicates that a breach of the nuclear ,'rystem process - barrier has occurred inside the drywell. The reactor
- vessel... low-/.y .
- water level and primary containment high pressurs' settings and .the '
instruments that provide the laitiating signals are selectodyend , arranged so as to assure proper system operatien eithout ta& acing F ' * ".T . A spurious systes startups. - .nft/f p [ Keylocked switches have been installed to permit blockage'.of the These switches 'are drywell high pressure initiation signal. primarily for use undee yest-LCCh conditions to permit shutdown of the applicable core spray pop motor without affecting tha' reactor vessel low water level initiation signal. ,
....... .s .s The schase used for initiating ocek ura apray sptim A typical oore is a trip system containing decision making logic circuits. The spray system trip actuation logic is shown on ylgure 7.4-6.
decision making logic in a trip system can initiate ' core spray equipment in one core spray loop. The trip systems are powered by reliable independent de buses. 7.4.3.4.3 Core Spray System Pump Control The control arrangements for the core spray pumps are shown on Tagure 7.4-9. Both purrps are automatically started without delay if 7.4-14 Revision 2 .7uly 1933 8
- w. wa n .. < e ':ax : c a u.x v+ /.m ~ . ' . : .c : '?. ;; , x m ::-- v:m ~ x. ..
--~ - --. -- . . . .
Attachment-3 Safety Evaluation
":'S-FSAR No 2124 Sheet g of **
z the preferred (offst te) ac source is available. Each , pump CEA ee manually " controlled by a main control room remote sultch or the b W automatic contfoi system. A pressure transultter on the discharge pipeline from each core spray pump provides a signal in the main control room to indicate the successful startup of a pump. If a core spray _ initiation signal is received when the preferred ac source is not available, the core spray pumps start 1/3 sec af ter the bus is energized from the standby ac power source. The core spray pump motors are provided wtth overload protection. Overload relays are applied so as to maintain power as long as possible without lamediate damage.to the actor.s or emergency power systes. j Loss of voltage trips are provided with time delays sufficient to permit automatic transfer from the unit aus111ary transformer source to the startup transformer source (preferred offsite) without tripping the' pump power supply breaker open. ! i Calibration and testing of the overload trip relays provided for , these motors is accomplished by passing a test current through these
. protective devices to verify set points and relay actuation. This test current is measured with fleid standard asseters. Current or voltage is measured with fleid standard ammeters and voltme'ters.- ,.~.
. : ".?
The actors are protected by long time induction overcurrent relay - elements and by low-set and high-set instantaneous 'overcurrent elements for overload and phase faults and by ground sensor relays for ground faults. ..
.. a .o.c .n.; , ,
The long tisie, high-set, and ground sensor elements are iset'in .
. general accordance vith recommendations in the IEEE. Induction Motor .. .
Protection Guide No. 288. November 1968. The setting of the . low-set '. ., . element is not covered in the Guide. . :
.l .',... -
The long time element is set at 115 percent to 125 percent .of rated . motor current with a time delay set about twice rated actor starting. ..' - time. The long time element is used for overcurrent annunciation.and . in series with the low-set instantaneous element, set at 'about twice ."7. ; rated motor current, it is used to. trip the actor circuit breaker for overload protection. This design permits continued motor operation . under amergency loading conditions wille alerting the operator to a nominal overload condition.
~
The high-set instantaneous element /wNes anort circuit protection and is set at about ten times rated motor current which is camp.atible with system minimum phase fault current capacity. This set point is higher than rated locked rotor current with a margin for inrush current and current asymmetry. . The ground sensor relays are instantaneous relays operating from ground sensor current transformers. The relay setting typically provices a 30 to 1 margin of ma:Imum ground fault . current to relay pickup when operating from any of the station service transformer sources. This setting is high enough to prevent relay pickup for t ground faults when operating on the diesel generator source. 7 A.1E eewitim 7 .1nIv 1983
w.xm:.w~ . a.+; re. : a ..u. w w.> . .. ~ : = . w: . . x - - - ~ w - -
. i Attachnent-3 Safety Evaluation No 31% 4 Sheet 7 of /3 Flow seasuring instrumentation is provided in each of the core spray )
pump discharge : lines. The instrumentation provides flow indication in the main control room. 7.4.3.4.4 Core Spray System Valve Control Except where specified othervise, the remainder of the description of the core spray refers to one spray system. The second core spray system is identical. The control arrangements for the various automatic valves In the core spray system are indicated on Figure 7.4-9. All.aotor-operated valves are equipped with listt and l torque switches to turn off the valve actor when the valve reaches the limits of movement. Each automatic valve can be operated from the main control room. Upon receipt of an inttlation signal the test bypass valve is I interlocked shut. The core spray pump discharge valves are automatically opened when .auclear system pressure drops 'to a . preselected value; the setting is selected low enough so that the low
~
pressure portions of the core spray systes are not overpressurized, yet high enough to open the valves la time to ,providelAdequate .., ..e ~. > cooling for the fuel. Two Eltherpressure switchswitches can initiate are. .used.. . opening'of tor. monitor..;3@'@ WY
*.'. J-nuclear systes pressure.
discharge valves. The full stroke design time of the' pump.111schargel o.O ." :. . . valves is selected to be rapid enough to assure proper'. delivery of y'. r"c' W - ; water to the reactor vessel in a design basis accident.1 'i.The its11g.4,MM ; stroke design operating times are as'follows: . ]' v. F r'v;.kgdbrhf~9);
. .:... A..U#y ' - -
Test bypass valve 45 sec. . ./ ' Pump suctic!n valve 90 see . ,11,.fk' ,. Pump discharge valves 18 sec *S.g.[ . y , j, . Y3t7, wr.4:
. .g. -
7.4.3.4.5 Core Spray Systee Alares and Indications..v, ,; .
- gY[ 3.g,t.y -
Core spray systes pressure between the two pump discharpe .. _
. > . .. :, -r,pH eakage . .i
.Q. d'3 .'. .
l I aonitored the nuclearbysystes a pressure into the switch core spray to persit detection systes outside of.,a {he e pr,laar.y,p' ,{,i@ f containment. A detection system is also provided to :. con't)nuously ,LI.c. u. / '/ confirm the integrity of the core spray piping between the,lailde,Dfjdy'/,';~ ' A differential pressure % r - . the reactor vessel and the core shroud. ; switch seasures the pressure difference between the top of the ' core * ' support plate and the inside of the core spray If the core spray sparger piping is sparger, pfpe just ~ .' '- outside the reactor . vessel. 4 a sound, this pressure difference will be the small drop If integrity isacrost lost, this the core resulting from interchannel leakage. ;7' pressure drop will also include the steam separator pressure drop. . An increase in the normal pressure drop .inttlates an alars in the Pressure in each core spray pump suction and main control room. discharge is ~ monitored by a pressure Indicator which is focally mounted to permit deterstnation of suction head and pump performance.
- 7. % IG 2e wsin 6 - M .y jqp(,
I
- w. u.m . w. a , . , , . a .= w ; . ~. w. .. x . : . . .i . . ., : : - ~ x~ ==
. .- : . :. ..a-._ -
Attachment-3 Safety Evaluation PWPS-FSAR No 2. l 2. 4 Sheet 8 of /$ 7.4.3.4.6 Core Spray System I:nvirorbental Considerations , There' are -no control and instrumentation components for the core .
- spray system that are located inside the primary containment and 'All that aust operate in the environment resulting from a 14CA.
components of the more spray system that are required for system operation are outside the drywell and are selected in consideration of the normal and accident environments in which they must operate. 7.4.3.5 tow Pressure Coolant Injection Control and Instrumentation 7.4.3.5.1 Identification and Physical Arrangement Iow pressure Coolant Injection (LPCI) isBecause an operating the LPCI mode system of the is Residual Neat Removal System (RHRS). designed.to provide cooling water to the reactor vessel following the design basis 14CA, the controls and instrumentatico for' it are discussed here. Section 4.8 describes the WOt3 la detail. * ?. Figue 7.4-10 shows the 'antire residual beat removal system The fo!!owing list ofincfuding the equipment used for LPCI operation. components for .which esotrol ~1 .er equipment itemises essential ~ ,. ' . ' . .( ' [ instrumentation is required to operate in the IJCI anda:
..,l u ?. ' y: ' .
J. .(-Q-@....
'S'!M.T.
Four NOTS pumps .~ . ' *
. l-Pump svetion valves (from suppressionTool) t .g ' .jd. ./ , ..
LPCI-to-recirculation loop injection volves
. M-.~;..;n,(7
%f Q , ".. k. T } j + .,@
Recirculation loop valves ,
.?,.
~ ,
the control .p .' .'.,.{. ,' This is angessar tin ' ensure Jh r-The instrumentation .for LDCI operation a.~ circuitry for other valves in the NHRS. that the water pumped from the , suppression.. pool by'.;the', - provide routed directly to a reactor recirculation. loop..: These/ interlock. ogactorf.V features recirculation looparevalvesdescribed are described inla this Section. - The this Section,)ecaus55 .%./,y? . actions o actions are accomplished to facilitate LPCI - operation.U ',' sG.j e , .r. -
. , m '.p,*; <r.;
identical' pump . subsystems, 4ach ,,subsystan f.lll. . , 17CI operation uses two The two subsystems are,' arraoged- to V . ' ' with two peps in parallet.
- discharge water into different reactor recirculation' loop Figure 7.4-10 shows the locations of instroents, control Except equipment, for and LPCI components reistive to the primary containment.
2001-68B and ;"40 Josetor ! valves 1001-68A, the LPCI check recirculation loop tweps and velves, the compecents pertinent .to LPCI
/
operation are located outside the primary containment. ' The power for the RHR system Each pwaps pair of pumps is supplied la each from ac buses that can subsystem receive standby ac power. Notive power for the receives its power from a different bus. injection valves on bothControl sidespower used for during the LPCI LPCI ope alte rnate standby power sources. cor ponents comes from the de buses. Redundant trip systems are 7.4-17 Revision 5 - July 1985 l.
c- WN K% , ' . - . ~ . .. .a ;, ...s. .. . . . .. ..: ~.. .,L . m. . :: '
---h% t s3 Safety Eval ation PNPS-FSAR No 2.12.4 W< 9 of /3 br 6,ome povereo from diff aront de busses. :The use of canonon 6 .
of the LPCI components is acceptable because the oore spray rfstems and LPCI operation are arranged independently to accomplish the same objective: ; provide adequate cooling for the fuel at low muclear system pressure following a design basis accident. , LPC1 is arranged for both automatic operation and remote aanwl operation from the main control room. The equiseent provided for annual operation of the system allows the operator to take action independent of the automatic controls in the event of a EACA. 7.4.3.5.2 LPCI, Initiating Signals and Zagic Tr.e overall operating sequence for LPCI followiog the receipt of an initiation signal is as follows:
- 1. If the preferred (offsite) ac power is available, the four pumps start simultaneously with no delay, taking suction from the suppression pool. The valves in the suction paths to the suppression pool are asintained open so thit ' no automatic action is required to line up suction ,
~
- 2. If the preferred source of ac power is' .aotl available,'['eee ,- -
;,. ~.
pump in each subsystem starts after a 5 sec;delsjr, after the' . standby power source is operating. The ,~ second. pump.*ia'anch"Msa.. w, .. subsystem starts after a 10 see time . :' Y. . delay
.. !.q@2flQ@pfg%5;f.'.p.jg. 1 ,
3. If the accident has not resulted from, rup 3 3 for water injection ,
. ( 4. + s @ fp ' W "
'- \
.. has resulted . fr;om;:y.>.
. . i 2.' .q.?. ,- )
- 4. If the accident ruptur ,,,
i reactor recirculation lines, cithe,-Q'JCI . [M,'p! identifies the damaged loop f, .d" :;4
' ' ~
4
- 5. The recirculation ymp discharge lve*h:imstfrTall os'eh ed .~anRN$fj[:<..$ '
recirculation loop ~tuta y :. reaetor > recirculation pumps are tripped .i:' ?.'.M.i'd. . :. tF.Y ! y. iv.- ). . e '- ,,
-4. i.
? .81 N*g,7= Q v .% . ,,
- 6. Valves in the LPCI systes respond' automaticaMy,s. e h.1-t.The
. ;,w. ..?q[.;,
water pumped from the suppression chamber -1s' routed .to'the , j ..#.,f' undamaged loop
' #'M?M%ss'.f . '.'. ;
- , . . f.: M .t,.e p .
- 7. When nuclear system pressure has dropped to' 'a predetermined, . ,
valves . to ' the .' undamaged . . value, the LPCI injection recircul'ation loop automatically open, allowing ..tho' IJCI , p'umpstoinjectwaterintothepressurevessel.j.'.:',;,g*.,j . i ..- 8, The LPCI system then delivers water to the reactor vessel .
)
via that recirculation loop to restore water level and l provide core cooling l l Figure 7.4-10 shows the locations of sensors. Figures 7.4-11, l l 7.4-12. and 7.4-13 show the functional use of each sensor in the 7.4-18 Revision 2 - July 1923 e
w w s :. . .r e , . x.<w .m . ~ ,, :u:a. . w- wm .. .mm....
. .. .. . . ... .... . .- . . . ....c.-.__ --. . ..
Attachment-3 Safety Evaluation No 112.4 control circuitry for the various LPCI components. Instru[ ment characteristics .and settings used in the initial plant safety analysis are given on Table 7.4 a. The current lastrument settings h F are listed in the Technical Specifications referenced in Appendix 8. . Two automatic initiation functions are provided for the LPCI: . reactor vessel low-low water level coincident with low reactor Reactor pressure and primary contalnaent (drywell) high pressure. vessel low water level indicates that the fuel is in danger of being overheated because of an insufficient coolant inventory. Primary containment high pressure is indicative of a break of the nuclear
\ syste>m process barrier facih na h ywell. '
[d (t'/ The instruments used to detect reactor vessel low-10w water level \ coincident with low reactor pressure and primary containment high pressure are the same ones used to lattiate the other C5CS. Once an l f altlation signal is received by the LPCI control circuitry, the signal is sealed in untti manually reset. The seal-in feature is
' l shown on Figure 7.4-11. .
Keylocked control switches have been installed to permit blockage W the drywell high pressure initiation signal. These sultches are primarily for use under post-LOCA conditions to perett shutdown of the applicable RHR pump actors without affecting the reactor vessel low water level Initiation signal. The scheme used for Initiating the LPCI system and the recirculation ~ loop selection logic is a trip systes containing two decision making logics. A typical LPCIS trip systes is shown on Figure 7.4-4. D Elther of the two decision making logics can inttf ate the LPCIS. .The trip systes is powered by de buses.
~
7.4.3.5.3 LPCI Pump control { The functional control arrangement for the pumps is shown on Figure 7.4-11. The reaction of the pumps If the to an tattf ation signal preferred (offsitM ac depends on the availability of power. source is not available, the four main systes pumps automatically start in a timed sequence when the standby, ac power source becomes available. If the preferred (offsite) ac source is available. the four pumps
- start simultaneously ytth no delay. Only three of the four RNR pumps are required t,o provide adequate flow to restore reactor vessel water level for the design basis LOCA. The time delays are provided by timers which are set as given in the Technical Specifications referenced in Appendix 8. The timers prov)ded in the LPCI circultry for the RHR gumps, as well as those used for the LPCI injection valves are capable of adjustment over a range of 1.5 tfees the specified setting listed on Table 7.4-4.
in the pump discharge pipelines Pressure indicators installed upstream of the pump discharge check valves, provide Indication of 9 proper pump operation following an initiation signal. A low pressure V ].4-I9 Eevet o'ch '2. -Jut y 1995 e . .. . . . , i,.1 v teat
p r. . .m w.s .nu a n .v w.m u.m :- :: x = . v e a- i a : --- w.wv me m. Attachment-3 Safety Evaluation No 2.13./ PILGRIM FSAR Insert "A - Page 7.4-10 (New Paragraph) When low-low water level is sensed, a high drywell pressure bypass timer (0 to 30 minute adjustable) is initiated. If drywell high pressure is not sensed before the selected time has elapsed, and if the low-low water level signal is still present, the ADS valves will be signalled to open without high drywell pressure. (See Figure 7.4-7) . Insert "B" - Page 7.4'-12) (New Paragraph) A manual " inhibit" switch in each of the two trip system logics allows the operator to prevent automatic depressurization. This switch is key-locked in the " normal" position to prevent inadvertent operation. An indicator 1111ht for each switch is illuminated when the switch is in the " inhibit" posit:on. An annunciator in the control room alarms when either switch is in the
' inhibit" position. The inhibit switch does not break the seal-in logic and uill not terminate an ADS blowdown once it has begun.
Insert "C" - Page 7.4-12 (New Paragraph) , Four additional timers (0 to 30 minutes adjustable), one for each channel of the two dual-channel trip system logici, provide bypasses of the high drywell
-- pressure system initiation signal. These bypasses permit automatic system initiation without high drywell pressure. The delay-time setting can be chosen to be long enough to prevent blowdown on temporary reductions in water level but not so long as to permit the water level to become dangerously low.
An alarm in the control room annunciated when any one of the high drywell pressure bypass timers is timing. The timers are reset. automat cally whenever the water level rises above the low-low setpoint. The bypass timers are also - reset manually whenever the reset pushbuttons, one in each of the two trip system logics, are depressed. Insert "D" - Page 7.4-14 (New Paragraph) The core spray system can be initiated by low-low water level alone, without reactor low pressure or high drywell pressure, after a selected time delay (0 to 30 minute adjustable). The timing function starts when the low-low water level setpoint is reached. The timers are reset automatically if th,e water level rises above the setpoint before the selected time has elapsed. The timers are also reset manually when the ADS reset pushbuttons, one in each of l the two ADS trip systems, are depressed. i
---J__-----_______L_.-_
%mm:; ::a ~.L., .<;c. . : a L ,.a , N .. u : n a
. ., 'L:.J: . m . .L ai'. : hi L mm% . . c----
\ . . Attachment-3 Safety Evaluation No 2124 i~ il of ' PILGRIM FSAR , , 2nsert "E" - Page 7.4-19 (New Paragraph) LPCI can be initiated by low-low water level alone, without reactor low pressure or hi@ drywell pressure af ter a selected time delay (0 to 30 minute adjustable). The timing functions starts when the low-low water level setpoint is reached. The timers are reset automatically if the water level rises above the setpoint before the selected time delay has elapsed. The timers are also reset. manually when the ADS reset pushbuttons, one in each of
.the two ADS trip systems, are depressed.
f nsert "F" - Page 7.4-11 (New Paragraph) Keylocked switches have Leen added to permit plant operators to disable the automatic logic. This manual action will be displayed on the control panels by indicating lights and it will be annunciated. These switches allow the operator to inhibit ADS per the instructions in the Emergency Operating
~
Procedures. Insert "G" - Page 7.4-10 (New Paragraph) The bypass arrangement increases the range of events over which ADS will respond. Events such.as a break external to the drywell or a stuck open SRV do not necessarily cause a High Drywell Pressure signal. i Insert "H" - Page 7.4-11 Each logic channel also contains a bypass timer, which allows automatic depressurization with low-low water level only, af ter a predetermined time has - ; passed. An annunciator indicates that the bypass timer is running and that a low-low water level signal is present. i I O e __ __ ___ _ _ - _ _ __.
mw3c;,1;m7- 1;--a;;-- , s ~ .e
, ,,. : O . .i w .
- c. w . ;: ..,u .- . si :-' . - ' ~.
'tt;; H;-t : "#a'o r"=luation
, No 2-124 F~~~~"~~' Sheet 13 af 73 "LWLdLWth ! 75 he ]~ l
.t0 * --
DECISION MAKING LOGIC d-c BUS l 4 LUL:xwLd 'ppp}:}DI::] (TRIP SYSTEM LOGIC) , e I
~
yy u - - - - -s HPCI TRIP SYSTEM ACTUATION LOGIC (1)
}
) c LWLJhc LWL dWL iLWL i WL
' HDd' M' h q M ~ '
OS q h, ' l 1 d c BUS I d tf-o ikd St C
. . , , --- r -
L. - - ADS TRIP SYSTEM ACTUATION LOGIC (1) r---- , . j
- h 1 LWl..LWL..
1 LWL<. 1 LWL.L .. I--- [' T.HDP.1, i d-c BUS - n, I
,.J [l DECISION MAKING LOGIC (TRIPSYSTEMLOGIC)
~
LWL::LWL:: lLWL; LWL; h iDT eB'ywRLF (It"HDP::] l
~ ~ '~
8 U i l (RLPC ACTUA[ , g ,, ,,L,,g ACT]TgLook c
~
LPCI.TR]P SYSTEM ACTUATOR LOGIC (2) LWL.1 LWL. 1 lHD(1,, --1HDP,.,; l
,, SECISION MAKING LOGIC d-c 8US I (TRIP SYSTEM LOGIC)
LWL:/ -- _ wlWL:: hD[:;HDPj;] [l RLP4 i$BDT 4BCg ) =Rt.P g s _ m~ u _ - . . . ., CORE SPRAY SYSTEM ACTUATION LOGIC (2) LE 7N1 Lar -LOW WATER LE EL e -- TSP-HIGH.QRYwELL PRESSN
- 2, ADS-ATJTOMATIC DE'# PRESSURIZATION ,
SYSTEM CSP-OPE CORE SPRAY PW #UNING LPCIP CPE LPCI PW IUNING FIGUR.E 7. 4-6 MLP-REACTOR low PRES $N TYPICAL CORE STANDBY COOLING IS-Inhibit Switch SYSTEMS TRIP LPSI-Low Pressure Pump Seal-In SYSTEMS ACTUATION LOGIC BDT-Bypass Delay Timer PILGRIM NUCLEAR POWER STATION FINAL SAFETY ANALYSi$ REPORT __ ______--__----a_ . _ _
- m. m.m m.. .m, m .m ,, _ . , , ,, , ,_ , , , ... . . - - m.- .. ,_ _
w.m .
- SCOPE AND JUSTIFICATION APPROVAL Now a
, m b ll.- 5 f u n t Blu K. A b;eteI % .,md s to 4-osin.M1/n 3 .. .,m,oc [.[,Ne_ eno.scr s,*u Au.dt.d estream-1 -saavutacy: Pot * --
'*-
- C8"'l'*"' ***' -
j aumG paamr,y. si j
,u -
te ..a occu m ns- ,J l ccgit CF MtEFCAT1dN- (ellach theth Knat99&aFy) [4,r de,, a d/4f.I forerePF e 8d "b bC ,., _31 w,,. wu. si. % +~ , G<-a- e J #a. e,ary*-y MM6 ; ., IW A.iM r. A* bsally conhol/c d drese.l yr re<ahr s1.65 a~// ee arred 4 . ., 6/*<ka + co bh.,s. v ,y17seCkJ.seGGy als4J /**J< da,,,,8
&<e. oktod) ( .mr,icawaw nm ncEwicarm - n. a 4/ 4 ,. 4 4 ,v Ar,y ,'. <4//,J e' ee o&y/ '
r res.aa. +L p.,La:Isy e4 e y lale. sL I,s., Ltuica F' .d k yieads. e.< en.km se es += im nee. 4Cs freinsL: thy of setteek'q l AC fe~ '
- L < a sf,4u s sak.&,
f5AA F9ACT7 WI WCH WlEC CHANIET l pu re SULAT10NAL 8 DES)GN CMTEMA-Oferdl,k ***$ SM!Yt'e$<nce, ytwtedussJ Gone} **edf
*ntT/toena r tsTetATis- CENTmW StBEDIAE M5 tam
$06 #ea ## edt ,,
l @ C,APRR C EXPENg DISOEDUNG ISATBilALS 8 M3 ### fel E ! - m . C OIST RJCTX M 8 G' a d o e (adage) f t. anak'r
& TL5DeMR g,,, g .
Pauk.t CLCEECW jg!f,### - jjf,, p,, se db mm- - e *==ie ta ya u:u I L
. -a: ,=
v a a wS .. * *=re W 587 a r.il .
***= res===ased ?
me fe/dkA/ Q. DIEPAMD SY: Appammtv.M/ Mr - AVarf(t - y A ::dt:, mg
= . nun n.;.av--
e -J-#r, .. s P' -
. ; . 2 . . .m.- i l
-4 ,A ,.A.,,x .
.' Y.A f ").,I 2.!
~" 5 M itMa' M * {f .. O h. $yStroJmf einAET: -l '
.. *) x .. .
GWAE
.Q .
- . u.,~ Au ,f
. .. ,.c h g svi.aufAE. a sors i,~.urerens g.. _- '3 *
~
.5)'.";JJ.-['EI..'*.
~
C IED ACCEPTANCE- GD6eW. TO PIEPAfER .
'i
.~ E: W-LED QA PShCCl'Ist. - 910eh
. C -TAEBEADY DATE ACCEPTED AS ABN. '
8E DFI ~ F ' ' " % ~~ "'" s . L88SFl.m [ 3 TM M MR N TO'"~"*** ' '
- JEDM -T4 . h E s t.L. . . . .- . .
PP .
- ~ ~ " " ' - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ,
m=Tmzmxtu:h;;t;; .i .n .. :.c. .c s. ; x . m :.- .c w.:> . w. n : . 2, : ,u x w,:a 3 5 15-57 4+feck. # / ) Safety Evaluation 28 W Sneet 8 of 9 Poser cabling from the blatkout d'*' (switchgear A8) is routed to the emergency switr T. a the routing established for the shutdo.n transformer as e4scussed in the FSAR, Section 8.3. Therefore, independence is maintained between the blackout diesel generator and the emergency diesels as between the shutdown transform 2r and the emergency diesel generators,
- 7. 11/1 Interfaces All conduit and cable installed by this design change is non safety related, however, all conduit located within safety related areas will be supported in accordance with seismic !!/1 criteria. In addition all drilling of seismic I structures are performed as Q activities.
E Analysis of Effect on Safety Functions
- 1. Secondary at power system The seco~ndary ac power system, the shutdown transformer and the controls of 4.16kV circuit breaker A600 are not safety related. The new blackout diesel gen'erator and associated 4.16kV switchgear are
. also not safety related. The nev 4.16kV bus enables either the diesel generator or the-shutdown transformer to supply power to the emergency service 4.16kV buses a.s required. By this modification, an additional 4.16kV circuit breaker is introduced into the 4.16kV power supply circuit between the secondary of the shutdown transformer and 4.16kV breaker A600. This circuit breaker controls the power from the shutdown transformer to the new 4.16kV bus. The addition of a new 4.16kv (A802) breaker in the secondary supply of the shutdown transformer does not adversely affect the availability of the secondary ac power except under conditions of malfunction (tripping) of the breaker and shutdown transformer overcurrent relay, in such an event, the blackout diesel generator will be available to supply power to breaker A600 through breaker A801.
This new 4.16kv metal clad switchgear will be designed, built, rated and tested in accordance with recognized industry codes and ANSI ~ standards. Thus the overall availability of the secondary power , system is not reduced by this modification. , l
- 2. Emergency service buses A5 and A6 ,
The emergency service buses A5 and A6 will be affected by this modification as these buses can be supplied power from the new blackout diesel generator. The addition of the blackout diesel generator to the exis. ting scheme will not impact the ability of the shutdown transfomer to supply power to the emergency buses during nomal operation, since the blackout diesel generator will not nomally be operating. The blackout diesel generator will be added scheme. to the shutdown transfomer differential current relay The tew differential relay scheme provides trip signals I to 4.16kv i:reakers A600, A501, A601, A801. A802 and the 23kv line circuit switcher.
~
)
l 6
m- - e. g , v . s .u .7 u w. - .. - ,. . . . ; , n w . , .- . s... .w. ,~. . .. , . .. . 5-15-87 A M ac4. 6 / Safety Evaluation 2; y Sheet 9 ofag 4 T/a new differential relay scheme provider t.1 sa..t .oc. of protection for buses A5 and A6 for fault, in the secondary ac power system. The controls for all of these breakers area non-safety related. There are no changes to the safety related portions of the emergency service buses. The ratings of the safety related breakers are adequate af ter the addition of the blackout diesel generator set. The safety functions of these buses are not affected by this modification.
- 3. Control Room Panels C3 and C5 The main control room panels C3 and C5 are safety related panels.
However, the control switches and indicating lights for the blackout diesel generator and associated 4.16kV switchgear installed on panel C3 are not safety related. The differential relay mounted on panel C5, is also not safety related. Even though the changes j incorporated by this PDC are not safety related, all modifications to panels C3 and C5 will be implemented to Q criteria to prevent 11/1 failures.
- 4. 480 Volt Load Center B4-This system does not perform a safety function. The maintenance loads of the new diesel generator will be supplied power from this load center when the diesel generator is not operating. The additional loading on the buses is within the capacity of the load center.
g Summary Based on the preceding evaluation, the addition of the blackout diesel generat.or does not increase the probability of occurrence or consequences of an accident or malfunction of equipment important to safety as previously evaluated in the FSAR. This modification does not create the possibility for accident or malfunction of a different type than any evaluated previously in the FSAR, nor does it reduce the margin of safety as defined in the basis for technical specification. Therefore, this _ modification does not result in an unreviewed safety question per 10CFR50.59. i ( 1 l i
.1
- u. .. ;- .c. ~ . . .:. .a : , : . n m.ra ..m , = - . . ~ ,w.u . , u . a. = . . .-. a . a-
. ; . . , w. v: .>. m.- = i- ww=n=
1.. i
- t.
, . Attachment- 4 Safety Evaluation NJ 2124 Shee+
l OTHER FSAR FIGURE CHANGES Figure 7.4-7 (ADS FCD) Figure 7.3-6 (NBS FCD) . See 24A1719 (ADS MOD SPEC) i Figure 7.4-9 (Core Spray FCD) for details of the changes Figure 7.4-11 (RHR FCD) d 0 9 4 e 4
, 2;J:;s m .= n <:e
,..r :m :' . .. x.. c i .2 s - w :-- a-w ~
= " T : ' ' w w z- - ~ '
i 4.4. k' s
. Safety Evaluation No.: 2.13 I SAFETY [ VALUATION PILGRIM WUCLE AR P(htER STATION Rev. No. O PDC PCN System Calc.
Initiator: Dept: Group: No.: No.: Name: Date: 5%/g7 7 M. W%ske. neb FS 4.M c. erg-75 fj[/ S@DS g%4p ,) d 87-SO4-l Description of Proposed change, test or experiment: Em eo'ekeel Baron LJ ifu'e s 4, *n u 4e SLc a n nt assoc u s s e d Teak ueet spee1 %. + ten ckan e s, SAFETY EVALUATION CONCLUSIONS: , The proposed change, test or experiment:
- 1. (M Does Not ( ) Does increase the probability of occurrence or consequences of an accident or as1 function of equipment important to safety previously evaluated in the FSAR.
- 2. N Does Not ( ) Does create the possibility for accident or as1 function
, of a dif farent type than any evaluated previously in the FSAR.
- 3. M Does Not ( ) Does reduce the margin of safety as defined in the basis
' ' (
for any technical specification. L BASIS FOR* SAFETY EVALUATION CONCLtJ510NS:
. ' .Ser A +4-ek o el %e#+5 Change Change .
be) Recommended ( ) Not Recommended (X ) LE Performed by wN, d Date 4 6ff 7 Exhibit 3.07-A Rev. 3 Sheet 1 of 3 ISSUED F% a ONSTRUCTh!.-a!. p ::: - l i
ci, .r::a.=:x .:.: . . . x. n: .. := x- c. . . - - w .r= ,- c a = m-- - -- - - - - - - - -
^
_,;_ _ . _ p , _ .. ..
-- - - - -~~~
4 9 Safety Evaluation c2G ( No.: () SAFETY EVALUATION PILGRIM WUCLEAR POWER STATION Rev. No. (0 ) A. APPROYAL () This proposed change does not involve a change in the Technical Specifications.
@ This proposed change, test or experiment does ( ) does not h involve an unreviewed safety question as defined in 10CFR. Part 50.59(a)(2).
Q This proposed change involves a change to the FSAR per 10CFR 50.71(e) and is reportable under 10CFA50.59(b). (4 Comments: IN ve/vre M- 8/h9 -t' The safety evaluation basis and conclusion is:
% Approved () Not Approved Ja AYA <*lV/Y7 t'nLI ffAf87 isiE1pline IEtoup LeaderffDite Supporting Discipline Group Lander / Bate S. REVIEW APPROYAL
( Comments: la cra Pu4 2t a.pohn.4h udi Ah
% % Drcctw n d ef borm s o
rA4bb F Y y7 5&SA/)roup Leader / Data h f7 C. ORC REVIEW - () This proksed change involves an unreviewed safety geestion and a request for authorization of this change must be filed with } Directorate of Licensing, NRC prior to implementation.
~
( This proposed change does,not involve an unreviewed safety I question. g ORC Chairman - / 8E4' Date M ORC Meeting Numbe IIY9 cc: Exhibit 3.07-A Rev. 3 Shee g _ ~ )
= - - - - - - _
j ISSUED Fog l [ CONSTRUCTION ----~~=_-..a.,- _: i l l
)
mm;.w=m. w .> ,. . . .. a x .: u- _ w. ,. .. .sm- -c .w +
. -O ' Safety Evaluation No. Ol30 A
%. o . s/A in i
A. Description of Procosed Chance. Test or Exoeriment: This modification replaces the Standby Liquid Control Systems's (SLCS) existing sodium pentaborate solution (natural boron with 19.8 atom percent B10) of 9.4 to 16 weight percent concentration with an enriched sodium . I pentaborate solution (borf5 enriched greater than 54.5 atom percent B10) of 8.42 to 9.22 weight percent concentration. In addition, this modification recalibrates and revises the setpoints of the level and temperature sensing instruments, relocates the SLC pump 2078 test button near the test button for pump 207A and locates a new pressure gauge near the test buttons. The recalibration and revision of level and temperature sensing instrumentation is required, due to the solution concentration change. The SLC's Technical Specification change includes the solution concentration requirements, surveillance requirements, and bases for the new enriched sodium pentaborate solution. B. Purcose of Chanae The enriched sodium pentaborate solution is being added to the SLCS to comply with the NRC's Anticipated Transient Hithout Scram (ATHS) Rule
. (10CFR50.62). The reduction in maximum allowable solution concentration from 16 to 9.22 weight percent reduces the maximum solution saturation temperature from 70*F to 38'F. This reduces the possibility of Technical Specifications requiring reactor shutdown as a result of solution temperature requirements. The additional system changes are being performed to simplify testing and minimize enriched sodium pentaborate loss.
C. Systems. Subsystems. Comoonents Affected: , Standby Liquid Control System - This modification affects the SLCS in the following manner:
- The performance of the system is improved by this modification. The modified system performs at increased reactivity control capacity to meet the NRC ATHS Rules equivalence requirements of 86 GPH/13 weight percent of normal sodium pentaborate solution.
= If the enrichment option was not used, two pumps would be required to meet the NRC's ATHS Rule. With the enrichment option, the reliability of the system is maintained, since only one pump is required to satisfy the NRC's ATHS Rule. The system retains one redundant pump.
The low crystallization temperature (38'F corresponding to a 9.22 weight percent concentration) of the enriched sodium pentaborate solution will further improve the system reliability. This reduces the possibility of reactor shutdown because of solution temperature requirements. ;,. _ _
^ ISSUED FOR CONSTRUCT CN
__:__ =. s
,_ =.s l
mm.a.- - - - . - . , , , . . ~ _ . - . -. .- -
~
Safety Evaluation No. J1' S 3
\ (2.u. c - 6/4 / 87
- The response time of the system is' improved by this modification due to higher rate of B10 injection into the reactor.
- The relocation of the test button for SLC pump 207B and the addition of the pressure gauge will facilitate the system testing and does not affect the safety performance of the system.
D. Safety Function of Affected Systems /Comoonents The safety function of the SLC system is to provide a backup method, which is independent of the control rods, to maintain the reactor subcritical as the nuclear system cools, in the event that not enough of the control rods can be inserted to counteract the positive reactivity affects of a colder moderator (Ref PNPS-FSAR, Rev. 6. Section 3.8.1). This modification has an impact on the safety analysis (Ref. PNPS-FSAR Section 3.8.4) and the Technical Specification Section 3.4 which need to be updated to include the NRC ATHS Rule (10CFR50.62) requirements. - E. Effects on Safety Function The enriched sodium pentaborate modification to the SLC will upgrade the system to the reactivity control capacity requirements of the NRC's ATHS Rule (10CFR50.62) and still provide the equivalent of 700 ppm of natural boron to maintain the original-system shutdown requirement. The low crystallization temperature (38'F corresponding to a a 9.22 weight percent concentration) of the enriched sodium pentaborate solution allows the reduction of the tank heater and heat tracing setpoint to 53*F. This temperature is 5'F above the low' temperature alarm setpoint of 48'F. The low solution crystallization temperature and the new tank heater and heat tracing setpoint will reduce the possibility of reactor shutdown because of solution temperature requirements. The addition of the pressure gauge and the relocation of the safety related test button for SLC pump 207B will not have any adverse effects on the safety functions of the SLC system. Materials for these changes will be procured, installed and tested in accordance with safety related requirements. F. Analysis of Effects on Safety Functions As per GE analysis (see Attachment 2 to this safety evaluation), use of an 8.42 or greater percent concentration of enriched sodium pentaborate (enriched to greater than 54.5 atom percent bio) will meet or exceed the NRC ATHS Rule 10CFR50.62 requirements of the SLCS at Pilgrim Nuclear Power Station. This analysis is based on an injection rate of 39 gallons per minute (Ref.1 & GE Calc. No. DRF C41-00095/2 L4, .Sht.13A, SLCS Volume & Concentration Chart). As each pump of the system has a minimum discharge capacity of 39 gallons per minute, the design is adequate to satisfy the NRC ATHS Rule requirements. The minimum conce io_n of 8.42 percent and aB10 enrichment greater than 54.5 percent pr ides a w urr-wargi3Li] E percent beyond the amount needed to shutdown e reaf& SUED F L CONSTm m,OR v. Y x m .e >
^
=a..-. . .m w a 1.-. w .- . - .. . . . . . , 5 Safety Evaluation No. d'3' - ks e . o - 5/4 Il The upper limit, 9.22 weight percent, concentration of enriched sodium i pentaborate has a saturation temperature of 38'F. To preclude i precipitation, the minimum solution temperature will be maintained above l 48'F, which is 10*F above the saturation temperature of the maxi, mum concentration. In order to ensure a solution temperature greater than 48'F the technical specifications will require determination of the , solution temperature daily. This frequency is considered adequate because i the room minimum design temperature is 60*F, and any temperature change I would be gradual. In addition, the daily monitoring will be backed up by I the tank heater, heat tracing, and low temperature alarms. If the i solution temperature in either the tank or pump suction lines reaches { 53*F, the tank heater or heat tracing will commence operation. If the j solution temperature in either the tank or suction lines continues to drop to 48'F, the operator will receive an alarm in the control room. Technical specifications will then require that the reactor be placed in a cold shutdown condition within 24 hours of a solution temperature less than 48'F. In order to comply with the ATHS Rule (10CFR50.62), the boron in the sodium penta grate solution must be enriched to greater than 54.5 atom percentofBgu. The technical specifications will require that the B10 enrichment be greater than 54.5 atom percent. If the B10 enrichment is found to be less than or equal to 54.5 atom percent, the Technical
- Specifications will require the operator to determine if the original shutdown criteria (equivalent of 700 ppm of natural boron) can be met. If the original shutdown criteria can not be met, Technical Specifications will require that the reactor be placed into cold shutdown within 24 hours. If the original shutdown criteria can be met, Technical Specifications will require that the B10 enrichment be' returned to greater than 54.5 atom percent within seven days. If at the end of this seven day period, B10 enrichment is still less than or equal to 54.5 atom percent, Technical Specifications will require that the NRC be notified within t bring the enrichment into compliance.
sevendayswithBECo'splan5enrichmentislessthanorequalto54.5 This ensures that if the BI atom percent, the operator will shutdown if the original shutdown criteria if the original cannotbemet,orbringtheenrichmentintocompliancy0enrichmentgreapu shutdown than 54.5(p(r)teria can be met. In order to ensure a B r ercent the Technical Specifications will require that the B enrichment be determined prior to restart from a refueling outage or any , This frequency is considered timeboronisaddeg0to adequate because B isthe a storage stable tank. isotope and enrichment changes can only occur when additional boron is added. In addition to ensure that the boron added is enriched properly, station procedures will require that B 10
- f enrichment be determined as part of the receipt inspection before release the matrial for use. Technical Specifications will also require that o{0enrichmenttestresul.tsbeknownwithin30daysofsamplingthe B
material in the Standby Liquid Control Storage Tank. The 30 day time period allows sufficient time to perform the enrichment test and receive the test results. It is considered adequate from a safety point, due to the station procedure requirement to determine enrichment as part of the receipt inspection before release of the material for use. The requirement to determine enrichment after the addition of boron to the storage tank functions as a backup check t Alcn ar_oce,dures. _ ISSUED FOR
] CONSTm a m
mt g
]J
* ." c-,_.
4 ,
m: w. a w n ;.w. .n . . .w:u w. . . . m, . .. x . ....,w .c .n.~ - - - -
\ Safety Evaluation q No. c2 8 b 2 07. o - 5/a /q 7 The storage tank high and lowlevel alarms are being maintained at their original volume setpoints. The high level alarm alerts the operator to a solution volume near the storage tank overflow. The original volume concentration requirements were such that, should evaporation occur, a low level alarm would annunciate before the temperature-concentration requirements were exceeded. For the original solution, the maximum possible attainable concentration at the low level alarm was 14 weight percent. This corresponded to a saturation temperature of 60*F which is less than the original 65'F setpoint of the heat tracing. This ensured the operator was given an alarm before crystallization could occur from high solution concentration. The requirement for a low level alarm to annunciate before temperature-concentration requirements are exceeded is not needed because of the new lower solution concentration requirements (8.42 to 9.22 weight percent). Since the maximum concentration of the new solution is 9.22 weight percent, the maximum possible solution concentration obtainable from evaporation without a high or low level alarm is approximately 10.3 weight percent *, This corresponds to approximately a 44*F solution temperature (low solution temperature alarm setpoints is 48*F). Due to the 53*F setpoint of the tank heater and heat tracing and the design room temperature of 60*F to 100*F, solution concentration changes due to evaporation would be slow. The operator would be alerted to a solution concentration change-from evaporation by either the low level alarm or the technical specification monthly surveillance requirements before the crystallization point is reached.
- Hiah Level Low Level Alarm (9.22) - g (9.22) = 10.3 weight percent Alarm G. Summary The SLCS by itself cannot cause a.n accident and it does not interact with any other system'whose malfunction could cause an accident. Hence, this modification on the system does not increase the probability of occurrence of an accident.
This modification increases the system's control capacity to satisfy NRC ATHS rule requirements. The modified system is more effective than the existing system in bringing the reactor to the cold shutdown condition from rated power. Hence, the modification does not increase the consequences of an accident. This modification does not call for the safety equipment of the system to i work at higher pressures, temperatures and more severe conditions than the existing levels. The modification makes the SLCS pumps redundant and it does not change the logic of the system. Hence, the modification does not increase the probability of the malfunction of the equipment important to safety. l This modification increases the margin of safety for system availability by reducing the possibility of system unavailability from solution temperature requirements. ; This modification increases the margin of safety for flow rate requirements (required 39 GPH; available 78 GPM) and for minimum volume of solution requirements (required 2068 gallons at a mid , range.c.qncentration of 8.82 percent; available 3960 gallonss. ISSUED FOR 4 _ CONSTRUCTS 1 --~r-=. : . . ~.: T '- __=____--_-__c _
;2 ;;7L.y;gq.~ymy,; ;, ]p :, , , ;.1; L.. a q., >.,.m ,s ,u; ;,, t .- . t. c, - ;c; 7. , , . 6, FSiRREv5th5YiET 4
References:
Safety Evaluation: Pbc B(,-75 Rev No.: O nate: f/+N 7 Support a change List FSAR test, diagrams, and indices affected by this change and corresponding FSAE. revision. Affected F5AR Revision to affected FSAR Section is shown on: Section Preliminary Final
.5Ec7,oals 2.8.33 3.B.4,3.s,5,a3.8.6 Attachment 1 PsG. 3.t.I M E49, Rev. E.7 ,
Fs C . 3 . T . F. M - 1 F - 2.3 FIG . "3.%. 3 E J. %. G, Attachment i, I Attachment i Attachment l
?. >
PRELIMINARY FSAR REVISION (to be completed at time of Safety Evaluation ; preparation). 1 Prepared by:bD.Nem.A./Date: 4/(f87 Reviewed by: to: F/l7 y ; Aporoved by: M /Date: M/f'7 ! l FINAL FSAR REY 15 ION (Prepared following operational turnover of related ! systems structures of components for use at PNPS). (1) i I" Prepared by: /Date: Reviewed by: /Date: (1) Attach completed FSAR Change Request F.orm (Refer to NOP). Exhibit 3.07-A Rev. 2 Sheet 3 of 3 l lSSUED FOR CONSTRUCTION .
)
l
m , .. mu.. m w w m. w m..-c m . . m>, . . . . .r am . . ,., x e mm-- . v - . m. - _ , '.r,.m-~-------
'. Safety Evaluation No. Ai 3 r '., EL:, o - s/4 /s7 The technical specification changes will provide adequate operational and surveillance requirements for the SLCS modification and will not reduce the margin of safety.
This modification does not involve an unreviewed safety questiop. References
- 1. Gs.aeral Electric Company Letter, Pilgrim ATHS SLC System Modification, R. G. Ferguson to R. N. Swanson, dated 2/2/87.
5 155UdD FOR CO N STRL.G. . ... -,::. ... .:. .- .
---=.=. .
-L--._* _ _ _ - . - _ . - _ . . _ ,
w . c.:ww.:.: c,x =
. . ;w . ,wu .. ~ . ,. . .e c.w . . v a - >
. . . . . - c.w . .a s . w.. ..
Safety Evaluation
'. No.: $%6
. SAFETY EVALUATION WORK $HEET Rev. No. O
~
.A. System Structure Component Failure and Consequence Analyses. i System Structure Component Failure Modes Effects of Failure Somments
~
SLES SLCS 5.ld.en See, A Nac.hed 'Shech 1,
%e 4 3s*F ,
- 2. SLCS .Lale; lib to See 4Wekecd Sheed Atdown '
4Le Rese4e+- by SLC 3. l General Reference Material Review CALCULATIONS REGULATORY FSAR SECTION PNPS TECHNICAL SPECS, DESIGN SPECS PROCEDURES SUIDES STANDARDS CODES l.2 3. 4./ 4. 4- G E SPEC. 2 5 7M A 49 loan Co. G 2 3 2, 3 f 3,t
- 5. P.-
- p. t.
APPf2MG
- 8. For the proposed hardware change, identify the failure modes that are likely for the components consistent with FSAR assumptions. For each failure mode, show the consequences to the system, structures or related components. Especially show how the failure (s) affects the assigned safety basis (FSAR Text for each system) or plant safety functions FSAR ,
Chapter 14 and Appendix G). 1 Date 4 !d' 7 Prepared by b ' NOTE: It is a requirement to include this work sheet with the Safety Evaluation. Exhibit 3.07-C Aev. 2 1 > _- l l
'SSUED FOR CCM_STPd)CT$$h
- mr= w
- :uw :.n:a n . w < . :.. .x:., an . . ::. : >.. w . . . n. w. -
a~ . .:..e:o
'. Safety Evaluation-No. 2eal ga. o - systr7 SAFETY EVALUATION HORK SHEET A. System / Structure /Comoonent Failure and Consequence Analyses
- 1. System / Structure /Comoonent: Standby Liyuid Control System Failure mode: SLCS solution temperature less than 38'F.
Effects of Failure: Enriched sodium pentaborate solution crystallizes in the pump suction pipe rendering the system inoperative. Comments: In order to ensure a solution temperature greater than 38'F;
- 1. Solution temperature will be determineri daily.
1
- 2. Tank heater and heat tracing commence operation when the solution temperature reaches 53*F.
l - 3. Solution temptraicra of 48'F will alarm in control room, reactor must be placed into Cold shutdown condition within 24 hours of alarm. l
- 2. System / Structure /Comoonent: Standby Liquid Control System Failure Mode: Inability to shutdown the reactor.
Effects of Failure: Core damage, release of radioactive materials. Comments: Analysis performed by General Electric to assure the modified Standby Liquid Control System will provide the equivalent of 700 ppm of natural boron to maintain the original shutdown requirement. I
- .=:- _-
ISSUED FOR ) CONST1UCTiON j ,._,.7._.., l I I
w n : c.:..aw.mx.w , . : .c:a._ - ..c.c. ~ w :. ua,.. .
. v . .
a'> , -: m - ~e :r =- -- v--- ; s : PDC 86-75, Rev. 213 I Safety Evaluation,(Rev. O Attachment 1
- SheetIofIK ATIACB Etfr 1 RECOMMENDED FSAR CHANGES
_ __ as The pages of the following sections, ab es & figures of the FSAR that need to be updated due to SLCS modification P 86-75) have been marked with l cuggested updates and included in this attachment for your review. I FSAR Sections: 3.8.3, 3.8.4, 3.8.5, 3.8.6 One T= m: _'-1. 3.04 W FSAR Figures: 3.8-3, 3.8-6 The following drawings will be revised as part of the Plant Design Change package (PDC 86-75) but are not included herein. Dwg. I. D. FSAR FIGURE TITLE - M249 3.8-1 P&ID Str System M-1Y-213 3.8-2 sir System Process Diagram e ISSUED FOR CONSTRUCT!ON ___ _ .... .. J 6\
w :: w. a - :n:x . ..m .x ,aw .r.,w . -.u - - -
. ax =;
6
- PDC 86-75, Rev. O
~
Safety Evaluation, Rev. 0 Attachment 1 N PHPS-TSAR Sheet 2 ofIQ
- 9 /3 7
3.8 STANDBY L2' QUID CONTROL SYSTEM - . 3.8.1 safety objective The safety objective of the Standby Liquid Control System ($LCS) is . to provide a backup method, which is independent of the control rods, to maintain the reactor suberitical as the nuclear system cools in . the event that not enough of the control rods can be inser.te d to counteract the positive reactivity effects of a colder moderator. 3.8.2 safety Design Basis
- 1. Backup capability for reactivity control shall 'be provided, independent of normal reactivity control provisions in the nuclear reactor, to be able to shut down the reactor !f the normal control ever becomes inoperative. ,
. . .. 2. The backup system shall have the capacity for controlling the
, .:,7 . : '- re, activity difference between the steady state rated operating condition of the reactor with voids and the cold shut &own ' " condition, including shutdown margin, to assure complete shutdown t., from the most reactive condition, at any time in the core life,
- 3. The time required for actuation and effectiveness of the backup .
control shall be consistent with the nuclear reactivity rate of
- change predicted between rated operating and, cold shutdown a
conditions. A fast scram of the reactor or operational control . of fast reactivity transients is not specified to be accomplished by this system.
- 4. Means shall be provided by which the' functional performance capability of the backup control system components can be
- verified periodically under conditions approaching actual use requirements. A substitute solution, rather than the actual -
neutron absorber solution, may be injected into the reactor to test the operation of all components of the Redundant Control system.
- 5. The neutron absorber shall be dispersed within the reactor core in sufficient quantity to provide a reasonable margin for imperfect mixing or leakage.
- 6. The system shall be reliable to a degree consistent with its role
~
as a special safety systems the possibility of unintentional or shutdown of the .,. . . .. a - ie a . __ - .. 3.... De uriptio" CONSTRUCTION a p re-3 t h-The piping and instrumentation for .... .... .. ..~... ,u. Figure 3 8-2 is a process daagram for the system. The SLCS is manually initiated from the main control room to pump a boron neutron absorber solution into the reactor if the operator believes the reactor cannot be shut down or kept shut down with the control rods.
- _ _ _ ___=_-___
w a .. m .. i.w. a u x.: : . n= c x .x. s . w, ; ~ . m ,, i+~v v m.mm PDC 86-75. Rev. 0 Safety Evaluatipn, Rev. O -% Attachment I- b PNM-Tsu SheetJofif M/ l b4 ; However, insertion of control rods is expected to always assure prompt shutdown of the reactor should it be required. The boron absorbs thermal neutrons and thereby terminates the nuclear fission " 1 d I
/ chain reaction in the uranium fuel.
The SLCS is needed only in the improbable event that not enough
,l w control rods can be inserted in the reactor core to accomplishgf e
- shutdown- and cocidown in the normal at a steady manner. rateThe within SLCS the' therefore is $ 4 E
sized onlyofto shut the reactor downthe shutdown cooling systems, *and capacity - keep the reactor fro d
.* going critical again as it cools. .
The boron solution tank, the test water tank, the two positive 4 } 1 jI had associated local a7 j I d
- di8 placement and pumps, the two explosive valves, controls are mounted g,f in the Reacto j
' g Mr valves The liquid is piped into the reactor vessel and x y primary discharged containment.the near bottom of the core shroud so that it mines with See section 3.3, Reactor
) !
the cooling water rising through the core. Vessel Internals MechanicalDesign,andSecpon'4.2,teactorVessel
- Ql I <
and Appurtenances Mechanical Design. $ gue,M ,, sodium pentaborate The specified neutron absorber solution is a 'n: _ -u - : : A r,, 7 n:r-
^
ggsolution.4 2t is prepared byp ineralized water. An air sparger is, } _:: - rr W i; ;/ in d2m To ' prevent system plugging, the . provided in the tank for mixing. tank outlet is raised above the bottom /' of the ~ t A. - a gg ' a strainer. ' f 4a4Mok core At all times when it is possible to make the at reactor-least d p gal of k critical, the SLt5 shall be able tosolution deliver or equivalent into the - O.C2 63) pereent (sodium pentaborate_ - m.ww ,,ews a w g reactor, fynd""2s accomp2 , e liqui r ey control tank and lling w t stan h .oo ate in olume. he "I $ desin alized w design er to aoncentrathon low east le 1 alarm arm
-at th low level po an 5 sol ion is erflow vel vol . to all w_ J~ {16 m be di ed up vapgfati 1 _li_ss s, __f o the to 1_ow_' the aJt s tiert tamoer shwa 9 >'"' -
g' temperature of the specified solution isWF so the g I The u saturation h
/ equipment containing the solution is installed in .- ' a room" in which:^~TtheM h G air temperature isAntoelectric be controlled immersion L heater in the tank, and a d)
,/ J. 2
-_-*-3.
5:o,i:: ' !: 2:inni "l: :^ ^ ir d : :4" M 7 contro11ere : Mi temperature ---*3- : _ . . ,. . . . . . .n r r n it is aboo used to elevate the
/; temperature High or and assure that the boron dissolves when first added or to a the water. low temperature, high or low liquid level, s1'E'I shorted heater causes an alarm in the control roomf-- 4=e % aa tach positive displacement pump \is sited: T.: ^ to inject the solution solution level in into the the reactor in 50 to 225 min,( , ~
The pump and system design tank, at all reactor operating pressures.The two relief valves are set to exceed the prassure is 1,500 psig. ns m p , .. ; = 4 d 4ral. v. reactor operating pressure by a suffic th the 10 leakage. The relief valves are installed w 3.5., CONSTRUCTION m
.g;. : u . ; , . ; x x - .rax w . , ;. ; . ~, , .:.w ;, a:: , ...x,
. , . 7 x.: = . , - . ; - ., , < . .r; c. , - - - - - - - . . PDC 86-75, Rev. 0 Safety Evaluat on. Rev.0 Attachment 1 Sheet 4 of /q 4
GEllERAL ELECTRIC CD. c2 / 3/ ! Nuclear Energy Business Operations l ENGINEERING CALCUL4.10N SHEET l. DATE , NUMBEa __ sy SMEET or SUSJECT_ re 1. das q .. i ses *.9 h Lh. 6 4L 1 e,. 9%t r aa.~m a L.a s g. .( s.n ua.:r% ## p,% -eu .r- ,g$ p .au%a%* d.Luh 1, x wm. ww
-s a u:r , -
z Ab 9 % s. y gqs ramu.w k q 's ~ k to
.x -R e . -m 9 - ,& . . .
; gu w.,
nu = 4 ..- M a
. R(.A(u 4 *g&+ c uy .A
~ .
%g pym,A r t ,
pg a uumt e w u. n ~.w . 1 vp s .. . s - w r-( "Ro a.cy a ifff k a f*Cnetes YJ of Yke Z~ m ee 6a Srmoor ,ci cu, o rrs). co~m n e
. 1 a
ISSUED pog
+ CONSTRUCTION = 2 ::- - Ij
---__ ___E, .,__m__,
PDC 86-75, Rev. 0 Safety Evaluation, Rev. 0 Attachment 1 SheetSof /g Q .h y PNPS-FSAR To to prevent eva'poration and precipitation within the valve. prevent typass flow from one pump in case of relief valve failure 16 - ! the Ifne from the other pump, a check valve is installed downstream l cf ea:h relief valve 1tne in each pump discharge line. I The two explosive actuated injection valves provide high leakassurance into the of i opening when needed and ensure that boron will notThe valves have a f reactor even when the pumps are being tested. Each esplosive va.1ve firing reliability in escess of 99.99 percent.The plug is circumscribed is closed by a plug in the inlet chamber. eith a deep groove so the end will readily shear off when pushed by This action opens the inlet hole through the the plug.valve plunger.The sheared end is pushed out of the way in the chamber, and is shaped so it will not block the ports after release. The shearing, plunger is actuated by an emplostve charge with dual ignition primers, inserted in- the side chamber of the valve. 4 Ignition circuit continuity is monitored by a trickle current,Indicator and an
-- alarm occurs in the control room if either ci,rcuit opens.To service a valve, ,
Itghts show' which channel primer circuit opened. length pipe-(spool piece) must be removed ' af ter firing, a 6 inlamediately upstream of the valve to gain artess to the shear plug is actuated by a three pos'ltion keylock switch .on the . The SLCS This assures that switching from the "off"
., control room console. Switching to either side' starts one
- position injection pump, is a deliberate act. opens an explosive valve, and closes the Reactor g
- Cleanup System isolation valves to ' prevent loss or dilution of the h boron. h$ . .
A green light in the control room Indicates that j, power is a l running). A red light indicates that the contactor is closed (pump g 3 h' running). ie hF l Liquid flow is confirmed by a decrease in reactiv1'ty, storage E tanka red Itg drawdown and pump running Indication. j {K# switch turns on when valve 1101-1, downstream of the explos the Itquid may not be flowing, the operator can tamediatelyJ is open.
-r turn thethis sw
- keylock equipment.
switch to the other side:Crosspf The ptng and chosen check pump valves -fa will start assure a flow path jF J
- either pump and either explosive valve.even though its local switch at p ovided in the c trol r .
Equipment drains and tank overflows Lr tptp q q fyft(m {g l but to separate containers (such as 5!i 1 r; _Gv 7 -- J
.ny J.u vf70rortrom- #
st . . u .... " ' 4 . - 1,5to pir sm"1%su h s m vs. = =wed. M inadvertently reaching the reactor. ( A*m a*A .4 m h.em lan4 h ee esa - ) the standby liquid control : Instrumentation is provided locally attemperature indication and control. tank and consists of solution
,es
m=;- - n .o.;w , a . , ,.w . a.- .a . .- w .. .: : w , e . m ., . PDC 86-75, Rev. 0
,p Safety Evaluati n, Rev. O Attachment 1 PNPS-FSAR Sheet 6 of /4 2/3 f M .
1 Instrumentation and control logic is 4c tank level. and heater status. , presented on Figure 3.8 4 ' 3.8.4 Safety Evaluation
< The SLCS, although not necessary for plant operation is required to '
I be operable when the reactor is in other than cold condition. Jagf Despite this precaution, the system is espected never to be needed for plant safety because of the. large number of independent control - To further assure this " dg rods available to shut down the reactor.
'h availability, two sets of the components required to actuate the % w y pumps and explosive valves are provided in parallel redundancy.
- Mgl .
The system is designed to bring theThe reactor from rated power gtoNaka reactivity compensation g d cold shutdown at any time in core life. b- y reduce reactor power from rated to zero and allow g provided willcooling the Nuclear System to a cold shutdown It 5condition 8'
"' with the U,
.l g control rods remaining withdrawn in the rated power . pattern.
5f includes the reactivity gains due to complete decay of the menon v ~$ inventory. .!t also ' includes the ' positive reactivity effects fron % J' eliminating steam volds, changing water density from Phot 2-. to cold.t reduced Doppler effect in. uranium, reduction of neutron leakage fresh boiling to cold, and decreasing control rod worth as the moderator ( : 4*4 s :-+ The. specified minimum final concentration of boron in thei l dg s. reactor core provides' a reactivity worth of apprortsately -0.12i #p .e f a cools. ,.... A kb 4 f .05 0 A k for . calculational uncertainties an g-c < ' < $ g
- gplus a margin ofassures a substantial shutdown margin. ,,
I J U^ > 7w
' fee fs 38 ge concentrate,on of natural boren in theT i y
The specified nintaus avereactor, to p ovideThe the alnlaus 5 ecified shutdown quantity 5 -f margin af ter o
-4 the SLCS Is ppm, L . _ V . _ _ 5.
J ! ofesodium pentaborate t be injected into the reactor is calculated ,n i based on the required ppm average concentration in the reactor el 6
4$ coolant, and the quant y of reactor " coolant
-,--J in the='reactor
'--vest Y hI and recirculation loops " .
D result ts increased by 25 percent to allow for taperfect mining, ; leakage, and volume in other small piping connected to the reactor. g 2h " Cooldown of the Nuclear $ystem will take several hours as a sintaus, to remove the thermal energy stored in the reactor, cooling water. nj l-
)
vp , and associated equipment and to remove most of the radioactive decay heat. The controlled limit for the reactor vessel cooldown Usually,is IJ y " QD 100*F/hr and normal operating temperature is about 550*F. shutting down the plant with the main condenser and various shutdown LtJ g ' ;
. cooling systems will take 10 to 24 hr before the reactor vessel is 3p '
opened. and such longer to reach room temperature gg 00*F) wh requires the maximum boron concentration. The Snjection rate is limited to the range of 39 to 79 gal /stn. 'Qy l The lower rate assures that the boron gets into the reactor in about 1 1/2 hr. considerably , quieker than the cooldesn rate. The upper M ** *
*A* I * ,
(0 08
$* n.%,( a,b gmc.gh b whaS. beenn >.
. _ n _ .- ,. ~ ~ . ~ . - - . , ~ , .c , . . --- . , -
- 4
- PDC 86-75, Rev. 0
, Safety Evaluati n, Rev. 0 Attachment 1 "' . Sheet 7cfy ,z PNPS-FSAR Itoit injection rate assures that there is suffletent alsing so the. -
boro 9 does not recirculate through the core in uneven concentrations ' thicn could possibly cause the nuclear power to rise and fall '~ cycIlcally. - The system piping The SLCS ts designed as a Class I seismic system. and eculpment are designef, installed,Nonprocess and tested in accordance equipment such as with USAS B31.1.0 Section I and Appendt A. - the test tank is designed as Class II. The SLCS is reautred to be operable inThethe pumps event of a station po standby ac power supply in the absence of normal power. and valves are powered and controlled from separate buses and . circuits so that a single failure will not prevent.. system. operation. Itghts are powered from the essential instruments and The - - 120 V at instrument power supply. 1865
.The SLCS and pumps have sufficient pressure mergin. up to t e sysinjection h tem J
l into reitef valve setting of 1.400 pstg. to assure solutt pstg in the . the reactor above the normal pressure of aboutThe nuclear system relief and e safety v bottom of the reactor. begin to relieve pressure' abode about 1.100 pstg;'.therefor f' positive M*A** A deg,, e. a. s }7 The The shutdown maroin from this concentradon Isystem ]ts adenseR to ? the reactorAof 700 ppmpilgrim's Supplemental Reload License Submittal in can be foundThe in analysis and models for the reload core are Appendt Q. described in the GE Standard Application for Reactor Fuel."' 3.8.5 Inspection and Tes'ttag j operational testing .of 'the SLCS is performed in at least_ two parts to . j avoid inadvertently injecting boron into'the reactor.juy ope ing ' N~5 I tifi its Io 4 fs u o abeMir[ateNy 1 $ M eI N r e$ ch. n the va <ves to and from the solution tant closed and theopened J1~three s valves (two locked closed)in the test tank can be rectreulated 7 by 3 .$ the demineralized water gimpw poren i . ( 4 turning on eithe_r pump _ locally. g m tprmeVtemuM aut4h. m ia P inec ut- %.v,a e nettonal testing of the injectio6 portion of the ' system is accomp ished by closing the locked open valve from the solution tank.
'cpening the locked glosed room valvetofrom the test tank, and actuating the either the A or 8 circutt.
keylock switch in the control in that This circuit. starts the pump and blows open the injection valveThe light system 15 operating. ISSUED FOR CONSTRUCTION
- .. . < ,. ..., e
~ - ;;mwa= ..: a.: : . :. m ., w :w -
..w . .n x. , , ;w n .
w.:
. PDC 85-75, Rev. 0 Safety Evaluathn, Rev. O PNPS-FSAR Attachment 1 L 2G/
Sheet 6of14 in the sy closing aleakage local locked open valve to the reactor through the injection valves can be detected at containment. 3 a test connection in the line between the containment isolation c valves. valve is closed for tests, or open and ready for A the local Leakage from the reactor through the first check valve d cperation.) connection whenever the can be detected by opening the same test er reactor is pressurfred. ::$
)
After the functional tests, the Injection valves and explosive ge charges must be replaced and all valves returned to their normal * ~~ positions, as indicated on Figure 3.8-1. S
~
The test tank, contains deminerallred water for about three sin of *
- 0. mineralized water from the makeup or condensate pump operation. storage system is available at 30 gal / min for rafMt11tng or flushin the system. . .
4 .f Should the boron solution ever be injected into the reactor, either < t'b.,,-
.. intentionally or inadvertently. then, af ter making certain that g the y normal reactivity controls will keep the reactor subtritical, the+j .i boron' is removed from the Reactor Coolant lystem by flushingg< for f gross dilution followed by operation of the Reactor Cleanup r System.
There is practically no effect' on reactor operaf tons when the # concentration has been reduced below approntmately 50 ppe.. -
. g .
M The concentration of the sodium 'pentaborate iri the solutton' tafik is M
- determined by chemical analysis periodically. " N e.eit b ast .4 4Is.
The gas pressure in' the two accumulators is measured periodically to detect leakage. A pressure gage and portable nitrogen supply are required to test and recharge the accumulators. O 3.8.6 - C.e Q *w Ud. naae, emas Gen.r dedk Yu, us4. hNt usw bla. t s' C # t. 85b A t , h ^
%b w 4.a m it.c.A gw4 AgM A e.3d h.
< e .% we g 4. p.sgu % p W cwmue e) 1 . -- -
ISSUED FOR CONSTRUCTION t Revision 6 - July 1986 3.8-6
~ ~~----m._ _ _ _ _ _ _ _ _ _ _
rmm.mm: ::: s. m.:; = u . ,e . . . . u:.m- - < -
. : .. e .
w.. . . . PDC 86-75, Rev. 0 SafetyEvaluatJon,Rev.O Attachment 1 %
' PNPS-FSAR SheetT of /4 E/3/ j 3.8.7 Current operational Nuclear Safety Requirements operation, survet11ance
- current limiting condition for -
The re cuirements , and their bases are contained in the Technical Specifications referenced in Appendix 8. . 3.8.8 References General Electrical Standard Application for
- 1. WEDE-24011-P-A, .
Reactor Fuel, applicable revision. 4 Gs A l Cegoi,ty-P, Standh i ifdel Conbl Sp amGaeva E le.c.h-lc. ,l Bd.c) 87 ! /z9 E p valenc.e Rarovt 3 z
~
ISSUE 3 FOR CONSTRUCTION .-_-=.===.. g Revision 6 - July 1986 3.3-7
a m . a:.re m a aca s :.= w ..u w . . :.;: .w. : w, , ,. , ,. u . ,.n _ , , , . _
' '. PDC 86-75 Rev. 0 SafetyEvaluatjen,Rev.0 Attachment 1 Q Sheet /d of 14 2. I 3 l 4
i 1 I I I I I I )
. = ,
E . \ i= - 4 I I I I I I .I s 1
- a
'h
?# ISSUED CONSTRUCTION FOR
- ' FIGURE 3.8-5 -
- S ATUR ATION TEMPERATURE OF SODIUM PENTABORATE SOLUTION PILGRIM NUCLEAR POWER STATION FIN AL S AFETY AN ALY.S S REPORT L-- l--- Lll-_ _L_-.* ---.. _-_L. . . .L ?_.
- -. - . . . _ - . . . . - - - _ _ . . . - - . - 1
=.= ..v.za. .w c..~ w -:.c a ., .1. . ., w: . . w. w.,w - -
...~ , a. =a>., . wa~- ,..v .
4, i
' PDC 86-75 Rev. 0 Safety Evaluat on, Rev. O 4
. Attachment 1 Sheet Ifof Ig ale { \
i I I I I . is .
~
n - . esseesiwe-tem sene s** ism-t m
^ g 34 at Il V06tast CanCEnfaatim Ie -
33 g
. i.= - /
- V t -
n - L J, _ .
, *350 gessons.gmi ,
- = _ d o
**ue sa'me emi l, Loe-LavsL acaer i
. r .- I i n.n. .. ' I -i IIID-LEVEL ALAM ' 1 l l ' 1 l l
~
TasIE OVERFLO. t i i 1 I I
= - -
, =
stT WOLE;; - _-- - 1 = Tweh . ISSUE] FOR
, CONSTRUCTION __=== ji.
s s. ssT r
" stege5- / FIGURE 3.8 -6 O SODIUM PENTABORATE SOLUTION VOLUME CONCENTRATION REQUIREMENTS PILGRIM NUCLEAR POWER STATION FINAL SAFETY ANALYSIS REPORT
v.n:ammm wx. w . - v.u w. .. a . . -.,e:n.; . -
. x w e ,- ..e _
s PDC 86-75 Rev. O
' GENERAL ELECTRIC CO. SafetyEvaluaQon'Rev'0 Nuclear Energy Business Operations Attachment 1 Sheetjgofl4 L 2 13/
ENGINEERING CALCULATION SHEET L.A ons sle1lM-m agn se:.' t.M - emo%6l1 , sy_DERE _gnggr 3 h op W- M %L 4 Cad . SU5 JECT _ 7 %L e e.s k l% 9.gt . 8 4 r fe,1y9 s t v. w< .. 3,**3e,e
~$
1 c' s f a$ - e e
)
y s is 3
.__ i.
.s . c g 25 N Y h *
.l h ve
- h a 3 '
g.. .
.r. . , . !:
* ,5 l 1 \{,g, o 1
. .o ,
s : 1> se 3h,p s 7 \ !. g 2 2 p 2 .l,, g;
- ss
/ f/ q E
s g-p i g .- l , ' pf2v _- J- _ } .!= c .
. k. g i ,j
~
!ijt. (SSOED FC%, !'
Fj31a E h e 4 4, CONSTRfhJS O 4 m
" e r o k
h t*
- Aa. =%m.%en m,%3mg ymm ,
i = .
< cwzzca mw. na u . . . . w. .. w= > a, m . u. = n r. ,, a . u. , + . ..
t. PDC 86-75, Rev. 0 Safety Evaluat on, Rev. 0
* ; ' Attachment 2. .
Sheetiof3 1.13/ 1 GENERAL h ELECTRIC STANDBY LIOUID CONTROL SYSTEM ., i 4 CONTROL CAPACITY l - \
~ - EQUIVALENCY REPORT a
PREPARED FOR THE BOSTON EDISON COMPANY PILGRIM NUCLEAR POWER STATION JANUARY 29, 1987 ( ISSUED FOR~ PaE m tD BY: R.T. zxatt E~C Ef.oAlr. 1/21M CONSTRUCTlVi '
,!r VT.RIFIED BY: J.K. SAWABE
) 'A.Q 1-24 ht?
l Verification Material in DRF C41-00095/2, Section L4. 6 J______. _ _ . ,
\
- w. :mm m . , . .,w, .
.s w, :=
, : w:s -
u: :.<. ruv co-io, nev. v
, . Safety Evaluati , Rev. O J Attachment Q,, , Sheet 2,of c) Alg j DISCLAIMER OF RESPONSIBILITY This document was prepared by or for the General Electric Company.
Naither the General Electric Company nor any of the contributors to this document: 4 f A. Makes any warranty or representation, express or implied, with j respect to the accuracy, completeness, or usefulness of the information contained in this document, or that the use of any - information disclosed in this document may not infringe privately owned rights; or, B. Assumes any responsibility for liability or damage of any kind which may result from the use of any information disclosed in this document. .
.e.*
~
~
'SSUED FOR i CONSTPsuCTIOiQ
- n. = = = - ,
i
p ==. m ;. m ::c.a=. . . se . . ..p.. .
... . m. v . a.. .
.a. =.. .,w.-- ..
PDC 86-75, Rev. 0
,' Safety Evaluation
{, Rev. 0 Attachment '2. Sheet 3 of 9 2.lq; ABSTRACT , This document was prepared for the Boston Edison Company to address tha requirements of the standby Liquid Control System (SLCS) at the .s Pilgrim Nuclear Power Station for compliance with the NRC ATWS Rule 10CTR50.62. The plant specific values used to demonstrate compliance with the NRC ATWS Rule are the same as the minimum values provided in tha system Technical specifications. e r l ISSUED-!FOR CONSTRUCTION _ .-_ -. . . .o g
e- n . au., .-a . . ~. ..~ ., .... ~ . .- ,, .c . . . . . , . . . . . . . . . . = - - , . - , =
- PDC 86-75, Rev. O Safety Evaluation Rev. 0 Attachment 1 l Sheetfof9
{2.1gj
. l TABLE OF CONTENTS 11 Abstract 1
- 1. Introduction ,
1
- 2. Discussion .
~
1
- 2.1 SL,C System Design Basis _
L 1 . i 2.2 NRC ATWS Rule _ 1
- 3. Analysis .
l 2 3.1 Equivalent Control Capacity Definition 3 3.2 Equivalent control capaci,ty. Calculation 4
- 4. Summary 5
- 5. References
} .
. _ -m l [ .15 SUE 0 FOR CONSTRUCTIOll)g -.- =... .
- iii -
.r .
- n .m.: m - m r.:c:a x "...w ..u . w .- n m .m ., .. .w
. c w.. e
- PDC 86- . ..v.
0 Safety Evaluatiop, Rev. 0
- Attachment 2., j
. (_,.213 1 SheetSofp
1.0 INTRODUCTION
Coston Edison has requested an evaluation of the minimum required concentration (weight percent) of sodium pentaborate for the Pilgrim Stendby Liquid Control System to comply with the NRC ATWS rule . requirements in 10CTR50.62 (Reference 3.). The minimum concentration is to be based on equivalency to the minimum 86 gpm, 13-weight percent sodium pentaborate control capacity requirement stated in the NRC ATWS rule. Equivalency is calculated using the ratio of the specific Pilgrim minimum values to reference plant values that the rule is based on. For Pilgrim, a minimum solution concentration of 8.42 percent is rcquired. This is based on the assumptions that the minimum - bgen enrichment of the sodium pentaborate decahydrate exceeds 54.5 B atom percent, one pump is required to operate and the actual c pacity of each pump exceeds the required minimum pump flow rate. 4 2.0 DISCUSSION 2.1 SLC System Design Basis Tha generic de:ign basis for the SLC System is to provide a specified cold boron shutdown concentration to the reactor vessel as described in NEDE-24222 (Reference 4.). The SLC System was typically designed ,, to provide the specified cold shutdown concentration in about one or two hours. During reload licensing evaluations, this shutdown , concentration is verified by analysis to be adequate to render the core suberitical. The considerations'in the reload evaluation are independent of ATWS and injection rate is not directly considered. The ATWS rule requires the addition of a new design requirement to the gsneric SLC System design basis. Changes to flow rate, solution concentration or boron enrichment, to meet the ATWS Rule, must not invalidate the original system design basis. . ISSUED FOR CONSTRUCTIONjj . = . - c
m;an.m r.2 ,_,-_ .y , , , , , , , s w.w m - 2 ;- PDC 86-75, Rev. 0
. Safety Evaluatiqn, Rey, o
- Attachunt 1 Sheet (, or cy [29 .
l 2.2 NRC ATWS Rule . Parcgraph (c)(4) of 10CFR50.62 states, in part:
"Each boiling water reactor must have a Standby Liquid Control ,
System (SLCS) with a minimum flow capacity and boron content equivalent in control capacity to 86 gallons per minute of 13-weight percent sodium cantaborate solution." Tho NRC Staff has provided clarification of equivalent control capacity (Reference 5.) as follows: (1) The "equivalen't in control capacity" wording was choosen to allow , ficxibility in the implementation of the requirement. For example, the
~
equivalence can be' obtained by increasing flow rate, boron . concentration or boron enrichment. ,
'(2T The 86 gallons per minute and 13-weight percent sodium pentaborate ware values used in NEDE-24222, " assessment of BWR Mitigation of ATWS, Volumes I and II", December 1979, for BWR/4, BWR/5 and BWR/6 plants .
with a 251-inch vessel inside diameter. The fact that different values would be equivalent for smaller plants was recognized in NEDE-24222.
"The flow rates given here are normalized from a 251-inch diameter vessel plant to a 218-inch diameter vessel plant, i.e.,
I the 66 gpm control liquid injection rate in a 218 is equival'ent to 86 gpm in a 251. This is done to bound the analysis....(pp. 2-15 [NEDE-24222))." ' (3) The important parameters to .cfon' sider in establishing equivalence , cre vessel b'oren concentration required to achieve shutdown and the , time required to achieve that vessel boron concentration. The minimally acceptable system should show an equivalence in the ) parameters to the 251-inch diameter vessel studied in NEDE-24222. 3.0 ANALYSIS
. j 3.1 Equivalent Control Capacity The NRC equivalent, control capacity concept of the ATWS rule is a very cimple, direct criterion that does not require consideration l of the
. mixing efficiency or to account for plant-specific core nuc ea ISSUED FOR '
CONSTPsUCTION i. ..
. ~- ~; t J-__-_--.- . _ _ _ _ .
w.a m n. w . w :a..auus w m c:,e .. m . w w w~ . , ri; ma rv m r .a - . . . . . 6-PDC 86-75 Rev. 0
- Safety Evaluation Attachment 1 Q , Rev 0
. Sheet}ofc) 2 y; ths equivalency requirement if the following relationship is phewn to be true:
,9
- E251
- C.,* E >= 1 (Equation 1)-
86 M 13 19.8 . i l whsre the plant-specific parameters are defined as: . Q = minimum SLCS flow rate (one or two pump operation as appropriate), gpm. M = mass of water in the reactor vessel and recirculation system at the hot rated conditions, lbs. . C = minimum sodium pentaborate solution concentration, weight percent. . s E = miT11 mum expected B isotope enrichment (19.8% for natural boron), atom percent. . .. ..
-. . . 2
., , ~
Th2 value of M (the. mass of water it) the reactor vessel and rceirculationhktematratedconditior.sintherefer'enceplantis 628,300 lbs for a BWR/3/4. This value was calculated based on rated temperature, rated void content, normal water level, control rods fully withdrawn, expected minimum vessel dimensions and nominal vessel , internals dimensions. 3.2 Equivalent Control Capacity C31culation The NRC requires the use of minimum pl. ant-specific values'to .
- demonstrate compliance with the equivalency requirement. For single pump operation, 54.5 atom percent boron enrichment, Pilgrim can demonstrate compliance if the following relationship is true:
- 8,1
- 19.8 (Equation 2.)
C >= 13
- g ,
~
- M 251 9 where C is this case is the minimum allowed concentration (weight percent) of the sodlum pentaborate solution, Q is the minimum allowed indivirhal pump flow rate, M is the mass of water in the reactor and recirculation lines, and E is the minimum allowed boron" enrichment ISSUED FOR CONSTRUCTION 6
-= _-_::-_______,__,__.__ ,
m.v.m.m ..m... ...m, m. .m.m., .....m. , . , . , -
, l PDC 86-75 Rev. 0 '
, Safety Evaluation, Rev. 0 Attachment 1
. Sheetgofg {2 9, Icval. The water mass is based on the same conditions as the geference picnt water mass. The minimum allowed individual pump flow r will become part of the system Technical Specification and Design Sp2cification (Reference 1). .
Q = 39 (minimum rated) gpm M = 507,850 lbs E = 54.5 % ' . Using the current Pilgrim plant-specific values (in Equation 2) gives o required minimum conc /otration .of 8.42 weight percent sodium pentaborate. (Equation 3.) C >= 13
- 507,850
- 86
- 19.8 628,300 39 54.5 ,
~
C.)= 8.42 ' 4.0
SUMMARY
Whsntheconcentrationofenrichedsodiumpentagrs.tedecahydrate ) is equal to, or (onrichment exceeding 54.5 atom percent boren B greater than 8.42 percent, Pilgrim meets or exc6eds the NRC ATWS rule-equivalency requirements. ,
~
ISSUED FOR CONSTRUCTION -_- ~ .= t
~
- 1 e
e e b_'______________m__ _ _ _ _ _ _ _ _ _ _ _ _ l
,ar::: 7;w.m. , a : ;c;;,, c:w ' . .. ., w . .: .u;.q,o
,:=- T :,:,. w ;;;:
;:':;:z-~ u. m e
- PDC 86-75, Rev. 0 Safety Evaluati n, Rev. O Attachment 9.
Sheet of 9 1;3j
,p
5.0 REFERENCES
1
'j
- 1. Doc. No. 257HA169, Rev. O, Standby Liquid Control System Design Specification. 1 I
- 2. Doc. No. 257HA169AV, Rev. 2, Standby Liquid Control System Design Specification Data Sheet.
- 3. 10CFR50.62, NRC ATWS Rule, June 1984.
- 4. NEDE-24222, Assessment of BWR Mitigation of ATWS, December 1979.
- 5. USNRC Generic Letter 85-03,Control Clarification Capacity of Equivalent for Standby
" '"' ... Liquid Control systems, January
- 28, 1985.
4 I; t %~ ~
.- , ISSUED c CONSTR(!CT ~ ON
?
)
i 1 m -._____1___ _____:__
m.irmcm ,w. mu .w .. w . . w.... w . . . . u n.. . - . s. ,.n.... --
..u.~ -
a 4-PDC 86-75, Rev. O Safety Evaluation No..l.131 Rev. O Attachment 3 Sheet I of y . ATTACHMENT 3 Recommended Technical Specification Changes ' The pages of the following sections and figures of the Technical Specifications that need to be updated due to SLCS modification (PDC 86-75) have been marked with suggested updates and included in this attachment for your review. Technical Specifications Sections 3.4 & 4.4 Technical Specifications Figures 3.4.1 & 3.4.2 _m fE.ee I ISSUED FOR ., CONSTRU -3;= cts ,t%.;.._ 7.'lu .:
- -----h.-..___. _.. . _ _
- X,74.!mma,a v. ;p : ;w;;.;.n.:;m;;,, g. , , ,u 9,, p- , , , , ,
. . .. ..,.a...
.s
* % Evaludici 4.20s tu.0 l 4HnWent 1
-,* SURVEILLANCE REQUIREMENTS W e+ 3 e# 7 LIMITING CONDITIONS FOR OPERATION 3.4 STANDBY LIOUID CONTROL SYSTEM 4.4 STANDBY LIOUID CONTROL SYSTEM I
- c. Manually initiate one of the Standby Liquid Control System loops and pump demineralized i water into the reactor vessel.
This test checks explosion of the i charge associated with 'l the tested loop, l proper operation of l the valves, and pump l operability. The replacement charges to be installed will be selected from the same manufactured batch as the tested charge.
- d. Both systems, including both explosive valves, shall be tested in the course of two operating cycles.
B. Ooeration with Inonerable B. Surveillance with Inocerable comoonents: Components:
- 1. From and after the date 1. When a component is found that a redundant to be inoperable, its component is made or redundant component shall found to be inoperable, be demonstrated to be Specification 3.4.A.1 operable immediately and shall be considered daily thereafter "Jntil the fulfilled and continued inoperable component is .
operation permitted repaired. provided that the component is returned to an operable condition l within seven days. l ISSUED FOR l CONSTRUCTION u
. 1
n uwamm: ..ww xw, s.wwu w., .a w ..c w m ,,- , e , ,.m mv - :. w.r u --- 4 6,, 5daty Enlofes% d.b'85i b o A l%. bane art 3 e,' LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS W e4 4 cT v
~
3.4 STANDBY LIOUID CONTROL SYSTEM 4.4 STANDBY LIOUID CONTROL SYSTEM C. Sodium Pentaborate Solution C. Spdium Pentaborate Solution At all times when the Standby The foliowing tests shall be Liquid Control System is performed to verify the required to be operable the availability of the Liquid following conditions shall be Control Solution: met:
- 1. Volume: Check at least
- 1. The net volume - once per day.
concentration of the Liquid Control Solution 2. Temperature: Check at in the liquid control least once per day, tank shall be maintained as required in Figure 3. Concentration: Check at 3.4.1. least once per month. Also check concentration
- 2. The temperature of the anytime water or boron is -
liquid control solution added to the solution, or shall be maintained above the solution temperature 48'F. is at or below 48'F. l 3. The enrichment of the 4. Enrichment: Check l liquid control solution Boron-10 enrichment level shall bg maintained at a by test anytime boron is boron B'O isotope added to the solution and enrichment exceeding 54.5 prior to restarting from atos percent. each refueling outage. Enrichment analyses shall D. If specification 3.4.A, B, or be received within 30 days C.1 or C.2 cannot be met, the of test performance. M en reactor shall be placed in a 5 :t ; 2 --- -:: t " , Cold Shutdown Condition with F a check sha 1 es mace to a all operable control rods ensure that Boron levels fully inserted within 24 meet the original design hours. If the enrichment criteria by comparing the requirements of specification # enrichment, concentration 3.4.C.3arenotmet[bnhag. and volume to established h "" criteria. G ne ScWH g.+bf b [theBoron-10Isotonicenrichment top 4.5 ' Atom levels ^clo 3 hol' WCET~ percent within seven days +h9n I g g4 l g ,p g.ger; , vels heci from theline of enrichment 1 recess.fsubmitareportto / the P8%Clo" Sb4ll be Placed in 9 >i
'F l the KRC and advise them of Cold 6hqialown fonJ, hen ~.th P " . "h / plans to bring the solution gli CPevdle (chf4l Ndi bll /
l85d) viter ic's up to a demonstratable 54.5 QseWed wemm 24 6o.4%.
, v'n 3 i
( atom percent Boron-10 Isotopic Enrichment. M af4er g g tb 5yTitn y i fy3e's o d ( J,6 in yeven ch7s
~
d S M ' M ( O n7 r c. s f Il ho t. M N \ 3,4, C ,3 1 ISSUED FOR 97 CONSTPsUCTION -- _ a k
(t :L .' f
- ' v :; 4; } '/ 5t l i+. . f'J' - - jy i 1 .
- v .o s
#5O
~? %1E j 3: . u 5 R 5 O s
.' 9Iai U 3q b %v _
) _
\
M
=& u G T t Ec T= .
o
** ~o s w.
t w w sef se
% g
%t 1
a* w% m. g m. s E 4 6 V , m % w - .
)
4t 6 e mva
.M II H.t
. M, '
- n o
- "
- i .
-
- t s 7 N n
- o. o Lm n.
mI
' (
=
ut l n oe S m e er
- w. el t i uN d 'r q a u n yM . o e
@. o oM . S. . ba Rn m e. s at no b u ,wGuW m -
" v Pt ei u
o s
. mu 3 mr ut a
n
' T i e
h c
* #EU.t s
%g
\
g
%d oc S n 1 C o
uNe.r~ oT5 .m \
\ -
- 4. dn wvec_
w9
.)
\
\'
\
' wl .
m 2 w 3 a ee rm uu gl
.i o y FV
} .
nE s o o i T* ti tN r e e oOE *- Q ngW - otMt E" B@ 4 NwNgM_ eu e o. C* z t CO "09u. - em ( 9' s. e e. e OOfWM \ i o (- - - 0 q . 1 g5QW3
, 23Pd 9o EaWhefrI h 4
w.s.x w . n m - ... u ~:ir. . . . v . m. .v. m . n.; r , ; o .a c .. :.x <=.-- a x= t
- Pbc e4-75' Rev. O !
/
BASES: Adcfy I b4 %fsch Me. 2131 Eev. O !
. AMa chNthf 3 3.4 & 4.4 STANDBY LIOUID CONTROL SYSTEM gg gg7 A. The requirements for SLC capability to shut down the reactor are identified via the station Nuclear Safety Operational Analysis (Appendix G to the FSAR, Special' event 45). If no 1- more than one operable control rod is withdrawn, the basic shutdown reactivity requirement for the core is satisfied and the Standby Liquid Control system is not required. Thus, the basic reactivity requirement for the core is the primary determinant of when the standby liquid control system is required. The design objective of the standby liquid control system is to provide the capability of bringing the reactor from full power to a cold, xenon-free shutdown condition l assuming that none of the withdrawn control rods can be ]
inserted. To meet this objective, the standby liquid control system is designed to inject a quantity of boron that produces a concentration equivalent to 700 ppe of natural boron in the J reactor core. The 700 ppm equivalent concentration in the reactor core is required to bring the reactor from full power to a three percent A k subcritical condition, considering the hot to cold reactivity difference, xenon poisoning etc. The system will inject this boron solution in less than 125 minutes. The maximum time requirement for inserting the boron J solution was selected to override the rate of reactivity insertion caused by cooldown of the reactor following the xenon poison peak. The standby liquid control system is also required to meet. 10CFR50.62 (Requirements for Reduction of Risk from Anticipated Transients Without Scram (ATHS) Events for Light-Mater-Cooled a
' Nuclear Power Plants). The Standby Liquid Control system must ~ (*J "Q have the equivalent control capacity /Dr 56 gpa at u percent wt. natural sodium pentaborate in order to satisfy 10CFR50.62 requirements. This equivalency requirement is fulfilled by a M
combination of concentration, B-10 enrichment and flow rate of - sodium pentaborate solution. A minimum R.42% concentration and - 54.5% enrichment of Boron-10 isotope atJ39 GPM pump flow rate N q satisfies the ATHS Rule (10CFR 50.62) equivalency requirement.
. Because the z.adsed concentration / volume curve has been Yevisad6to reflect the increased 8-10 isotopic enrichment.1m-
;;d'.".'._-- ... n.. .::- Q , : : ..... v. 6uw o- 0 Z ;-- ^ 7. 7 7 0 ^7 C . a . Inla twusaavn . . . . - " A M N F-- ! Q
. J . . .. .
e r
--
- e:d ;;r.;;7.tTCt'-- 'v& s --
' Experience with pump operability indicates that the monthly f test, in combination with the tests during each operating cycle, is sufficient to maintain pump performance. The only practical time to fully test the liquid control system is during a refueling outage. Various components of the system are individually tested periodically, thus making more frequent testing of the entire system unnecessary. mm ne AtS6UEOisFORw Javl wana,a u enIde.c.g.u w a . mmc ,,s d enever de aao e, + eet l CONSTRUCT!Oi@" -
= ry ~ <<*a m sems Le,naaa,
y ::::.KEL .5 . a u. . _.w.. vg 5gpa , ;_ ,y - ; y w,3-h ' ' .. ' YbC %~ 75 Rev o M&y Evoludbm No. 2l3) Rev, O t BASES: A gg,g 3 3.4 & 4.4 STANDBY LIOUID CONTROL SYSTEM (Cont'd) S b88'f 7 CP 7 The minimum limitation on the relief valve setting is intended to prevent the loss of sodium pentaborate solution via the lifting of a relief valve at too low a pressure. The upper limit on the relief valve settings provides system protection from overpressure. B. Only one of the two standby liquid control pumping loops is needed for operating the system. One inoperable pumping circuit does not immediately threaten the shutdown capability, and reactor operation can continue Shile the circuit is being repaired. Assurance that the remaining system will. perform its intended function and that the long term average availability of the system is not reduced is obtained for a one out of two ; system by an allowable equipment out of service time of one third of the normal surveillance frequency. This method determines an equipment out of service time of ten days. Additional conservation is introduced by reducing the allowable out of service time to seven days, and by increased testing of the operable redundant component. C. The quantity of Boron-10 stored in the Standby Liquid Control System Storage Tank is sufficient to bring the concentration of Boron-10 in the reactor to the Point where the reactor will be , shutdown and to provide a minimum 25 percent margin ',eyond the amount needed to shutdown the reactor to allow for possible ' imperfect mixing of the chemical solution in the reactor water. ! Level indication and alarm indicate whether the solution volume has changed, which might indicate a possible solution concentration change. The test interval has been established in consideration of these factors. Temperature and liquid level alarms for the system are annunciated in the control room. The solution shall be kept at least 10*F above the maximum saturation temperature to guard against boron precipitation. Minimum solution seem-temperature is 487 Each parameter (concentration, pump flow rate, and enrichment) is tested at an interval consistent with the potential for that parameter to vary and also to assure proper equipment performance. Enrichment testing is required 4m4y when chemical addition occurs since change cannot occur by any process other than the addition of new chemicals to the Standby Liquid Control Solution Tank. e og ISSUED FOR CONSTPsLICTION =~%T c.--- u -.-____-._-_m________
vm . -; , n. ~ - -. - .. ., , - . . - - - . 1
. NED Prop > sed Change Safety Evaluation j
[* Mo.: 2lQ l SAFETY EVALUATION PILCRIM NUCLEAR POMER STATION Rev. No. O t PDC PCN System Calc. Initiator: Dent: Group: No.: Name: No.: Date: W. Babcock NED I&CS 86-104A RCIC 5/15/87 ) 1 Description of Proposed change, test or experiment: RCIC Exhaust. Pressure l Trin Setonint Chango ? SAFETY EVALUATION CONCLUSIONS: The proposed change, test or experiment:
- 1. (X) Does Not ( ) Does increase the probability of occurrence or consequences of an accident or malfunction of equipment important to
- safety previously evaluated in the FSAR.
- 2. (x) Does Not ( ) Does create the possibility for accident or malfunction I of a different type than any evaluated previously in the FSAR.
~
- 3. (X) Does Not ( ) Does reduce the margin of safety as defined in the basis for any technical specification.
BASIS FOR SAFETY EVAtt1ATION CONCLUSIONS: See Attachment #1. t Change Change ( 4 Recommended ( ) Not Recommended j SE Performed by h Date / 87 Exhibit 3. 7-A ' I Sheet 1 of 3 I { l 3.07-13 Rev. 4 i j FOR INFORMATION ONLY
m m . x .= m u..z.:.;;:.: m mecu,u . . . w e r .n . .. . : a
. w .. .:c. . c.:. . . .. ; - w .. . ,
6 Safety Evaluation No.: J l Get
,, c SAFETY EVALUATION , PILGRIM NUCLEAR PO4ER STATION Rev. No. O A. APPROVAL This proposed change involve a change in the Technical Specifications This proposed change, test or experiment does ( ) does not (X) involve an unreviewed safety question as defined in 10CFR, Part 50.59(a)(2).
g This proposed change involves a change to the FSAR per 10CFR 50.71(a) and is reportable under 10CFR50.59(b). Comments: See attachment #1. paragraph "H". l The safety evaluation basis and conclusion is: ( 4 Approv () Not Approved YN lG Discipline Group Leader /Date 61AI 6 0t Sybporting Discipline Group Ledder/Date B. REVIEW APPROVAL Comments:
!H L 6 '
O IjS11A Gro# Leader /Dath C. ORC REVIEW . This proposed change involves an unreviewed safety question and a request for authorization of this change must be flied with the Directorate of Licensing NRC prior to implementation. This proposed change does not involve an unreviewed safety question. ORC Chairman Date ORC Meeting Number Cc: Exhibit 3.07-A Sheet 2 of 3 3.07-14 Rev. 4 FOR INFORMATION ONLY e w __=__ _:-__-- _
. w . , , > :., _ ~ :.::,.s. .. :n. 2; . .. . .w :w . . : n:- ,- rz ;a. , . - ; n ~ . . :,, , , : . . . ; .;. . . - ,,; x .. . - u .. l s PILGRIM STATION
- FSAR REVIEN SHEET
References:
Safety Evaluation: 214 / Rev. No.: O Date: m Support a change List FSAR test, diagrams, and indices affected by this change and corresponding FSAR revision. . Affected FSAR Revision to affected FSAR Section is shown on: Section Preliminary Final Attachment 1 Table 4.7-2 Attachment 2 X Attachment 3 - Attachment 4
. Attachment 5 Attachment 6 PRELIMINARY FSAR REVISION (to be completed at time of Safety Evaluation preparation).
Prepared by: Al O A * *'*1Date: dif/f* Reviewed by: d b? /Date: / f +- Approved by: b M Date: N U U FINAL FSAR REVISION (Prepared following operational turnover of related systems structures of components for use at PNPS). Prepared by: /Date: Reviewed by: /Date: Attach completed FSAR Change Request Form (Refer to NOP). Exhibit 3.07-A Sheet 3 of 3 3.07-15 Rev. 4 FOR INFORMATION ONLY l
7.rg5y3 ;m; yq p;3, ;;;,, . ;, . n, , w g g,; ,,, a ; s a g; w ~.; , y ,,,a,;,,,., . ; g., ., g ; y - , , ,, Attachment #1 to Safety Evaluation No. 2/& /
,RCIC Exhaust Pressure Trio Setooint Chance" A. Description of Chance This modification raises the setpoint of the RCIC turbine exhaust pressure trip from 25 psig to 46 psig. No new equipment is r.equired for this change.
B. Purcose of Chanae
References:
(1) GE report: "PNPS Station Blackout Study (draft)", dated 3/10/87. This modification, when implemented, will allow RCIC to operate during an event which is beyond its original design basis. The changes described will enhance RCIC response to the " station blackout" (SBO) event defined in Reference 1. This event was not included in the original design basis of PNPS. RCIC is one of two systems available for injecting water into the reactor vessel during the - , SB0 event because it is not dependent on AC power to function. Throughout the SB0 event all AC power is presumed lost. Torus pressure will continue to rise during the SB0 event. The RCIC turbine exhaust steam is directed into the suppression pool. Any increase in torus pressure, therefore, directly affects turbine backpressure. If RCIC is
- to continue to run during the postulated event, its turbine must operate against continually increasing exhaust backpressure. Up to 50psig backpressure there is no effect on turbine performance.
Pressure switches (PS-1360-26A, B) are located in the turbine exhaust piping to detect high exhaust pressure, indicative of restricted or blocked exhaust piping. These switches act to trip the turbine when their setpoint is reached. Present setpoint of 25 psig was selected to be as low as possible to detect blockage of the exhaust line without causing spurious trips on turbine starts and is consistent with the maximum discharge capability of the gland seal condenser vacuum pump. In an operating situation, the RCIC exhaust pressure is less than 10 psi greater than torus pressure (Ref. 1). Exhaust line and turbine casing are protected against overpressure by a rupture disk set at 150 psi. Assuming turbine exhaust pressure trip setpoint remains set at 25 psig, the trip point will be reached at about 8.5 hours into the SB0 event. Raising the trip point to 46 psig allows RCIC operation until about 15 1/2 hours into the event. Refer to Reference 1 for details of the SB0 event and associated analyses. Page 1 of 5 FOR INFORM ATION ONLY
. . .a . .n w n . :.a- a. w.: .n..,.- . , . ~ ~ . ~ . m u. . . , uw.o w + .n . m,.- - m. n---
Attachment #1, S.E.# &lG/
]
C. Systems and Comoonents Affected
- 1. The only system affected by this change is RCIC. ;
- 2. Components directly affected are pressure switches PS-1360-26A&B.
These will be re-calibrated to actuate at 46 psig. This value is well within their adjustable range of 5-150 psig. The instrument (ffects of increasing the pressure switch setpoint has been evaluated in Bechtel calculation no. E-676-1 and it is concluded that the new setpoint falls below the G.E. analysis ! limit (and the process safety limit) for exhaust pressure even j when instrument inaccuracies are taken into account.
- 3. Comoonents Indirect 1v Affected 3.1 The high RCIC exhaust line pressure trip will be set at the increased pressure for all events which require RCIC response. As discussed below, no adverse effects result from RCIC operation with a 46 psig backpressure either during non-accident conditions '
or during accident or SBO. 3.2 The mechanical and operational limitations of the RCIC equipment have been addressed as follows: 3.2.1 Turbine casing The design pressure of the turbine casing is 165 psig per Attachment 3. This is more than adequate for a back pressure of 46 psig. 3.2.2 Exhaust line piping design Piping specification M-300 for the turbine exhaust line gives a design pressure of 100 psig and a design temperature of 325 degrees F; therefore, there is no concern with the exhaust line piping design if the turbine exhaust pressure is changed from 25 to 46 psig. These design values are used in allowable stress analysis and nozzle loads. Page 2 of 5 . FOR INFORMATION gONLY
- m u.c ,
mm .z .. 1;O .h ;? < m - a- ..w w ~ -
. . i c.:. . , , - ",?" n- c -
Attachment #1, S.E. # 4101 3.2.3 Gland seal condenser sub-system The gland seal condenser vacuum pump will not be functional if the turbine exhaust pressure is near 50 psig, as the vacuum pump maximum discharge prcssure is 25 psia. The total gland steam leakage of approx. 220 lb/hr for 50 psig (reference attechment 8, page 6 of 8) will still be condensed in the barometric condenser (part of the gland seal condenser sub-system) provided the condensate storage tank is used as a suction source for RCIC (ref: Attachment 5). The barometric condenser, if cooled by CST water, is capable of condensing the entire amount of gland steam leakage; however, for conservatism when calculating room temperatures. -4 70 lb/hr (Reference 1, based on eroected gland steam leakage with the seals degraded to about 10% beyond the recommended replacement clearance) of steam was assumed to escape to the room, environment. The non-condensible leakage from the vacuum tank relief valve is expected to be linear with backpressure. Therefore, this forms the basis of paragraph 3.3 below. 3.2.4 Turbine control valve / Pump-Turbine Performance An increase in turbine exhaust pressure up to 46 psig is expected to cause the turbine control valve to open further to provide for a given power demand. Also, the turbine will deliver less power at very low inlet steam pressures when the exhaust pressure is higher (Ref: Attachment 5).
~
The pump / turbine performance curves published by the turbine vendor already consider an exhaust line pressure of 50 psig per Attachment 3. Therefore, RCIC system capacity and flowrate are adequate during those accidents or transients in which RCIC response is part of the current Pilgrim Nuclear Power Station licensing basis. 3.2.5 Nozzle loads External forces and moments distort the turbine exhaust nozzle; pressure alone does not. An increase in allowable turbine exhaust pressure could be accompanied by a corresponding increase in nozzle loads transmitted into the turbine exhaust nozzle by the piping system. (Ref: Attachment 5) Per item 3.2.2 above, the piping design already considers nozzle loads at the piping design pressure of 100 psig; therefore, the limits provided on the turbine outline drawing are still met. 3.2.6 Automatic Operation The automatic initiation and operation of RCIC are not affected. 3.3 Environmental conditions: The operation of RCIC with an increased exhaust pressure does not result in RCIC equipment room temperatures and/or humidity in excess of the environmental service conditions in which RCIC is qualified to operate for a period of approximately 15 hours, at which time room temperatures are approximately 140*F (ref. 1, - figure 5.5). This is based on a conservative estimate of 70 lb/hr l steam leakage. I i Page 3 of 5 R WFOP.M ATION ONLY j
maw:. n . . =-- :a, . nwn ~ - . -- :w:. , -~ .~- ..~ .- - Attachment #1, S.E. # .21Lr / 1 These RCIC room temperatures were reviewed as part of the PNPS . Station Blackout report based on gland seal steam leakage into the j RCIC. room at 10 and 70 lb/hr, which bounds the expected leakage at 46 psig (Ref: Attachment 6). During an SB0 event, the RCIC pump { suction is maintained on the condensate storage tank. Cold CST water will be effective in condensing gland leakage at exhaust pressures of 46 psig in the barometric condenser. As a result, this leakage will not, in itself, cause a harsh environment. l i During a small break LOCA, when containment and torus pressures are high, RCIC is not credited in the Pilgrim licensing analyses (FSAR Appendix G, event 39); therefore, RCIC qu 'ification is not required for a LOCA environment. Qualification lhe RCIC isolation a valves is not affected since the maximum mission time for E.Q. purposes has not been increased. , During normal RCIC operation, increased turbine exhaust pressure is not expected to occur, therefore environmental profiles are not affected. 3.4 Mark I Containment Analysis l The potential for significant containment dynamic loads associated with steam condensation at the RCIC turbine exhaust discharge line has been evaluated for SB0 and small break LOCA conditions by l -- BEco, Teledyne, and General Electric. Per attachment 4, the increased _ pressure has a negligible effect and per attachment 6 the increased temperature of the torus water has no effect. t. Containment response as given by the limiting steam line break is also uu ffected by the increased operating time of the RCIC system since the total steam discharged to the torus during depressurization will remain essentially the same. 3.5 Radiological Considerations As discussed in Attachment B, extended operation of RCIC during design conditions at 46 psig backpressure is acceptable from a radiological standpoint. Off-site doses will be well below 10CFR100 limits, and on-site exposures will be controlled within 10CFR20 limits. 1 Page 4 of 5 FOR INFORMATION ONLY I
m;ywg .3,.; ; e .;z ac ... u,u ,,,,w. , m.w u , u .: c, . g ,v .,-, ,- Attachment #1, S.E. # 3 / /e /
.' D. Safety Function of Affected Systems /Comoonents The PNPS FSAR, section 4.7, states the safety function of RCIC it:
"The Reactor Core Isolation Cooling System (RCICS) provides makeup water to the reactor vessel following reactor vessel isolation in order to prevent the release of radioactive materials to the environs as a result of inadequate core cooling."
E. Effect of Chance on Safetv Function - There are an effects produced by this change on safety function. F. Analysis of Effect on Safety Function RCIC operation is unaffected for FSAR Appendix G events as follows:
- 1. Event #39 results in increased containment pressure; however, for this event RCIC operation is not credited. Extended RCIC operation for small break LOCA is evaluated in Section "C".
- 2. Events #27, 28 & 38 require RCIC operation, but do not involve increased containment pressure.
G. Summary This change involves re-calibration of pressure switch setpoints only. Modifications described allow enhanced operation beyond the original design considerations for RCIC. No potential for an unreviewed safety question exists for this change. H. Additional Comments The Emergency Operating Procedures will cover the use of the RCIC System beyond the design basis. Other Attachments to Safety Evaluation:
- 2. FSAR Table 4.7-2 (revised).
- 3. Telecon - H. Riggs (BECo) To R. Herbert (Terry Turbine Co.) dated 5/15/87.
- 4. Telecon - H. Riggs (BECo) to D. Encis (Teledyne) dated 5/15/87.
- 5. Letter-G-HK-7-202, 5/19/87 G.E. to R. N. Swanson concerning resporJes to BECo Questions Regarding the RCIC Exhaust Pressure Trip, dated 5/8/87, and inputs for Exhibit 3.02-0 of Safety Evaluation.
- 6. Telecon, Boston Edison,'and GE regarding RCIC turbine exhaust pressure setpoint modification, dated 5/18/87.
- 7. Telecon - H. Riggs (BECo) to G. Hilson (Bechtel) dated 5/14/87.
- 8. Telecopy tra'nsmittal - P.D. Knecht (GE) to P. Antonopoulos (BECo), dated 5/29/87.
Page 5 of 5 FOR INFORMATION ONLY _ - _ - _ - - d
5 8
' 9 1
o r o r l M e e J [ z z T c . S g nn a l l Y n ii V 0 e e 5 S I mm g s s s g t / / gH H m - s G t l l i p ne ne i N e aa sn g g n g r i v i v s i I S gg pi i i i i p L s s s p 0 5 e 5 e i O p 00 05 p p 0 0 v v C O l_r R4 71
- - 0 8
1 0 0, 0 o 3b 3 o 3b 5 R T 1 L 1 HL M % 1 + 5 5 5 a 4 a 2 N 1 O I T A y m p L O c r n n S a I i I r 1 1 1 1 1 1 0 f i c 5 2 1 2 5 2 5
+-
2
+~
2
+"
&+.
E R M c O l0 A C l n . i R R A g O E t ' g g 0 i
- g T P a 0 s V p R
2 C A O g i s i s 2
+ n n i
s p . E H g / 0 i . i p A 7 T - 0 0 S R N e nsp i 0 5 5 1 F 4 F A g i
^
^
0 0
/
0 5, 1 1 6
- f S
E OP L n A 2 5 1 2 00, 1 A 2 2 2 a / 0.5 - 2 - / - - P N L B L L R N 37 ^- 0 - H 0 N 0 0 0 I P A OA R - 5 T T l i
% e N i c
O C N I h h h h h i v h c c c c c e c
' R R t t t t t D t h h OO F F h i
w i w i w l i aw i w l c c i w t c S S S i S S a t t S S n t t g i i N e i e e e ne e u w w e O m w r r r er r f S S r I u S u u u ru u i u T r s s s es s r l l s A t w s s s f s s t e e s C I s n l o e r e r e r f e l r e r n e v e v e e r F I F P P P OP P C L L P I C e ) E r e P e u r S r s u u s h t s T s e g w ns N s r d i o ee E e p e h l mr np M t t h r e U e s sh gp w p l l i R n g ue ug i o s e e ah T o r ne a r ai hl l r s p s t g S l a or h u x s hh x a e e v sl i eer sl ee nl oh N t h i u ei I c c t s es emge nt n o v vt vv c( n sw cs e i n i e e u i o ue s r er np ear nru l e r l e n rl e o n rl o yl rl F dl f p i ih s me m i t ri t r a e C p p bh b ps af a b ceb ce mw I mL mL r g r ae ef e r at r at i y C u/ uI ui uir t i t u eau ea r r R PH PH Th Td p Sd S T R wt R w Pd l O2 Z TO3 E >.{- OZ
!I
- JO::&C :2;.:: N,{u:3 2;M M \ > , &:.::E , ' ?F.~:u: a.w: . a mdk. w, / c , .y=. FA 6% -
ArrAcwnr 3 3 BOS'IVM EDISON COMPANY h it #1 Unit #2 Ow PILCRIM STATION ' ] NPD TELEPHONE CALL RECORD
] NOD a TO *>% E~ R E l?T DATE 6 f 'l ME hD/*1 FROM, ,
/ dI MP /0TFICE 4(/ A/ME SUBJECT EC/bA.' e.s*3 0 2 G. N = W W 8/4fd. h5/NG PHONE N0 l~h 5 ~
siao 67zeer3/iE,YJksaw Exhssr /WEss<'AE- tet-M// l b9('Yff iWm W4v' 4 kW-rAA) ma wA4 OJM xd4/WPum - dbd(
"up.cGa Are W4M d +0nar
-Nm' aw N O ct #<e u x e W W zu S A. eAeb. r / hl NS l4 w&'d V
U / $ f % RGre // Pcz~ 1 mat 60setnx!c; Edesis .ShsiA Ex/tg5w-/Passas ' (Lustuapresssae 5 Efex"E TvA 6 i A)E PEdFo&1AvcE . PEG 9).WS)
- RG.12 MRC A. //14/ c UES/M /WESSUME Mkl6, YOOMb di= TK6lDE 0+SW6/&McHtx1e
/
S. 6vc.lcs\cp & & M -6 m AdE6 W S Action Required: ( s) (No) 4,- M8('c*'<- - M l" M N'
<$ r ,' <l W -
h 4 'M W SIGNED (pt?
.ggj DATE
.)~OdAAA 6.DASGJT!!
j' [/ e Chrono File No: 2-0.7
/
Page of Subject File No: ,
"Sco ra= x-52 ' FOR INFORMATION ONLY
z m : im;:.- .u;- ;> m . ..;. 8
.~r.:.. .e v .n:w.:. r.:::m:.:.c z.. r u m.-
. :=a.: v.:r .. .. a . ;, . u e :u. , ..n w ....w.,, .
[fTdCHM ENT - k 6 /
#1 ED' BOSTON EDISON COMDANY qAp Unit #2 PII. GRIM STATION 0 x?o t
- ] NOD 70,, - e' < d _
'h. DATE # If TIME ._#'2
& j-q rROM /[! '
,6, .6 4 Tete d dEx*ns c0xrANy/0rricE ~/s2E./'>n'E_.
SUBJECT:
dfb M f W. PHONE NO.
's' m WM f 9'
&dWJ t A ~ ~ /& a A % uAdd % RD 2 A M n A k L,. a lz r s k a J M R'rezD'f~ % n v - A M z:- : /
. sG L xhuc a kut.sja r
~' '
< /
A a .e L A w b . Y/ ud W Y3-
% d's!/J %%%%GL
& Ye - < &W M eleu L~ twJ.iwWAxuLM- '
A n- dksk e 6es4 di A / & M J a 'ad A # d : -
// / M L 1A % ,a w % A d
^
L+ b $ w AS'rsi & 5 t mm <w ~ TML %eAA'LAZ
,, a gs n m-: : ^f :
Action Required: (Yes) (No) j CopiesTo:g gf 6 Od667d dl / gg
.h gcdC pg g )
Chrono Fil No: 2 0.7 y/' // ' '
/see e Subject rile No:
BEco ror. x.31
- FOR INFORMATION pNLY
.. ..w:.w;m; xxv.: : . 2. . x. .n ce u , u zz m ra;r. . w,.. w:r-r. e. . wv: =i::c , - .Ka w ' Arr6 CHM Euf S lOb GENER AL $ ELECTRIC NUCLEAR ENERGY BUSINESS OPERATIONS GENERAL ELECTRIC COMPANY a ONE UNIVERSITY OFFICE PARK a 99 SAWYER ROAD = WALTHAM. MAS 5pCHUSEUS 02:54
)
)
G-HK-7-20' cc w/att: C.S. Brennion l May 19, 1987 P.T. Antonopoulos W. Babcock R.C. Slovic (Bechtel) J.E. Torbeck(CE) Mr. R.N. Swanson Nuclear Engineering Manager Boston Edison Company 25 Braintree Hill Office Park Braintree, MA 02184 1
Subject:
PILCRIM NPS Safety Enhancement Program (SEP) PDC 86-104A
Dear Mr. Swanson:
Transmitted herein are inputs for PDC.86-104A, " Increase in Trip Point of RCIC ' i Turbine Exhaust Pressure Switches." Attached are: (1) Inputs for Exhibit 3.02Q of the Saf ety Evaluation, and (2) Responses to BECO May 7,1987 questions on the subject PDC. This information was previously transmitted to Boston Edison Company on May 8, 1987. Please call me if we can be of further assistance. Very truly yours, Ralph G. Furgeson Services Project Manager Nuclear Energy Customer Service Enclosure (2) FOR INFORMATION ONLY = .
w:w;a;x.mwyn;../;cm:>. x m a - ..nzw .:- c .- wa
.c ~ c= . o :.' - . . . ~ ' -
. u. w.-
Mch 5 Safety Evaluation
' 2o[9 No. 86-104A Rev. O SAFETY EVALUATION i PILGRIM NUCLEAR POWER STATION APPENDIX TITLE: Increase in Trip Point of RCIC Turbine Exhaust Pressure Switches PREPARED BY: .
REVIEWED BY: A. Description of Change: o Problem: To assure normal operation of the RCIC turbine and its auxiliaries, the turbine is tripped off at an exhaust pressure consistent with the maximum ' discharge capability of the gland seal condenser vacuum pump. This setpoint also permits detection of blockage of the turbine exhaust line, but avoids spurious trips on turbine start." However, this low setpoint results in loss of the c RCIC pumping capability at a time where such ability may yet be needed. l 1 o Design Change
Description:
Increase the trip point of the RCIC turbine exhaust pressure switch from 25 psig to 50 psig (nominal) to extend the time period when RCIC is available to pump water into
.the reactor.
B. o Safety Classification and Boundary Limits C. o Procurement Requirements f i l 1 l
*~
FOR INFORMATION ONLY
mx :.:xw wn:.c::. :. ~ca anw.. a.c.:w ws:w. m w.:.n . - a , .w a a., . , .m.r .. . n.x, . Safety Evaluation Atte k.S
- No. 86-104A 3 Of i Rev. O Sheet 2 of 3 D. o Plant Imoact t
- Maintenance The pressure switches which provide the turbine exhaust trip function require testing on a schedule consistent with other RCIC trip functions. If they are proven to be unreliable, steps must be taken to procure new switches, qualified for this service.
- Industrial Hazards
- Constructability Unless new devices are used, no construction work is required.
E. Governine Documents o Revised Documents - FSAR Table 4.7-2: Change trip point of devices 1360-26A,B from 25 psig to 50 psig. o New Documents ! None o Reference Documents SIL 371 F. System Requirements See Item A of this document 3.02 - : FOR INFORMATION ONLY
Dx: ' ~LV& i ": us: a5 W:.5 y ig; .:..&. s %:.: . :. , .;, : c.. .: . ; .. i: * .;:.m; . . , ... + . , e,i.c 7-
. . ww 1 Safety Evaluation Am. s
. No. 86 -LDE. .
449 i
)
Rev. O F- l: 3 I l v. G. Analyses of Desinn Adecuacy GE has analyzed the affects of increating the RCIC turbihe exhaust setpoint and concluded that the existing RCIC turbine hardware specified exhaust line design pressure and temperature are adecuate for an in-creue from 25 psig to 50 psig. The gland seal condenser will not function properly when pool ambient pressure approaches 25 psia, but the consequences (gland seal leakage) is considered acceptable from a radiological viewpoint. H. Accectance Criteria 1360-26 A,8 shall be demonstrated to provide a switch
~
Pressure switches action at an input pressure of between 50 psig and 50 psig minus device ac cu.ra cy. . J l 1 l l
*~
FOR INFORMATION ONLY
E 7.W ~BJ~i m i TES P O m.alA u- . . w V,wan; .s:.a. .z mm mm c nw Attt k . 6 e g of q RESPONSES TO BECO QUESTIONS REGARDING THE RCIC EXHAUST PRESSURE TRIP What was the basis for the original setpoint? Q1. A. The original RCIC turbine exhaust setpoint of 25 psig was chosen to*be as low as possible to detect blockage of the turbine exhaust 1.ine without spurious trips on turbine start. 4 Q2, What is the basis for the new setpoint? A. The new RCIC turbine exhaust setpoint of 50 psig was chosen as a value applicable to BWRs with both Mark I and Mark II containments, based on m analyses of expected containment responses It is equally applicabletotosmall and intermediate b a station LOCA's, where RCIC is most useful. blackout (580) event. The exhaust pressure should also be high enough such that the new - containment vent can be used to prevent further incre event. Analysis of the containment pressurization indicates that an exhaust pressure trip setpoint of 50 psig would be reached after about 17 hours of RCIC operation when the containment pressure 571-28-17322, the FigureisC) about 40 psig. Based on revised calculations by Bechtel (calc. Therefore, use vent line is capable of removing decay heat at this point.of pressure trip of 50 psig,would not be exceeded. Q3. Are there mechanical limitations on: a) Turbine casing? , A. The design pressure of the turbine casing is more than adequate for a back pressure of 50 psig. b) Exhaust line piping design" A. The original GE specification for the turbine exhaust line is a design pressure of 150 psig and a design temperature of 325'F.
,Therefore, there is no concern with the exhaust line piping design if the turbine exhaust trip pressure is changed from 25 to 50 psig.
c) Gland seal condenser? A. The gland seal condenser will notthebe functional if Therefore, pump maximum discharge pressure The leakage is 25 is expected psia. leakage to be relief valve into the RCIC room. linear with the back pressure (see Figure 1), however steam leakage will still be condensed in the barometric for RCIC. FOR INFORMATION ONLY
WENDE@c .: a , w m : x m m .u u s m.a::e : .h::c. m:, n. c.:. + v. , w . . w a, (%tek,5 la ok 9 d) Turbine control valve? A. An increase in turbine exhaust pressure up to 50forpsig is expecte to cause the turbine control valve to open further to provi a given power demand. at very low inlet steam pressures when the exhaust pressure is higher. Nevertheless, adequate flow for SB0 and small breaks will be provided. . e) Nozzle leads? A. An increase is allowable turbine exhaust press The limits 4 the turbine exhaust nozzle by the piping system. provided on the turbine outline drawing must still be met should be confirmed. turbine exhaust nozzle, pressure alone does not. f) Environmental conditions A. Extended operation of the RCIC turbine Offsite at 50 psigdosesback pressure is acceptable from a radiological standpoint. ld would be well below 10CFR20 limits, and on-site exposures cou be controlled within 10CFR20 limits. RCIC room temperatures were reviewed as part of the PNPS S Blackout report based on Gland Seal Steam Leakage in room RCICat pump 10 and 70 lb/hr. suction is maintained on the condensate storage t Cold CST water will be effective in condensing Gland Leakage " shown in Figure 1 at exhaust pressures of 50 psig in the , barometric condenser. Are there any changes to the FSAR? Q4. A. The RCIC turbine exhaust pressure There trip setpoint in the Pilgrim is no other impact be changed from 25 to 50 psig (FSAR Table 4.7-2). in the FSAR. What other Mark I plants have made this change? Q5. Hatch 1, 2 A. Other plants which have implemented a similar change are: Brunswick 1,2. \ I FOR INFORMATION SNLY
pns m _ u..;wa. u. +,12 ;=m .,7. iA.cuc- . .u. - . . w .,.:, c u . w.. .c.. ,a.- -. c.:. . 7; , ,4 . r- . , :.1 t . .s .~ ,- 3e0 b kb. b 3 0(R 3ao - l 220 - 200 - ; I tso - 2 2 160 - c c T 140 - . m I 120 - . 100 - i ,
! l i !
l no -
} } I l l o to 20 .c 40 so i
TunsINE EXMAUT" PRE 53UmE (pe$ , u.2:4: . j Figure 1. Maximum RCIC Turbine Total Cland Leakoff vs. ?! 3 .
~
Turbine Exhaust Pressure I 42 y EORMATIONANLY
mmm%m:2:h. ;::.ww . w.s xac .=.mx n a .w: n . = a. . m:.m n ~ m ... c.w . l l MEETING lx l TELECON l l CONFERENCE REPORT A&ch. 5 8 ef4
Subject:
RCIC Turbine Exhaust Pressure Page 1 of 2. Setpoint Modification i PDC 86-104A - - . .
... _Date Issued: 5/22/87 ,
Date, Time, & Place 5/20/87 11:30 EDT File: 676 Action Item Originated By- Randy Snapp p Recorded.By: rs No. & Party
Participants:
D. Richards, Boston Edison G. Vozniak, Bechtel S. Levine, Bechtel R. Snapp, Bechtel R. Gilbert, Bechtel J. Torbeck, General Electric . P. Knecht, General Electric - E
References:
- 1. General Electric Co., " Modification Evaluation for Increase Reactor Core Isolation Cooling (RCIC) Turbine Exhaust Pressure Setting to Enhance Pilgrim's Capability to Respond to a Station Blackout", dated April 17, 1987
- 2. General Electric Co., " Pilgrim Nuclear Power Station Blackout Study (Draft)", dated March 6, 1987
- 3. General Electric Co., " Responses to BEco Questions Regarding the RCIC Exhaust Pressure Trip", dated May 8, 1987
- 1. RCIC operation at higher backpressure will result in increased steam leakage from the gland seals to the RCIC turbine room. The magnitudes of both onsite and offsite radiological effects need to be considered.
This was evaluated several years ago for the BWR owners group. Although that evaluation did not consider Pilgrim NpS specifically, the margins below the 10CFR20 limits were more than large enough to conservatively account for any plant differences. That evaluation was carried out at 220 pounds of steam leakage per hour under LOCA conditions and envelopes the maximum leakage expected at Pilgrim. l l 2. 'he increased steam leakage may also lead to room temperatures which are higher than the area high temperature isolation setpoint. This was considered in Reference 2, however, it is not clear whether that evlauation considered the higher exhaust pressure. FOR INFORMATION ONLY
~ mw.am-u ;.m..a w .w. . .s.u w & :., we u . w.. . r . m :.m u o a. . r.
. m ,..
~
- Attch. 5 i Since the RCIC suction will be maintained from the condensate' storage 9 o[9 tank during the SBO, the barometric condenser will continue to condense the steam leaking through the gland seals, etc. Although some steam may be exhausted v'ta the condenser relief valve along with the non condensibles, the total steam leakage is expected to be below the 70 -
pounds per hour assumed in Reference 2. 1
- 3. The steam condensation rate at the RCIC turbine exhaust line discharge in the suppression pool should be evaluated at the higher pressure and higher pool temperature which accompany the SB0 event in order to assess =c-the potential for significant structural loads on the torus. _.
( Steam condensation phenomena at elevated torus temperatures were evaluated in'another BWR owners group report related to eliminating the pool temperature limit from the emergency procedure guidelines. Boston Edison Co. was a participant in that study which found that torus loads do not get higher with temperature. 1'
- 4. In reference 3, GE presented the basis for the 50 psig setpoint. That basis includes a 10 psi difference between torus and exhaust line i' pressure and a maximum 40 psig. torus pressure before torus venting. At 1 this pressure the torus vent prevents any further pressure increase.
However, the maximun allowable torus pre'sure s prior to venting must also '
~
include a margin for instrument uncertainty of 8 psi. If the torus vent must be initiated below about 38 psig, additional margin ne'eds to be added for subsequent torus presure increase before it peaks. GE confirmed the above and recommended that Bechtel calculate and specify the maximum pressure before which the containment must be vented to prevent turbine trip. , Although the calculation has not yet been performed the estimated result is 35 psig which occurs about 15 1/2 hours after 5B0. When the calculation is finalized, the result will be forwarded via an FRN to PDC 86-104A for Boston Edison Co.'s use in developing the emergency operating procedures. Cc: D. Richard W. Riggs P. Knecht (GE) J. Torbeck (GE) bec: C. Reid K. Khianey G. Wilson G. Vozniak W. Smith S. Levine , R. Snapp FOR INFORMATION ONLY
23:1LT5;z-:ciFF:m.::: :a:::=.' Jit.: 5&w.:O:w: 2.EU" w.m:. u . M . .: 2.n: aca: m ' W. M . a: ArranMar 6 g ,g3 NED. BOSTON EDic0N COMPANY O 4AD 0 unie #2 PII.CRIM STATION O Noo TO . eo kAfE(HT DATE # *"s ~I .' Rjp q FROM s' f th4 6 COMPANY FFICE [N
SUBJECT:
1-4 pac _r ov' MT(* EX HM6TME65dR E PHONENO./-Md[*fdf EbET EC UJT' 6AIC d YdY kl wYf/ & t/+A Wf x m A^$6co !;I2e L n 4% Are -# Abo /Acd%a2f+ ,j u
& .t '
~ bzw : ' ' 6' o '
/
/* -*Y f && d
.D15NJ L /A dz% ~ % ,; AA>
A R a % +1 & AeA w A f a , 66 m V aaoA/DLaffAl iMan 1 W 0PSIs a l%A 'uL'L/ m~ hl1) wh2d wsL &<-.; A.4 d, Are- ., . " / A **< & Adh 9:f M M f P.//Z.d#f//)f6 bJ t Ptfc 5D A /' W ub&
.M bd'$/6 M iG z w4 ,
M1 ,k 4- <- AL<*.dtt/ /td ? MMM - Action Required: (Yes) (No) V Copies To: 6 44.$ m (/3) g/ M 4(#( 4 , [ , d DATE / 7 Chrono File No: 2-0.7 Pan j// rd / p* Subject File No. mEco ror x-51* FOR INFORMATION ONLY
i u. y..u. .xs : ;r :e.- x, .. :s., m .w . , : . a . ,. ~ n.v :.u v <- w.~..n.r:n...w.1 <. c ; :;m.:;c. v,- - vi ne .=m-hf4Uff'1SW N NED t #1 BOSTON EDISON COMPANY O o^n C Unit #2 PitcRIM STATION gpp TELEPHONE CALL RECORD
] NOD TO. b., E.C H T~ / [7 DATE TIME / 'lU/W(
TROM e' I f [G 6 6 COMPANY / FFICE b SUEJECT: lMPAC.t- op R T. C., Ex k r4WSr hf'S S UM'6 PHONE No. DET~ Fo/4f h!/f JJ -K- A (PSiG h!d ($sb a
/da E d ofk.shu d d WAedZ~suffs~0/
M4nsLetA <t1? / E& /$dd. A s M[ 5<' e _ , __
'e) # 6 E L u/ M d th f d d j]' ow} o
. A'h W A'A Y M. /At"![ * -
fed / sfA A / W ) $W r
/,
h/t'd #~ A't' / 4 - 7 7 J~w JMAu
% Maw 2J WA Jwa
-zaL2 wD/D anr !
A s Aues s k M 4 'l 4% sw D ab% 4% Ahd L ddLDM &td ML~e W 'l 4 bd <'u' &*-n'JA & HA W / Allt bf Y "C $b/ ' \ Action Requir : (Yes) (No)
.- s Copies To: SIGNED f g
[ DATE /fff Chrono File No: 2-0.7 Subject File No: p,
]
"*'*'"""' FOR INFORMATION ONLY ,
- JLEl:Mx.a m. : :a; : a uc. ' :. > .. u ..
. : : m , u . :,
. ,. - ..>......a
.. , ;_ , .. ....,c.
174' Ht'IEN f b g' BOSEN EDISON COMPANY 3 of 3 gxerrn g I Ow PILGRIM STATION
] NPD TELEPHONE CALL RECORD ,
[ NOD -
~5CNY TO ' -
DATE l$' fI / TIME 8 'ctpm FROM, e' M46 COMPANY! OFFICE dM
SUBJECT:
IN/ hic 7" of .[C b //d5 ' ESS O)#EPHONE NO. 4Er 8/NV i O f A .s st4./ /n ) a.<C &&# st " x 7 k e.sr hath m Pal AT6rea r%.
'r y /
r c8 fn & WJO^ kE
- h~ Y /.D E/ J --
h &M /d # # M
'12 Lo ~ 2 w I .A A ened uL9 ? % G M 3 R J w & G' Wa% & McAddhb x stad e M & s M' '
.3 yfL're.
24 ma./JJ2a . N % ji
~
d M A r/bl Q & J-6ll/dre-AMA~dd s A t & a% s Y=0 =
" / f Action Required: (Ye s) (No) a .__ [
Copies To: S ED DATE
/
f Chrono File No: 2-0.7 Page j /of Subject File No: Btco ror x-51* FOR INFORMATION 10NLY
MGi%TC54D:J v;. : x;,I :::. ' Li' i .Y: % . ..?.:u:' ~ ' PsaL 'u .3 n n. n: ,;.J;i &. . . . .. . w:
. .T:. .a TTACH A168.)T 7 iofI EL BOSniN EDISON COMPANY Q dit #1 Ow PILGRIM STATION C Unit #2 O Nro
, TELEPHONE CALL RECORD .
'TO MEN fL I 6>OM DATE d/ [7 TIME 8 'Od p<*H FROM , -
klk i (, 6f". COMPAh dFFICE [E(*. H 7~66
.\
SUBJECT:
N?0[ /ALD AG I C. Exy Ad_s7 .<7fAM PHONE NO./ $ 0l W 0 fi/~ /'U6 ANl6N kAWll6/$ 35=A b' l'aL&d' }7m -fr it M /~M4 -&6 mem fnMg R h->e 4 Mh sul L A MY. kL AJA-sL A HMzJ ' ' Acre -+zW / N
. A% id-dirJ1see JAL \
nb deAas usNA'& Assee sweb k1. k u~ m ' hmwdd , b ./ hwy ' 1 6 R ^JLA & /L dii n o as2' A ddM/ /) tt-
&&& GJ'Mun% k AA
~
r1in abs, M hso_p. Jmf-*f 4Ar/+A perJ Mrc A/A.
/#w eAa a ). '
/' DJ . L' n&< ink aL e m_,aA wa~e unf & n fdEM
'fM ~%fAM/AJ& AMJ Action uired s) (No) AAv<c //NW M M drs;aba& # % M Ast pop /L '
a
' SIGiED c
] 8A g [/~/r DATE O/ je Chrono File No: 2-0.7 ,/ <j' Subject File No: _ ._
Page // o f ~~
/
arca r== x-52o' FOR INFORMATION 0NLY
- .::;2c:5 m . W =. : .<wa2 ~:-i1.x cn:u a;- n a6;n:. :. .:?r.w.w. 2r :, aa .v.:m '
@ SAN. JOSE BLDG J RM 1455 TEL No. 51200 May 29.87 15:26 P.01
^ _
~
2%R.AL ELECTRIC / @ JOSE Rapifax (408 925 1687 Athhmed 8 8 31 Com 8-425- 687 .. I cf 8 milite 1GE 9251687 .........................................s........................... I TELEC0PY TRANSMITTAL , ifE: 2 y//7 TI!E: Y ,* 2 r ): g /r,y A A ,A 4 /4[ff_-+'"N~d EXT sIL CODE: _ BLDG- RM. __
- DER: 0/f='1 M ____
' M f>'*C 977 d~nr 8L GDE: BLDG. -
RM-I
- SS TRAMAL INCIAKES CCNER SHEET + 1PMIS ,
BTi: UN !=W & PN // MM \
- .I: 44 p -o r- ..
e
~
I 6 J
. l I
. su_ i
. l FOR INFORMATION ONLY I u
- w w.u.w w. ..:.w.w.s :.m.w= w w o.n:x =: ~, 5.w:w ,:x.w.: ir.x:ua -c..m.r.v.w v -
'GE SRM SOSE BLDG J RM 1455 TEL No. 51200 May 29.87 15:27 F.02 SMms.d 8 .
2 of 8 NUCLEAR FUEL & ENGINEERING SERVICES San Jose, California
- i May 29, 1987 SIE 8799 To: Petros Antonopolos From: P. D. Knecht
Subject:
RCIC Turbine Exhaust Pressure Temp Modification The following statements are provided in support of the closecut of the , design review of PDC 86-104A.
- 1. The subject modification is acceptable from a radiological stand- $
point. Attachment A provides a detailed basis for this conclusion.
- 2. The subject modification' is acceptable from an environmental qualification standpoint. PNPS EQ bases for the RCIC system is believed to-be as described in Attachment B. The subject modifica- l tion wil,1 not cause these conditions to be exceeded. ,
- 3. None of the radiological calculations in the PNPS FSAR include I leakage from the RCIC system. Therefore, the change in setpoint has no effect on these calculations. .
" Official" transmittal will be provided through Ralph furgeson, fp?c &
i P. D. Knecht Systems Integration Engineering MC 740 - Extension 56215 PDK:kp , FOR INFORMATION ONLY !
w; mum . x. . . a ;. 2 . ;.i u x.. . . e: .. ;.x
. = v.:x . w.w a: ,u a u, w GE SAtj J05E BLM J RM 1455 TEL No. 51200 May 29.87 15:32 P'08 ATTACHMENT A RCIC TURBINE EXHAUST PRESSURE TRIP hM 8 i RADIOLOGICAL CONSIDERATIONS .3 g.fg i i
The RCIC turbine gland seal steam Isakage characteristics given in Figure / l vere used as the basis for assessing potential radiological effects. 'The f design condition for notual operation is typically 10 psig at the turbine sxhaust, with the trip setpoint at 25 peig. For.this analysis the calculated doses from the steam leakage were based on operating at a dis-charge pressure of $0 psig. h e calculated doses are for the radiological effects of the gland seal leakage only and do not include the radiological effects from other pathusys (e.g., break flow or S/RV flow), which are contained by the primary containment. Consistent with the ' expected conditions f or the type of accidents under con-sideration, no fuel failures were assumed. Consequently, the controlling source term for this analysis is the " spiking" release of iodine and noble gases r,esulting from the differential pressure across the fuel cladding. In order to bound the radiological consequences of the spectrum of small and intermediate . break accidents, it was assumed that all the spiking activity is released to the reactor coolant at time zero. The following additional conservative assumptions were made: . (a) The RCIC turbine would be operated continuously for 10 days af ter the break. (b) The gland seal leak rate corresponds to a maximum exhaust pressure of 50 peig. (c) Removal of activity from the reactor pressure vessel through any other path, such as leakage through the break or steam flow to the S/RVs, is neglected. Both onsite and offsite radiological effects caused by radionuclides in the gland seal leakage were evaluated. For the of f site dose calculations the following assumptions were judged to be reasonables (a) Activity released by the gland seal leakage is uniformly mixed in the reactor building air volume. FOR INFORMATION ONLY
o .' m.m: .m x a: ....w. .- . . .. v.v . . ..#
. _. . . ..m GE SM JOSE ELDG J RM id55 TEL No.
51200 May 29.87 15:32 P.09 l
, (b) e Reector building air is removed at the rate of one ai; change per day .
through the standby 3r treatment systems ($GTS). "
, f l
(c) The SGTS filter efficiency for iodines is 90 percent. A vf a . ^A Off-Site Radiological Impact 2 3 9 With these mixing and transport assumptions, thyroid inhalation and whole-l{ body gamma doses were calculated for exclusion area and low populazion zone ? J j distances using meteorological conditions typical of one of the le*as't favorabic omestic Epstrom an atmospheric dispersion point of view.* [. i 4
..y I I J.sy j pt' The results of these calculations are sumarized by the enveloping values in the '
table below. l The dose limits of 10CTR100, which are applicable to a design basis y,j - (large break) LOCA, and 10CTR20, which are applicable to abnormal operating events ,
- i are shown for comparison, f]h.' .
g; 2-hr Dose 30-Day Dose at .i , 10CTR100 Limits 10CTR20 Limits , ; at Exclusion Low Population (Design Basis e Area Boundary (Abnormal s"; Zone Loca) 4 i Transients) Sq, 9 (millires) (millirem) (millirem) 0 Thyroid (millirem) Wi
<0 5 <1 300,000 .r1000 .k Whole-Body <5 <0.5 25,000 500
(! Because the bounding dessa tabulgted above are orders of magnitude less than . i3. l the 10CyR20 limits, it is concluded that the offsite dose consequences of the .P increased gland seal Isakage are accepts'le o for the conditions under consideration. on-Site Radiological Impact .{
"I '. 5 Several ensite situations were evaluated including an initial evaeustion time from i the RCIC turbine room, reentry to the reactor building, and reentry to the RCIC
- turbine room at the peak concentration times.
Wpper bound assumptions
)
were made to maximize the calculated dose races. For the initial evacuation of j the ROIC turbine room, the doses calculated for an assumed 10 minutes residence h{f.,: time are toout 2.5 : rem thyroid dose and .5 rem whole-body gama dose. These N y 1
.h .1 atmospheric dispersion (y/Q) values used were 1.4a10~3 sec/m3 for the exclusion q [
area boundary and 4x10~5 to 5.4x10-6 sec/m3 (time-dependent) for the low popula- d y! i tion tone. i bt - y
..f ~%.
FOR INFORMATIONa ONLY ~
w.x m,x. : .ma . m, a -c .= . .x.w: , :w . 12 . ~. a.a m. .~ . e SE SAN. JOSE ELDG J RM 1455 TEL No. 51200 May 29 87 15:28 F.04 bkbknme d 8
- f 6 of 8 :
upper bound doses are well within the quarterly occupational limits of 100TR20, 1 which are 7.5 res thyroid and '1.25 rem whole-body. ' Noble gas and iodine activity in the reactor building would increase to a time as the turbine continued to operate and then eventually decrease because of the centinuing removal of air from the building and radioactive decay. The potential inpact of the gland seal leakage on the accessibility of the building if reentry shsuld be required was estimated by using the peak post-accident reactor building l cetivity. The maximum thyroid dose rate was calculated at about 13 rem / hour with-out breathing protection. Reg. Cuide 8.15 shows protection factors up to 10' for certain types of self-contained breathing apparatus. Thus, personnel with adequate breathing protection could reenter and remain for substantial periods of time. Approximate calculations based on the peak post-accident noble gas inventories cnd conservative geometric assumptions indicate whole-body ganuna dose rates on
. the order of 0 5 rem / hour. It is concluded that both inhalation dose races and whole-body gamma dose rates associated with the increased gland seal leakage into the reactor building are sufficiently low that personnel reentry following [
~
tha event' , would be feasible for reasonable periods of time ('1/2-1 hour). R32ntry into the RCIC turbine room f:r a limited period of time may be desirable (e.5., if needed co restart the RCIC turbine after shutdown). Using similar bcunding assumptions, an analysis of this reentry at the peak concentration times v2s made. At the peak iodine concentration the calculated thyroid dose rate is abcut 1000 rem / hour without breaching protection. For the time of the highest concentration of noble gases, the calculated whole-body dose rate is about 10 r:2/ hour. Provided that adequate breathing protection were used to reduce the thyroid dose rate, reentry to the RCIC turbine room for a short time period (.e 0.1 hr) could be made to perform an essential function. In sumnary, extended operation of the RCIC turbine at 50 psig backpressure is acceptable from a radiological standpoint. Offsite doses would be orders I of cagnitude below 10CFR20 limits, and onsite exposures could be controlled within 10CTR20 quarterly occupational limits. l i
~ 1 l
FOR INFORMATION 10NLY i l
wa x;mm.=, m .. :.a= ~2x:: . m w . ..'z , wa -:ma u 7, .ww rm c.u rn 1 - - .. . .. 1 GE SA14 305E BLDG J RM 1455 TEL No- 51200 May 29.87 15:29 P.05 me bhclwed Ef ;
'f. I 6 of 6 24o - .
e. I i 220 - l noo -- l iso - 4 e
.* too -
a f i 3 14o - l l imo - , e I .I too - .
?
1 so - t i 1 1 I I : o to 2D 3o ao So Y.*
'2 TUMstNE EXHAutiPMESSUME (peit) 1393242 i
Figure 1. Maximum ECIC Turbine Total Gland Leskoff vs. ~ Turbine 7,xilaust Pressure
......- f FOR INFORMATION u,ONLY
n ..ai.- ~ . . : . , c.- , a .. .: ~~ -. : . ~.= . . <. c+. . ,~ m: <- - ; <=nmn GE SAN JOSE ELDG J RM 1455 TEL No- 51200 May 29,87 15:31 P.07 NEDE-2455 r QSQg g ~ j! Normal environment, equipment not running - 10407, 905 2 7 cfB Normal environment, equipment running v- 150cF, 905 M i ) i Abnormal environment, equipment running Turbine mounted equipment -- 2120F for 0, to 6 hours 1500F for 6 to 12 hours 1005 M Electronic controls, remote mounted - 1500F for 0-12 hours 1005 M Woodward Governor Company, supplier of the turbine control components, has identified a maximum operating temperature of 1680F, with 80% relative ; humidity for their electronic controls (any time interval) and 2120F with 100% relative humidity for up to 6 hours opemtion for the control components which must be acunted on the turbine assembly. Typical turbine control systems have been subjected to thermal aging and accident environment In:nditions in specifio qualification tests in accordance E g with IEEE-323 The electronic controls were ' aged" for 50 days at a temperature of 1850F in a non-operational state, followed by operstion in a temperature environment of 1500F, with 1005 relative humidity, for 12 hours. The remaining control components, which must remain mounted on the turbine, were ' aged' for 50 days et a temperature of 2250F, followed by operation in an environment of 2120F,100% M for 6 hours, and 1500F,100% RH for an additions 1 6 hours. 6 c O FOR INFORMATION ONLY
m.a. : ..,.z,, , , . . , , ,,
, ~'
" ' ' " " m ~
GE SMJ JOSE BLDG J PM 1455 TEL No* 51200 May 29,87 15:30 P.06 ATTACHMENT B NEDE-24955 *' 0 ek 0 ggg INTRODUCTION J This study presents the realistic equipment environmental capability for the g The, turbine HPCI and RCIC pumps, turbines, and turbine control components. control components are the limiting equipment with regard to environmental . onpability. The following information is applicable to HPCI and RCI.C systens (PNPS included) using Terry steam turbines for the plants listed in Appendix B'. CONCLUSIONS . _The realistic equipment room environmental limit for the HPCI and RCIC sy;tems using Terry steam turbines and Woodward electronio control systems is 150cF, and 1005 relative humidity for the required 2-hour period. If this environmental limit cannot be maintained in the equipment rooms, consideration can be given to remote 1y loosting the electronic control panel. The equipment supplier has identified the acceptability of a remote location, up to 600 feet removed, using 16 gage, twisted, shielded wiring for interconnecting cabling. (It must be noted that the remote location of the electronic control panel complicates the control system calibration If procedures and increases the potentia 1 for notse interference to the contro1 system.) O the electronic control panel can be remotely located for an environmental condition not ' exceeding 1500F and 1005 relative humidity, system operation with the turbine / pump equipment room reaching an environmental limit of
~
2120F and 100% relative humidity is possible for the required 2-hour period. B_ACKGROUND INy0RMATION AND DATA The following information and data explain and justify the conclusions defin above. The design basis for the equipment is: HPCI and RCIC System
* "WROC Evaluation of NUREG-0737, Item II.K.3.24:
Environmental Capabilities", June 1981 I O _ FOR INFORMATION ONLY
c , _v. _w . ,,,...~c._:
,...m.._-.
. s- . . . . ,; .n c, , , . . .o . . im.r u :c. --,--m-i Safety Epajuation No.: 11(o /
,' SAFE 1Y EVAlt!ATION WORK SHEET Rev. No. O A. Systes Structure Component Failure and Consequence Analyses.
System itructure f h onent Failure Modes Effects of Failure CImments
- i. n sq ua or, 2.
3. General Reference Material Review FSAR CALQJLATIONS REGJLATORY SECTION PNPS TECHNICAL SPECS. DESIGN SPECS PROCEDURES GLIIDES STANDARDS CX) DES 4.7 Bechtel Calc. #E-676-1 Reg. Guide 1.105 ISA-S67.04 i B. For the proposed hardware change, identify the failure modes that are likely for the components consistent with FSAR assumptions. For each failure mod 6, show the consequences to the system, structures or related components. Especially show how the failure (s) affects the assigned safety basis (FSAR Text for each system) or plant safety functions FSAR Chapter 14 and Appendix G). Prepared by IV. DAIbceeA o,,, dis /g+ NOTE: It is a requirement to include this work sheet with the Safety Evaluation. Exhibit 3.07-C 3.07-18 Rev. 4 FOR INFORMATION ONLY 4
w.~ ,.w - - - - - - . x ~ e-
- l. . x.; .ua w
. . . = a u . : u.v.. u - ..: .
-NED Proposed Change Safety Evalua91on
' No.: c2LO
$AFETY EVAL 1]ATION PILQtIM NUCLEAR PWER STATION (
'Rev. No. . O PDC PCN System Calc.
Initiator: Dent: Group: No.: Name: No.: Date:
;f.P. N#VILLS NRo PS 81'30 A7k/S , ffU/?f Description 9f Proposed change, test or experiment: #4B ## #77#S
/WI77nt2s Tk/A Ys Tha Asc/Acut.Anoo Pomp More a GSNMAA Tbn DRIVE Mowr BREAkfAr (Als/(AVo/)
1AFETY EVALUATION CDNrlIMIONS: The proposed change, test or experiment:
- 1. ()6 Does Not ( ) Does increase the probability of occurrence or consequences of an accident or malfunction of equipment leportant to '
safety previously evaluated in the FSAR.
- 2. (p) Does Not ( ) Dosis create the possibility far accident or malfunction of a different type than any evaluated previously in the F5AR.
~
- 3. 00 Does Not ( ) Does reduce the margin of safety as defined in the basis for any technical specification.
RASIS FOR SAFETY EVAf flATICNt CDIFt151MS: 8#4 477MCA/M / E4 7- #/ Change Change OfNecommended ( ) Not Recommended SE Performed by T* A NBb'/L.lA Date Tb7 lP [] Exhibit 3.07-A n Sheet 1 of 3 c-hf' a .....,, R.v. . = _ - - - - _ - - - - -
wm . = ann.n ., . .: r. - . :.:. a. +.n.<. . + , -, . .,, .. . . . No.: 2 g4~& 1AFETY EVALUATION PILCRIM NCLEAR PCNER STATION Rev. No. O A. APPROVAL (K) This proposed change does not involve a change in the idchnical ~ Specifications. (f) This proposed change. test or experteent does ( ) does not M i involve an unreviewed safety quettion as define ( in 10CFR. Part 50.59(a)(2). (.p This proposed change involves a change to the FSAR per 10CFR 50.71(e) and is reportable under 10CFR50.59(b). ( ) Comments: The safety evaluation basis and conclusion is: , J % Approved () Not Approved 7I tht/47 Discipline Group Leader /Date Supporting Discipline Group Leader /Date
*h. REVIEN APPROVAL
( ) Comments: wL dw G 2, ti
#cUY@ Group 9'ader/Dat's' C. ORC REVIEN .
() This proposed change involves an unreviewed safety question and i
. a request for authorization of this change must be filed with the Directorate of Licensing, E prior to luplementation.
( This proposed change does not involve an unreviewed safety ,, question. CRC Chatruan k / Date 6/IA/P7 ORC Meeting Nu er D-intnie3.o7-A j
..go OE'
./ ,s
\'.) C 3.07-14 Rev. 4
\0 %.
y - l
^ -
(. . .w .umc. , u : a . O. . ~.w:: v..:
. . w a:
.,.w-
~
~ - - - -
- ;- =
. t PfLGR8M STATf0N FSAR REVIEW SHEET
References:
Safety Evaluation: 218 Tf Rev. No.: O Date: Support a change List FSAR test, diagrams, and indices affected by this change and corresponding FSAR revision. Affected FSAR Revision to affected FSAR Section is shown on: Section Preitsinary Final 39S/ Attach nt 2 , O 9* 8'3 Attachment 2,
$ 6 U R S S' 9- Attachment 3 Mfff b6v## 8'f~f Attachment y M/7/4 794 2 ,
Attaeh.ent s-F/G< M 2 7 f~'L ' Attachment f M/ff-d PRELIMINARY FSAR REVISION (to be completed at time of Safety Evaluation preparation). E. Prepared by: y O /Date: 8 !U leviewed by: i - - N /Date:5/21f7 Approved by: /Date: 82f!87
/' '
FINAL FSAR REVISION (Prepared following operatione.1 turnover of related systems structures of components for use at PNPS).
/Date: Reviewed by: /Date:
Prepared by:
/
Attach completejl FSA ange Request Form (Refer to NOP). q.) ~ J f' , Exhibit 3.07-A f Sheet 3 of 3 6P
' ,j 3.07-15 Rev. 4 1
p .m .w.m.w a ,.x a mar..n m. mm w= . . c. 5 w - w m = .m =w-
.* PDC 87-30 Safety Evaluation No. 4 d Rev. O Attachment 1 Description of Chance PDC 87-30 will add an ATHS initiated trip to the 4160 volt drive motor breakers of the Reactor Recirculation Pump Motor Generator Sets A and B. This modification will provide an automatic trip of the drive motor breakers at either high reactor pressure (1175 psig) or low reactor water level (-46
- inches indicated level).
The trip signal'will be initiated from level transmitters LT263-120A,B.C and D and/or pressure transmitters PT263-122A,B,C and D. These level and pressure transmitters will operate trip units 263-121A,B,C and D on low level and trip units 263-123A,B,C and D on Reactor high pressure. The above transmitters and trip units are in place at present (PDC 79-25) and currently trip the field breakers of the Reactor Recirculation Pump Motor Generator sets. Eight spare relays, four each in Panels C2277 and C2278 will be rewired to operate in parallel with a trip output relay on trip units 263-121A,B C and D and trip units 263-123A,B,C and D. The ATHS system for the existing field breaker trip and for this modification which will trip the 4160V breakers of the Reactor Recirculation Motor Generator sets consists of the following; reactor vessel water level and pressure sensors, solid state analog trip units. Trip relays, power supplies e and redundant trip coils for each drive motor breaker have been added for this modification. The system is separated into two divisions with two separate trip channels (sensors and trip logic) in each division organized into a "two-out-of-two" for water level initiation and a "two-out-of-two" for reactor pressure initiation. A . trip signal from either division can initiate a system trip. The system will be an energize to trip system. One trip coil on each breaker shall be actuated by Division I RP1-ATHS Logic while the second trip coil is actuated by Division II RPT-ATHS logic. Cables'(routed in Seismic Class I mounted raceway) will go from C2277 and - C2278 located on Reactor Building 51 ft. elevation to the switchgear rooms located in the Turbine Building 23 ft. and 37 ft. elevations. The two divisions (I and II) will be routed in separate raceway. A second trip coil will be added to each of the 4160V drive notor breakers (A301 and A401) so that each breaker will receive a trip signal from both divisions.- TheadditionofthemotorbraagertripsmakesthePNPSRecirculationPumpTrip reliability acceptable and'high than the industry wide Recirculation Pump Tripreliabilityinitiatedby4trippingofthefieldbreakersalone. or a y
..g. 1 o, ,
rw- - - - -- -
p:. . www , , u zw n .w.w . .w<= . 'w
= - '
- x PDC 87-30 Safety Evaluation No. OO Rev. O Attachment 1 Purcose of the Chance The purpose of the Recirculation Pump Trip is to reduce core flow and create additional volds in the core, thereby decreasing power generation and limiting any power or pressure disturbance. The recirculation pump motors are designed to trip on an ATHS high vessel pressure or low reactor water level signal or by manual initiation. At present, an ATHS signal trips only the field breaker of the recirculation pump to effect a stopping of the pump. This field breaker in the past has proven to be less than totally reliable.
Changes in maintenance procedures and lubrication should result in a greatly improved reliability. To increase the reliability of a recirculation pump trip, it is here proposed to also trip the 4160V breaker to Reactor Recirculation Motor Generator set drive motor. Systems / Subsystems Affected
- 1. The ATHS system and the recirculation system are directly affected by this modification. The ATHS system is affected by the addition of eight trip output relays, each to be connected in parallel with one of eight existing trip output relays. ,.
- 2. The recirculation system is affected by the addition of a second trip coil to the reactor recirculation M.G. Set Drive Motor Breakers (A301 and A401). Two additional trips will be added (one to each breaker trip coil). This modification will not degrade the recirculation system because the logic will be energize to trip and also two out of two twice (pressure and/or low water level) to trip the breaker. Therefore, one failure will have no affect on the system. In addition, this installation will be handled as "Q".
- 3. Safety related AC and DC power supplies are indirectly affected by this -
modification. Table 1 lists the affected power supplies, breaker providing isolation, and associated loads. Table 1: ATHS Power Sucolies Egngl Breaker Lgad Reference l Y3 8 C2277 E14 sh. 1 f Y4 8 C2278 E14 sh. 1 J i D36 '2 C2277 E13 037 g 2 C2278 E13 D4 .. y- Q & 7 A3 Control Power E13 A4 Control Power E13
,,J5[, g T Q
, 'y egh N M fek 1 provide the isolation between the safety
' relatet pqer s\ipp ,d anthe non-safety ATHS system. In addition, the ,
el but C2278 provide coil to contact separation of the trip cards der of the circuit, including cables and trip coils. s ap , r V .;;.' Page 2 of 5
aa . . ma. w n .- , , awm ,a.n m:.- . - n.; . .a: . =mu . x z a.2 w - : ,- , s= n l j PDC 87-30 Safety Evaluation No. Dif f Rev. O Attachment 1 Comconents Affected - ATHS panels C2277 and C2278 will have eight spare relays, already mounted in , the ATHS cabinets, rewired so that one each is in parallel with existing trip l relays. No additional heat load to the ATHS cabinets will be added as the i relays are energized only in the trip mode. The additional load on the power supplies has been reviewed to verify that the capacity of the power supplies is not exceeded.
.The Reactor Recirculation MG set breakers (A301 and A401) will have a second trip coil added. Each trip coil (the existing coil and the new coil) will receive a signal from the ATWS system. One coil will receive a Division 1 trip signal and the other coil will receive a Division II trip signal. The t71res within the switchgear will be separated from each other as much as practical so as to minimize the potential of the failure of one division '
affecting the other division. The non Q Reactor Recirculation MG set breakers will not be degraded by the modification for the following. reasons: 1) The existing trip cards were purchased from Rosemount as "Q" items and have a proven record of reliability throughout the industry; 2) The logic will be a two out of two logic so that a single failure of.a trip card or transmitter will not cause the motor drive breaker to trip; 3) The cables from the ATHS panels to the switchgear will be routed in seismic class raceway (separate raceway for each division) to ensure that failure of one division will not affect the other division; 4) The wires within the switchgear will be separated as much as practical; 5) The new trip coil will be purchased via the CQI process to ensure the quality is the best available for commercial parts; and 6) installation will be performed and inspected per "Q" standards. As a result the Reactor Recirculation 4160V breakers and the Reactor recirculation system will not be degraded. Safety Function of Affected Systems /Comoonents Recirc system has two safety functions: a) Adequate fuel barrier thermal I margins following recirc pump malfunctions; b) Failure of piping will not - compromise floodable volume. The ATHS system (Recirculation Pump Trip and Alternate Rod Insertion) has a design objective of introducing negative reactivity to the reactor in the unlikely event of a failure to scram (see FSAR 3.9-1). The existingAe rculation Pump Trip (field breaker) together with this modification (4160V ive motor breaker trip) will supplement the ! functional performance afr,th tor Protection System (RPS) and is l independent of the S" 9.2). The RPT occurs at either high reactor pressurs (1 low reactor water level (-46 inches indicated 1 Specification pages 44a and 59a). w Qe The f c Panels (C2277 and C2278) is to process the signals (pr r or presst(re) from PT263-122A,B,C and D and the signals ( oport , eactor level)from LT263-120A,B,C and D. The panels cur' n e a trip signal to the Recirculation Pump Motor Generator Fiel . o, at 1175 psig or -46 inches indicated level. This modification i will lays to provide a parallel trip signal to the Reactor Recirculation Pump Mo or Generator Drive Motor Breakers (A301 and A401). I Page 3 of 5 1
- 1
www.ummmo ., . m w a m m m m v . a : m m . :::.v c.zz : := = -. PDC 87-30 Safety Evaluation No. 238E Rev. O Attachment 1 Effect and Analysis on Safety Functions The first recirc system safety function is to ensure adequate fuel barrier thermal margins following recirc pump malfunctions. These malfunctions are l described and evaluated in FSAR section 14 station safety analysis. The addition of the new drive motor trip coils does not introduce any recire, pump 1 malfunctions beyond those previously analyzed. PDC 87-30 does not affect the recirculation system piping and, therefore, has no affect on the second recirculation system safety function. This PDC could affect the power generation design basis of the recirculation I system due to adding new methods to trip the Recirculation Pump Drive Motor Breakers. Failures in the circuit that could cause the trip coil to be energized, could cause an inadvertent trip of the recirculation pumps. The potential failure has been minimized by purchasing the indicating lights and cable to the same criteria as safety related equipment. The trip, coil, fuse blocks, and fuses will be purchased via the Commercial Quality Item (CQI) process. The installation of all items will be performed and inspected by Quality Control personnel as one safety related equipment. The raceways will be supported per the criteria for Seismic Class 1 Raceway. ! The trip of the generator field breaker on low water level required the addition of a time de'xy which was to be set at 9 il second. This timer was added.to insure the original plant licensing FSAR basis for analysis was still bounding. That is, the pump coastdown (flow versus time) which was based on the trip of the drive motor breaker and was used in the original Loss of Coolant Accident and transient analysis would not be invalidated by the addition of the field breaker trip. The level sensor, logic and breaker time delay of 1.03 seconds is in addition to the 10 second (9 2 1 second) timer. The timers are usually set at 9 seconds to allow for instrument uncertainty. Transient analyses for Chapter 14 and reload cores are unaffected by the new pump breaker trip since the existing field breaker trip (which is included in _ present analysis) results in the faster coast down of the pump. The two trip coils (one existing and one new) have to be mounted side by side in the breaker cubicle ( Separation of the coils in the non-safety related breaker is not possjtle. ' e wires within the switengear will be separated as much as practica1 An h t to prevent failure in one circuit from l affecting he ar F. witchgear A3 and A4 do not have existing ! seismic.;g i allable and, therefore, mounting of the additthnal , g use blocks cannot be analyzed for their effect on e.MeWeret 'itchgear. The equipment will be mounted as rigid as D poksM( T s lle unction of the vital DC (Y3 and Y4) and DC (D4, 05, 036 and D37) r els will not be affected by PDC 87-30 because the isolation breakers, as,) sted in Table 1, are not being modified. During normal operation, there if no additional load on 04 and 05. Page 4 of 5 h
ws7ia ;w;;. =.v.w- _- . a w.m . , 1 a; ; 4 ;_. u; - -
. ._.s ,.a PDC 87-30 Safety Evaluation No. J 85 I Rev. O Attachment 1 Fuses are being placed in the trip circuitry (located in both the switchgear and ATHS panels) for ease of maintenance, not for separation of safety and non-safety equipment.
Failures in the trip coil circuitry cannot affect the remainder of the ATHS system because of the contact to coil separation provided by the relays in the ] ATHS panels. The logic circuitry will be installed in C2277 and C2278 which are located in separate fire zones, separated by a water curtain. Therefore, a fire in the Reactor Building on the 51 feet elevation will not disable both trip systems. The raceways from C2277 to A301 and C2278 to A401 will be separated as much as possible so as to reduce the possibility of a fire affecting both trip circuits and preventing the ATHS system from performing its design function. This modification will have no affect on ATHS Instrument Racks C2275 and C2276. j
- Summary The above analysis demonstrates that the modifications to oe performed by PDC 87-30 do ani degrade the existing Recirculation System ATHS System or safety related power supplies. The modification will nat affect normal plant operation and there is not a potential unreviewed safety question involved.
f~ S VO
# * , Y y
\f.
C
,,~ s
? ./ _ f 7
j, Page 5 of 5
= = _ _ _ _ _ - _ _ _ _ _ _ -
. % cc 4.,.c 9.v - .. . .sa N v. m . .' J.; a .u t - t .,
. _ z :.T. .- -- '
i O.f,7#9 m b e b.,;,d. { FS75-TSU 2 2.9 x::15: .x us P v.F :217 (KF:) no A:.:rw:t n= INSIx=n tuli i SYSTTM.S ( 3.9.1 Oes ;n Ct;e:*.:ve , j The des 2;n ot;e:::ve of the RF; and AR Systers is to prov:de a j ba:k .p tre hed for :ntrodae:n; negas ve react:vity to the Res: tor :n the uni:kely even: cf a fa:1ure of the Kea::or to scram from power. 3.9.2 Des:;n Eas:s The desig. bas:s for the RF* and AR Sys:e-s is as follows: mit: gate the
- 1. RFT sha*1 provide the means to help l
I co .seqsences of a f a:1ure of the Reac-or to serarr,.
- 2. RF7 shall supp eren: the funct:onal performance as well as prov:de of the ex:::.n; Rea: or 7:cto: .on Syster. (RFS) from the ORF3.
reia.nda .:y, d: vers:ty. and independence (Keferen:e Se::aon 7.2)
- 3. Means shall be provided by which functional performance ca;at:12ty of the RF~ control system coeponents can be
, ver:fied per:odi: ally under conditions approaching actual use requ2ren.ents.
- 4. These sys t e:r.2, although classified as non-safety, shall be e
( des:g .e f' and operated to provide a degree of reliability j cons stent w:th its functions. shutdown of
- 5. The poss atility of uninte .tional or accidental the Rea: or by these systems shall be minimi:ed.
- 6. di hall be diverse, to the extent practical, from the RPS.
s5 kO AR :ated scram for allcontrol initiated, CRDs sha~1 rods shallstart be with'
. 417
' < Once v$,) 's - d prior to fi*1ing the Scram Discharge Volume (SDV).
\~
'^f<o * % .; AKI funct:en shall be elec rically independent frors the !
.f $. v 1
/b%k
/ h h . The RPS.
.3 Descrip :en b 3.9.3.1 Re::r:ulation Pr.p Tr:p (RPT) System The RF- System consists of instrurr.entation and relay logic, that is
- d: verse and independent from all other instrumentation, wKich senses water level and high pressure conditions in the Reactor vessel. The RT~ system causes a " trip" of the Recirculation Pump MO Se: f: eld break.er upon dete:: ion of either high reactor pressure4.2. or low-low reactor water lev
- ond:tions. (Reference Sections
( g ste/ Ist d,Mrdh v Rev:s on 5 - Jul *951 3.9-1
~^ ~'
U.cITWmm;L 4 u.B - o. - d u UzzEiv.; 1 v.I ~rv- ?.h ; , , ;.a 2 '
. . 5; ,
FNFI-TSAR
' ~ N .. -
J,D7.A3px N a:........ i
- l. ;M : _ ..... . .,
/ =; m. ; - i H. 2 . d: m~.nw .
,e r.. ,_,-.___-e . , .;; ~._,
.- .- . ..;. .~.... - __
{ m.-
~ _ _ - }
3.9.3.2 Alternate Red Insertsen (Ni!) System As an added design feature ofseras tne RF;the Systen, ins t ru-ethe reactor through nta!. ion a.H Alterna:e relay log:c was provided to The AR: system ser>es : Rod Insertacn (AR:) system (see Reference 1). The mechanis. fer as "a d: verse electrical log 2e to the RF5 scram. valves (A and B) installed in the AR: cons.s s of two scieno:d the Control Rod Drive - Hydraulic Contro* instruttent air header of additaonal valves are redundant to the These Units (C C-H~.l). Protection System (RPS) backup scram valves. Upon entsting Reactor scram valve air supply heads is energization of either valve, the vented to atmosphere to initaate transfer of all HCUs and thus (Refer to Section 3.5 and 7.') ansertion of all control rods. j 3.9.3.3 System Trip Logic-althoug$non-The RFT system trip logic (Division 1 and Division 2),A and B batteries respective *., safety related, are powered The RPT from system is initiated through coincident (See Figure water level and/or high pressure signals from the Q(j) . recespt of low-lo Electronic Transmitters provide RPV reactor pres e vessel (RPV).
~ level and pressure signals to Analog TripThe Units trip(ATU) channels thatareare adjusted to energite auxilaary relays. (Refer to Tigure 3.9- )
arranged in a two-out-of-two-once logie. Division I Logic consists of channels A and C instruments contining in a two-out-of-t%o logic to trip Recirculating PungsARI A and 3 (Refe. function. 1 to Figure 7.9-2) and energize the A solenoid of the Y
/,
Likewise. Davision II Logic consists of channels B and D contining in a two-out-of-two logic to . trip Recirculating Pumps A and B and
, [ d. energite the B Solenoid of the ARI function. ~
y The mechanism for RPT consists of ' redundant shunt trip devices (trip [ coils) installed in each Recirculation MG Set generator When RPT field1pgic breakeT,. These trip coils are normally de-energized. t' is satisfied the trip coils are energized to trip the Recirculation trip of the R ,
!' MG set field breakers 4_and_ thus effect -
~
/ l
.. Pumps (See Figure 3.9-3A - CaJ okc e:kr hitbH '
y
; 3.9.4 References NDC-30952.
b 1. General y (ectric, P:1grirr. ATWS, Assessment, April 19ES. LRc hw- avnhnbw dF /1TW6 ygpg,,, p g
~f .
o vr ! be b/W ASCw WodUr Os w '
- h D,) ' .N $."h/(
-{
' ,/
Revision 5 - July 1915 3.9- l s
cim~.maan am:-.: .. m. m .- <.um.., . . x 7,w n ,. . . w. . .nw . .. . , m , .m. _ v.
.. , I <.
- .' -t .
i i - -
- .i ::
9- 09 ~
. . ! .x .-
f- Y .t\
- . -}i.
D g e .- . r ,1 - o c-- I :..a. -. :- 8 .l B
- y. ,
l &_u. .L. rn. - 13 - b ..i...
?
. qJ N
; s' gC lc. I
~~
vd s- t }o 4 5 O.yg i =IE - n. 4 5s-rq i<W 1 :. C . '. v, ..g ;' ,
,-, -. -ca. ,. . .
I. 3....j; :'
-., 7 ;
1 I 'f.,.ii,[i l ', p. .p 4.1. pu w , - - -- i i ,
- l
- i,. - - ..i .
1
._...E L-., A +,%.h .: j,
$.$?_#
L - rw
- 1.uZ. U m ., i :.
p
. r. __.,_4.
- rf .
P..,,., 11 <tg y, ' i t-
,I i; .- ;
g . 6.
, i. ,
'- 7,l1 gg ' '-
g 1
'I
. ij ,; l. 1,1 -
.i.t,
- d. l j '.: ' t -
,g1,( ,
.- - - . , ,..-- j .
L 4 4. . .
' 8- o g.gf g- e i g
a j +q, i l - 1 i i ... J
}l L ' *
- t. .'
. .J ' L, ".. 1 1,1 , . a o ' . ...
. l
- 11 It ll 1 ...
I, P g L .. . -=.v . l "I 7_ I e [18,ti i L i. e gI
- g. *--J ; ,l,
- l. *f
- 4 w ,1
.e i .( ,
i ei ,! L
. .ie '
3 ..- .t ( , T. f f* ' l ' 1.' i g '
! "T ,
t -
! t l- -.
gg i er } I
- i. L ., i- .T i i{
L ,t a. l,. .- o i, . _ ..- . . _1 _# '~T - <
. . l
- L .=' gj .I i j -di '~
l 1 l- :! l' .. ,'9 3, .
. L . : gg n l
'-a- 1 t
l I p= c.T I}l
. , i - I-
!] . g . - - I i lp. I f ""! .
r-1 b l "*' !
' d ( .
lj
.T j ll .
l 0 i i.
.I L -
- 3;, = f -
i t -- F _ L t . t.t.
, g I.I n!!, ,
i - -
.J i . v' , l i.f .
-- . 4 ! e,.
4 ,8 - t -e ..- -. 1.j , i e 1g
,a- +
p .4 t _.. _ <3 .--.. , j g. .I ,g ' b d F,.- 1~ . il . { ld i i
; i l
- p. .m
! ,'l t e
! li ~~ f.. ;
.'j l. - -
'l ,' 7 'T L ' i.e !
== 6.
Ig " l
.j, 1+
. w,, _ ' li M ,- '.
.i i l.- --
II _o it
.- _4 i l
g- I..j . j ,S -)- *- ! m 3g..- !
. r. .
'. : 1.--
.w i . .1, - ,. .- t; , . , . i
'
- _1 f
- e[...- m
- }f
,e ,,W _.
g
-. , ,, f ' . . ~_ ,
e A1 ._.
,1 . t. - - -
d <. *g[,>, > Iy. l: ,g/f , g.\. .. - - - n.s]. ( Q(v
. . = v.m2 ,,
(~. J'y/ I 1
. . -- - ei e
TR da f m,- MSM EGE T EE SMR TA F4 f Y S C.B S t 9 RR 2 o s S/OP WCT
# A d TE 0R I
- AR PT 1 t 7 5 0 3,
- l r 8 u o 9 7
"' t n
z n 1 K
~
1 z~
~
h" , 27 ~M r " Tz e c A o o, S ' r r u 6d 2e . l rm n *o* s F IA S O u' f le ' Sl #' o f > > ,
) y
, ), $ ba U A,H j y('
,M
*s - >S
- *r -
o 9 a.rn! eT
" c. _eu n
4 g 5u0 59 _"s n a9 rs y a, %7 s a ,
~a o
L
' "' y:" .
r nfN a v l lwn y~ O@e ' sr S r
"! e r a f
~
~
G" y-T h %e y a. s o O F
< e s A
3* 2
=
s m _ n,a j s o A 1 p
'y,h'z o )
_F* _ ;0 < b s p ,
- c, p 4 g ,
~
T ,y , 1" "'k"
~ ~
p
" h r!r,,L ,
e s p
)
(
\,a, _
N A : l
~ }' -'s T
/ < ' s t \
/.,mf', .\
,- .s .
; \
c = -
-. e& c c.
o_ -
'r n - e S n. n o n eA6." '
"' ,'a,. g.
_ ," n a W o [o- o a i , m, Df.oM m '" , _,s. o AM s- , . 3
- , 'n' c f rm)s os! nMl r -
lg ,fLa'
,rAt
,CCr" a I '; 'fl
- E -
rI4
," e f o.
,<xJs uo%
r r ec3 ,MEu"'^ sCM/ f",y,a,, > P s c,
,r
*" tA
- aE
- i
- as
- z, :s ,, 'n n
- l'
/
mi,am.m.cacy.cm m m u~ sucu. ma . # u = :. m . v. . m - . m. l .
- 5 t. 18 Mc:or Generator Se:
7.c.4.2 Auch& 5 30 7- A Ea:h me:or generator (MO) se: supplies power to its asso::ated
- re::::u' a :ng p=p meter. Each of the two M3 sets and Its con:ro.
are aden:::a'.: tnerefore, on;y one descript en is gzven. Tagure 7.i-5 shews the general arrange-en: and rat:n; of the M3 set. The MO se: can cont:nuously su;;;y power to the pu.p meter at any f'equen:;
- e:we en appronmately 19 and 96 percent of M3 set drave meter synenrone.s frequen:y. The M".; se: is capatie of starting the pu ;.
and, ac:elerat:ng it from standstill to the desired operating spest wnen the pump motor thrust bearing is fully loaded by rear:or pressure act:ng on the pep shaft after the pep has been sittang idle an a hot pressurized loop during a snutdown period. Contacts are provided that trip the re:irculation pump MG drive motor breaker on low-low reactor water level (see Figure 7.9-2). l Means are also provided to trip the MG set field breakers the Recirculation p=p Trip (RPT) system in the unlikely event tha* y { the RFS fails to accomplish a reactor scram f rom power. (Re#er Section 2.9 and Figure 7.9-2). lA L mew Baka The main components of the MG set are a drive motor, a genera or, and a variable speed converter with an actuation device to acjust output speed. ,. Drive Motor The drive motor is an ac induction motor rhich. drives the input shaft the variable speed converter. The motor can operate under of (, electrical supply
- variations of 15 percent of rated frequency or 110 percent of rated voltage. The ac power for each drive motor is supplied from a different auxiliary bus.
Generator The variable f requency generator is driven by the output shaft of the variable speed converter. During normal operation, the generator is self excited. The generator is excited from an external source dursng pump startup. Variable Speed Converter and Actuation Device The variable speed converter' ra fers power from the drive motor to The variabQ sp converter actuator automatically the generator. adjusts the slip between hh e
- er input shaft and output shaft as a function ofjf sYgia) speed controller. If the speed controller sj.gn41",ip. 19s".'.t etuator is locked to cause the speed K,rgryn ' " is." Manual reset of the actuation converterg s}1 tp ,r,e drn the speed converter to normal operation.
devicgA 1 7.9. S g ol Components The sp d trol system (Figure 7.9-6) controls the variable speed converte' of both MG sets. The MG sets can be manually controlled 7.9-2 Revision 4 - July 1954
'--- -i______m___..______,
- u. . ~ . , ~ . . .... . = > - .. .u
. ... - ~ .. -. .- . = -=.-
i SE #cp iff 1 30?G Rdocmwr6 Rm 1 A 2-mod; 4 r e, + 2 'i A 17 L 2. .shu i- 2i l I BEC DWG. M/E6-4 REV E3 ZONE J-4 THRU8 _ ADD TM/ FOLLOW /NG 70 7"Hf $707 fU' NC7/CA/ GF fWf
$16. Jf7 DMh'd 2070/'l AREARfA REAC70A jow Roc 70A pie gueron sow gguron gian mzssun wsrea ievri aren Levst PAe1 sun LT 77 LT LP PT Lp LP -BU-/22)
LP N2-//ca ast22e 24H2%M
" o o u
gr ir if if MLACTe# NMM MW LOW SKACTCA y/sp R44C70A 4 W WArgA 4gy14 PAKSSURE WATEA 41 vfl MLS$ypt LT PT LT PT LF AP LP LP to n21e 24341A0 In-/12D sts12oc , g k 1b
. . s. .. .y T '* /
,..I3 d' ' # \k C
cQ q y sroe swc.a 1
)
~
, m w a:u . . "v. m ,, . . . . .w - ,, . . . . /. e > . . . . r 4
Safety Evaluation r No.: 2 /S Ir .' 1AFETY EVALUATION WRK SHEET Rev. No. O I A. Systes Structure Component Failure and Consequence Analyses. System , Structure % nent Failure Modes Effects of Failure Comments
- 1. DRIVB /Wo D M fA/ Lune f e A 1?sts M /V K F11LD f ff. ,
8ft#4N#4 P !!" M913A N/P . 8914 4 TM/F
- 2. N1"WJ $PY FAILDA!r 6P Na Olb! Af f~ RJMzkW
).OG/4' C/wTWd 7> /N/17417sa/ AW////# floJ' ;
&Ad/A J / # # 4 L. Anwr>l Jrp/JJblE '
- 3. ORIVfh7shAAQ F4/wo#F## M 749 bil. BorND4pW4/p
. fgjp Crg. JAJNJA130 ft)PJf?sy
- Ge/L)M240 g)fg Mahdl 4pfttMALSr .
Y, Afw4 R P1~ 3/N648 *W yo SPFMef- 9e oef ef X,rs J. C S/C M *nn f4//"AA ** efhefs ya.g D 6 D ' Ceneral Reference Material Review 7"# TAlp FSAR ,. CALCULATIONS REGULATORY SECTION PNPS TEOtNICAL SPECS. DESIGN SPECS PROCEDURES ELIDES STANDARDS CDDES 9 f 1 d. Shf/ ') Z 2. Asv i /sc M J~s d z d? 9 Nnoc-3/y1.f . L
- 8. For the proposed hardware change, identify the failur6 modes that.are likely for the components consLstent with FSAR assumptions. For each failure mode, show the consequences to the systes, structures or related -
components. Especially show how the failure (s) affects the assigned ' safety basis (FSAR Tcat for each systee) or plant safety functions FSAR Chapter 14 and Appendix G). Prepared by M*'/ O DateII Pf NOTE: It is a requirement to include this work sheet with the Safety Evaluation.
,. C ' Exhibit 3.07-C
' [,
f p.f %.\ 49 3.07_i. Rev. 4 O CO
- a. : - ':< .
.:.. ... , .: .x . . , - , -.- . .. . a. > - - -
3 NED Droposed Change Safety Evaluatien No.: 2/33 Sheet / of 21 PILGRIM NUCLEAR PWER STATION Rev. No. PDC PCM System Calc. Initiator: Dent: Group: No.: Name: No.: Date: T.M.Hauske NED T C ; ".C 86-52A Residual Nh 6 Heat Removal Description of Proposed change, test or experiment: Replacement of the RHR Containment Spray Caps SAFETY EVAlijATION CONCLUSIONS: The proposed change, test or experiment:
- 1. (x) Does Not ( ) Does increase the probability of occurrence or consequences of an accident or malfunction of equipment laportant to
- safety previously evaluated in the FSAR.
- 2. (x) Does Not ( ) Dois create the possibility for accident or malfunction of a different type than any evaluated previously in the FSAR.
. 3. (x) Does Not ( ) Does reduce the margin of safety as defined in the basis for any technical specification.
BASLS FOR SAFETY EVAt11ATIM MIMIMS: See Attached Sheets Change Change (x) Recommended ( ) Not Recommended SE Perfonned by - Date 5!6 F7 l Exhibit 3.07-A Sheet I of 3 3.07-13 Rev. 4 d
w w m m,.m u= w. x mx. . . = ~ . .
.. , , . . . a ua. u. =c o ~
Safety Evaluation No. 'l f 1 7 PageJof32. SAFETY EVALUATION OF REPLACEMENT OF DRYWELL SPRAY HEADER CAPS A. Description of Chance I The design change replaces the 104 upper and 104 lower drywell spray header caps. The torus spray header will remain as is. The replacement spray caps are identical to the existing spray caps except that each of the replacement 1-1/2"-7G-25 Fogjet caps has one open spray nozzle and six spray nozzles blanked off, whereas each of the existing p f/f/I, 1-1/2"-7G-25 Fogjet caps has all seven nozzles open. B. Puroose of Chance g4 The replacement of the RHR containment spray header caps will result in' ( ,
@e h:da. This flow reduction will minimize the possibility of 4 % g,,
damaging the drywell structure by sudden decompression following the E Fy 4. 4, initiation of drywell spray. The reduced spray flow rate reduces the risk of structural damage from an inadvertent spray initiation in a hot dry atmosphere. This also increases the availability of sprays for severe accidents. (sw Mrwe.Jr- Q
~
C. Systems. Subsystems. Comoonents Affected
- 1. The system that is directly affected by the change is the Residual Heat Removal system (RHR). The applicable documents are General Electric System Specifications 21A5790AR, 22A1430 and 22A1430AE.
The subsystems that are directly affected by the change are the drywell spray, the torus spray and the RHR suppression pool return valve. The torus spray flow will be affected only slightly, and the suppression pool return valve needs to be open during containment spray so that rated flow through the RHR heat exchanger will be maintained. The components that are directly affected by the change are the RHR Drywell Spray Header Caps. Six of the seven spray nozzles are blanked off.
- 2. The systems that are indirectly affected by the change are the to.us-to-containment vacuum breaker system and the reactor building-to-torus vacuum breaker system. The response time of the vacuum breakers is affected by the reduced drywell spray flow.
D. Safety Functions of Affected Systems /Comoonents
- 1. Residual Heat Removal System The RHR cools the suppression : col water and provides for containment spray cooling (GE System Spec 22A1430, paragraph 3.1.3). It is used for a wide range of postulated LOCAs as well as MSIV closure, struck open relief valve, and alternate shutdown events. Impact on previous safety analyses are limited to those that utilized the containment l spray mode of the RHR operation (FSAR Figures 5.2-2 to 5.2-7).
j l
L: .us ;u.- u : ' a c. .R .
. :r m a.. . a:/ .
' " sa .
. a: . . a 'in: ,xiii aG;=
-, o 4
Safety Evaluation No. A ISE l I PageJofg
- 2. Drvwell Sorav and Torus Sorav Subsystem The Drywell Spray Cooling subsystem provides water to spray header systems located in the drywell and suppression chambers. Under post-accident conditions water pumped from the suppression pool through the heat exchanger may be sprayed into the drywell and the suppression chamber to remove the energy associated with the steam in these regions (GE System Spec 22A1430, paragraph 4.1.3). The containment spray is used for a wide range of LOCAs. Impact on previous safety analyses is limited to those that utilized the containment spray mode of the RHR operation (FSAR Figure 5.2-2 to 5.2-7).
- 3. Vacuum Breaker Systems The safety function of the vacuum breakers is to equalize the pressure among the drywell, suppression chamber and reactor building so that the structural integrity of the containment is maintained (FSAR Section 5.2-3.6). For accidents such as those presented in FSAR Figures 5.2-2 through 5.2-7, a reduced drywell spray will mean lower !
rate of drywell depressurization, resulting in delayed opening of the vacuum breakers. This delay is included in the analyses of Section F and has no deleterious consequences.
>I . Effect on Safety Functions 1 The proposed change would affect the safety functions identified in ,
Section D as discussed below: I
- 1. Effect on Drvwell Resconse The propcsed change raises the concern that the reduced drywell spray may not be sufficient to remove the post-accident energy deposited in the drywell, causing the drywell atmospheric and the structural temperatures to exceed their respective design limits of 340*F and 281*F.
- 2. Effect on RHR Heat Exchancer Efficiency Because of reduced drywell spray flow, there is the concern that the total flow through the RHR heat exchanger will also be reduced, resulting in decreased heat removal capacity through the RHR heat exchanger when it is used in the spray mode and higher long-term containment pressure and temperature.
F. Analysis of Effect on Safety Functions
- 1. Effect on Drywell Resoonse In Reference 2, GE reported the result of their evaluation of the reduced containment spray. GE reanalyzed the FSAR containment response for break sizes ranging from 0.02 to 0.5 ft.2 assuming the reduced containment spray was initiated 30 minutes after containment pressure reaches 10 psig. This assumption is consistant with the present
n m x.. u n w wr:."zn =r n w -
.~ ' - o wc .m ; ' s w.- .c c 1 . t Safety Evaluation No. 1 / U i l
Pageyofg 4
- 1. Effect on Orvwell Resconse (cont.)
FSAR requirement. It was determined that a containment spray flow rate of 300 gpm is sufficient to reduce the airspace tempera,ture to below 281*F for all break sizes analyzed. Holding airspace temperature below 281*F is essential since it eliminates the driving
' force for the wall temperature to exceed 281*F the design temperature of the containment liner.
The containment spray flow with the proposed design, with one header l j operating, has been calculated (Reference 1) to be 543 gpm when the l suppression pool bypass valve (1001-36A,B) is open with total RHR flow j limited'to 5000 gpm and 1150 gpm when the valve is closed. It is ' concluded the containment spray from one header will deliver sufficient flow to maintain the design temperatures in the drywell during a LOCA.
- 2. Effect on RHR Heat Exchanaer Efficiency When operating the containment spray, ne operator will be instructed to open the RHR suppression pool bypass valve (1001-36A,B) so that rated flow through the RHR heat exchanger will be maintained. This 3 l
assures that the heat removal capability through the heat exchanger will not be reduced. G. Summary This safety evaluation has identified two safety issues arising from replacement of drywell spray header caps: (1) effect on drywell responses and (2) effect on RHR heat exchanger efficiency. These safety issues were addressed by (1) performing analyses to show that the reduced drywell spray flow will still maintain drywell atmospheric and structural temperature below their respective design limits and (2) sequiring an operator action which will be incorporated into the revised operating procidure to keep the suppression pool bypass valve open while operating in containment spray mode in order to maintain rated RHR flow through the heat exchanger. No otner potential safety issues were found during this ; safety evaluation. Thus, no conclusion of the FSAR is affected and there is no reduction in margins of safety. No new accident is introduced nor is the probability or consequences of previously analyzed accidents increased. H. References
- 1. Bechtel Flow Analysis Calculation 17322-M-660-1.
General Electric Letter G-HK-7-157, dated 4/20/87, " Safety Evaluation 2. of Proposed Capping of Certain Drywell Spray Sparger Nozzles". 1 x _ - _ _ _ _ _ _ _ - - _ _ - _ _ __ _
,:r *
. .a y., e s ~.,.. , > > -- -
.;v : _a
- 4. T-- -- cm-' ms w. ;_
q l Safety Evaluation No. S I 3 3 Page5ofy ATTACHMENT 1 RECOMENDED FSAR CHANGES . The pages of the following sections, tables and figures of the FSAR that need to be updated due to the Containment Spray Header Caps modification (PDC 86-52A) have been marked with suggested updates and included in this attachment for review. FSAR Sections: 4.8.5.5, 5.2.3.2, 14.5.3.1.2 FSAR Tables: 14.5-1 FSAR Figures: 4.8-2, 5.2-1, 5.2-2, 5.2-3, 5.2-4, 5.2-5, 5.2-6 5.2-7, 5.2-8, 6.4-3
= - ~ - ~ ~
, ; ,. c., o t . ; . > _. . a , " .; n . ' .7 D b&fety Lvosustivia No.: 2/33
~
SAFETY EVALUATION WORK SHEET Rev. No. D Systen Structure Component Failure and Consequence Analyses. Mb0 N A. Systen Structure Commonent Failure Modes Effects of Failure Comments O
- 1. RHR_Conta_inment DW Temo 2_8QF ,See Attached Shee_t Sorav Header Caos
- 2. RHR Containment Elevated Suo- See Attached Sheet Spray Header Caps pression Pool Temp.
RHR Containment Pluqqing of See Attached Sheet 3. Spray Header Caos Nozzles. General Reference Material Review CALCULATIONS REGULATORY FSAR SECTION PNPS TECHNICAL SPECS. DESIGN SPECS PROtiutlRES (EllDES STANDARDS CD
. 4.8 ___
3ed d d S. /7512 d-sto-/ 10CFRSO s.? GL La f(a, 0 Ik-7-Ir v 14 4
- 8. For the proposed hardware change, identify the failure modes that are likely for the components consistent with FSAR assumptions. For each ;
failure mode, show the consequences to the systaa, structures or related components. Especially show how the failure (s) affects the assigned safety basis (FSAR Text for each system) or plant safety functions FSAR Chapter 14 and Appendix G). Prepared by N u Date Jf r 8 7 NOTE: It is a requirement to include this work sheet with the Safety Evaluation. Exhibit 3.07-C ; 3.07-18 Rev. 4 l i _ _ _ _ _ - - _ i
= r ;;.gz m ;;;;;;. ~ g :;;.,; . .: =n . . .
, x- e , x .:. 1 I
. j Safety Cvaluation No. il3 3 Page / of 32,,
' I l
SAFETY EVALUATION WORK SHEET 1 A. System / Structure /Comeonent Failure and Conteavence Analyses The following apply only to the proposed modification per PDC 86-52A:
- 1. System / Structure / Component: RHR Containment Spray Header Caps I
Failure Mode: Inability to maintain Drywell Temperature below 281*F during small steam line break from reduced spray flow rate. Effects of Failure: Drywell failure from thermal stresses. Comments: Analysis performed by General Electric to assure drywell wall temperature below 281*F during small steam line break with reduced spray flow l rate.
- 2. System / Structure / Component: RHR Containment Spray Header Caps l
Failure Mode: Reduced RHR heat exchanger flow rate. Effects of Failure: Elevated pool temperatures which could lead to loss of ECCS pumps. Comments: Procedure will be implemented to open suppression pool bypass valve (1001-36A, B) when in containment spray mode to ensure design RHR heat exchanger flow rate. Since pool rise Temperature is monitored in the control room and the rise is relatively slow, adequate time i exists for operator to open bypass valve or to place other RHR loop in torus cooling mode.
- 3. System / Structure / Component: RHR Containment Spray Header Caps Failure Modes: Plugging of spray headers.
Effects of Failure: Inadequate control of Drywell temperature. Comments: RHR water is maintained free of particulate and q contaminant spray piping is periodically purged ! with compressed air. System is thus clean and plugging of the nozzles is not probable. Since only one spray header is required, a redundant system is available. 1 1 ( A l
= w y ...o. . . m , w, ::.; w ,1
- w. w: n. .- .
sa - , ~ ..v. mai
. t 4
Safety Evaluation SAFETY EVALUATION PILGRIM WUCLEAR POWER STATION l Rev. No. () A. APPROVAL , 59 r s oF 3 L QQ This proposed change does not involve a change in the Technical Specifications. - { (k) This proposed change, test or experiment does ( ) does not N involve an unreviewed safety question as defined in 10CFR, Part 50.59(a)(2). (y This proposed change involves a change to the FSAR per 10CFR 50.71(e) and is reportable under 10CFR50.59(b). () Connents: bbrenw) e m = Yr M
- for ~ a) u a J.~ u s J.u- l
& A. ememy rqa. ;, e su n n.r e esi
- w The safety evaluation basis and conclusten is:
(>d Approved () not Approved Ju P-infbfy Discipline Grdup Leader /Date k $/~S/~/~? Supporting Discipline troep Leader / tate
- 3. ' REVIEW APPROVAL .
() Comments: - YW YV SSA Gro6p teeder/ Bate V
'C. ORC REVIDi
() Thisprohsedchangefavolvesanunrevisuedsafetyguestianand a request for authorization of this change umst he filed with , the directorate of Licensing, ISC prior to tuplammtaties. (d This posposed change does met involve an unreviewed' safety question. ORC Chairmen k !! Date 6[A 7!87 ORC Meeting Number 77 -Il9 cc: Exhibit 3.07-A Rev. 3 Sheet 2 of 3 1 L________ _ _ _ _ - - _ _ . _ - _
7 >, ; m s7, y2.v,: u t 1,; # . : 7 . - ,....;, ..
, ,- ;,x . f:z ; , ...
q PILGRIM STATION Sheet '7 of JL FSAR REVIEW SHEET
References:
PDC 86-52A Safety Evaluation: 1/f3 Rev. No.: O Date: MJ 7 l Support a change to the RHR Containment Spray Caps ). List FSAR test, diagrams, and indices affected by this change and corresponding FSAR revision. Affected FSAR Revision to affected FSAR Section is shown on: Section Preliminary Final 4.8.5.5 Attachment 1 5.2.3.2 Attachment 1-14.5.3.1.2 Attachment 3. 1 Table 14.5-1 Attachment l' Fig. 4.8-2, . 5.2-1 to 8,. - Attachment 1 6.4-3, s'14.5-5 to 8 ~ Attachment 1 PRELIMINARY FSAR REVISION (to be completed at time of Safety Evaluation preparation). t Prepared by:Nd /Date: 5.sh7Reviewedby: /Date: / Approved by /Date: J ~) FINAL FSAR REVISION (Prepared following operational turnover of related systems structures of components for use at PNPS).
)
1 I Prepared by: /Date:. Reviewed by: /Date: l l Attach completed FSAR Change Request Form (Refer to NOP). Exhibit 3.07-A Sheet 3 of 3 l
)
3.07-15 Rev. 4
m .:....... u.+- u; % 5'. z . ; Y a7;;%: M.
- x;~ ,e -
- .u.. x ? i:a M -' c
]
Safety Evaluation No. A / 3JT j Sheet /e of __?2L ATTACHMENT 1 RECOMMENDED FSAR CHANGES The pages of the following sections, tables, and figures of the FSAR that need to be updated due to the containment spray Header Caps modification (PDC 86-52A) have been marked with suggested updates and included in this attachment for review. FSAR Sections: 4.8.5.5, 5.2.3.2, 14.5.3.1.2 FSAR tables: 14.5-1 FSAR Figures: 4.8-2, 5.2-1, 5.2-2, 5.2-3, 5.2-4, 5.2-5, 5.2-6, 5.2-7, 5.2-8, 6.4-3, *' - _, _.._ _, 1?.5 ', it.5-S 4
,m w +. ......: -a r _ ;.
r.; . - - - .<. - s- n rinmo ll of Q Coolant flow to the RHES heat exchangers from the Reactor Building Closed Cooling Water System is not required ineediately af ter a 1,0CA
' because heat rejection from the containment is not,1,necessary during the time it takes to flood the reactor.
4.8.5.4 Suppression Pool Cooling Subsystem of the RHRS and is placed in The SPC Subs: stem is an integral part operation to remove heat from the pressure suppression Thispool to reduce system is pressure in the primary containment following a I.OCA. also operated as required during planned operations to control suppression pool water temperatures within the limits assumed in the Station Safety Ana7 ysis. With the RHRS in the SPC mode of operation, the RHRS pumps are aligned to pump water from the suppression pool through the RHRS heat where cooling takes place by transferring heat to the exchangers, The flow returns to Reactor Building Closed Cooling Water System. to the the suppression pool via return lines which discharge below " pool surface. The RHRS in the SPC mode functions to transfer heat from the primary containment to the Reactor Building Closed Cooling ItWater System is concluded thereby lowering the primary containment pressure. and 3 are satisfied by this mode of RHRS safety design bases 2In the event of reactor vessel isolation, the RHRS in the operation.SPC mode is capable of maintaining the It istorus water temperature concluded that below 130*r for at least 2 hr of RCICS operation. of RHRS power generation design basis 2 is satisfied by this mode A' operation. 4.8.5.5 Contairaient spray Subsystem ' The Containment Spray Subsystem provides contai:unent spray capability - alternate method for reducing containment pressure following a ' N.' 8" ' b %as an puroped through the RHRS heat excharwgers can be P TheQwater headers in the drywell and above the suppression diverted to _ spray - Thebups.ygneareragin thereby lowering the drywell condense any steam he Pe*48c" d + k pool. drywell containment presskre. - RHR5Acat exist in in the bottom of the drywell until the water level Cechger spray collectslevel of the pressure suppression, went pipes where it rises to the {Jg g g overflows, and drains back to the suppression pool. Approximately 5 percent of this spray flow may be directed to the suppression chamber spray ring to cool any noncondensable gases collected in the suppression pool. The Containment Spray free volume above theenergy f roen the drywell by condensing steam, remon Subsystem willthereby making available the drywell volume to accomanodat guantitles of gases from any postulated metal water reactions above that which the containment can inherently accournodate without spray. I The containment spray mode of the RHRS cannot be operated unless the level inside the reactor vessel shroud is above the two thirds core height set point and the drywell pressure exceeds 2.5 psig. ahES ,PoN8ch Ve,hh5 fo he, Sqpphessch pocl '
- 4.8-5 V8] the- Supphesssoh pool %s) f m C. ,
~ ~ - " - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _
.. + .. ; ., ~ - . . :. : .a.x . . . .. .,;. -. ; ; ,~ s:x.c-
- t2 e{ 64
~
g , .. . . . . .
.g,, p. . . . ( .u.c...........;,,=... ; ,
o
,,g.,
,imm *-
gg 1 .
,I .i t I '.M* i .
I "' ~ 3: " l I gi Il I!f bl 3 - g .,_"p_!!W.@h L "
!!ds,
! !! i . . !W __ ".,." "_ _
i
;p 3 _
_ i g_ p__._ _____ u ,, o , > .<_ _ _____ _____ p____ a _;ic.____ ___ ___ . ,o l_g_ _ _____ ii j. i!._______ yl)I.-. j ,,,,,,, M 1 6 -- . _ _. - - gl "* _-_____.- e I T 6.g ,. _ ___.__ _ ____ _
!".u_ I l 4 'Z' ,
h*y ; ] '_ ._, li-S
* .p,[fr -
*- --* ;9 d h*-$. -
l.j e J - .
+
- c. ._a.
.- m t r w
..-g
+.._+ 4 , r-y , tr i
*J +m
- '+-r. -
11 ( .:r { III g.. .
., 7j ,.._g.
5-d - - ' --- 8r.-?- ;;; I *I ,.::: Im_y-.: IT:r I m,'I *L :h:: tnd . i u.; a I
- AI. 1::::s-:
v._ y a 4 : u_ 7 -_+. . g... n-.. ( I__o ,, .
._p,l '
- e
.1.. ... _ ' / p.. --- I ._'
u. l f.' l L_4 ._42 _ , l
':ue. ] --_7, ,-..-f+ .. . .
l ..+ .-.. _ , i j!
' [ -
,.._A. ;_ : L
_;. _l + ..-+- _e_] l
.r. __' y
,4 _
i .._. r_. . -- 4_2. ,'
.a l;ti ll_F+
l iff l Itl .l !#: .l U nd: :i:: - - - - - - IM[i09 [,a'i 2 1
~
l,! b) l $iN l111 Did E d a " ass _, i
- i. I w!; Tim,Mn!!.p i o . _
e . . I
,s -
- -1 ,
, i, ,
I.- - ep(A,5 10, - d, 6 - u l .
) q gn - . ;, .
~
I l ki / " el1. i i , f' *' 4h
- l s '
l i;lJ. T i ' i @ j!D ii h lvi l 1-+i: .j t L p.4
,y 6 8 '
( _ ,, l , J I ( ': ! *GI Oi l{ Q l 4t: ti. _ I lQ !
' v, i , .
j - s_'W7 g 1 i 1 tl
~
z l
. it :41 j.k t
. i.4 - - i , i .
q,
. , . , . 3
. 5 i.) i
\ l t ,
v ,
4 % .' .'
, , ~ ..:. ,,, '. ^', - . r .2. 's',
. l 'i ' ' e r T.'.W.Wm' A - 7~~
1 ] o -f g PNPS-FSAR
'4 The Reactor Building Closed Cooling Water System (RBCCWS) piping supplying the drywell coolers will be revised to seismic Class I to maintain the pressure boundary integrity of this piping under seismic loading. Refer to Section 10.5.5.1. The drywell coolers were originally purchased as seismic Class I equipment to se rve as pressure boundary only.
The PCS design loading considerations are given in Section 12 and Appendix C. The Station Safety Analysis presented in Section 14 demonstrates the effectiveness of the PCS as a radiological barrier. In addition, primary containment pressure and temperature transients from postulated DBAs are also presented in Section 14. 5.2.3.2 Drywell The drywell is a steel pressure vessel eith a spherical lower portion, 64 f t in diameter, and a cylindrical upper portion 34 f t 2 inches in diameter. The overall height is approximately 110 ft. The design, fabrication, inspection, and testing of the drywell vessel complies with requirements of the ASME Boiler & Pressure Vessel Code, Section III, Subsection B, Requirements for Class B Vessels, which pertain to containment vessels for nuclear power stations. The drywell is designed for an internal pressure of 56 psig coincident with a temperature of 281*F with applicable dead, live, and seismic. loads imposed on the shell. Thus, in accordance with the ASME Code, Section III, Code Case N-1312-(2), the maximum c.vwell i1 ! pressure is 62 psig. Thermal stresses in the steel shell due to I temperature gradients are taken into account in the design. Special . precautions not required by codes were taken in the fabrication of the steel drywell shell. Charpy V-notch specimens were used for impact tasting of plate and forging material to give assurance of proper material prcperties. Plates, forgings, and pipe associated with the drywell have an initial NDT temperature of O'F or lower when tested in accordance with the appropriate code for the materials. It is intended that the drywell will not be pressurized or subjected to substantial stress at temperatures below 30'F. The drywell is enclosed in reinforced concrete for shielding purposes, and to provide additional resistance to deformation and buckling in areas where the concrete backs up the steel shell. Above the transition zone, the drywell is separated from the reinforced concrete by a gap of approximately 2 in. Shielding over the top of the drywell is provided by removeable, segmented, reinforced concrete shield plugs. I In addition to the drywell head, one double door air lock and two bolted equipment hatches are provided for access to the drywell. The locking mechanisms on each air lock door are designed so that a tight seal will be maintained when the doors are subjected to design pressures. The doors are mechanically interlocked so that neither I .. _. door may be operated unless the other door is closed and locked. The 5.2-3 6
m wr.pr?-a -: . a.. .c a .. a m x .. ~= . . ,.. ,. .
. . u. . . .
. N j f1 PNPS-FSAR drywell head and equipment hatch covers are bolted in place and sealed with gaskets.
The spectrum of primary system leak rates up to a double ended blowdown of a recirculation line has been analyzed relative to the temperature and pressure response of the drywell. Steam issuing from a leak and expanding at constant enthalpy may result in a superheated containment atmosphere. The maximum amount of superheat possible is a function of both the source pressure (reactor pressure) and the receiver pressure (drywell). The enthalpy of saturated steam goes through a maximum value at a reactor pressure of 400 to 500 psia. Steam issuing from a leak at this pressure will result in the maximum superheat for a given containment pressure. If a steam leak occurs, the containment pressure and temperature increase at a rate dependent on the size of the leak, containment characteristics, and the pressure of the reactor. The containment pressure and temperature rises as noncondensable gases are swept into the suppression chamber. Containment pressure levels off after all noncondensable gases are driven into the suppression chamber. The containment shell temperature rises as steam condenses on the relatively cool wall. When the drywell shell temperature reaches the
~
saturation temperature dictated by this c:entainment pressure, steam condensation is terminated. The only energy available to further increase the vall temperature is the superheat energy. The result is a decrease in the rate of temperature rise of the drywell shell and an increase in the bulk atmosphere temperature of the drywell. I Figure 5.2-1 illustrates the reactor vessel pressure response to steam leaks ranging in size from 0'.02 to 0.50 ft2 Figures 5.2-2 through 5.2-6 illustrate the containment response to steam leaks covering the'same size range. "'h: ti-^ it t-'-= Far the a-t ri. d.t
==11 M ::;;h 2 01 * =' r -
'" - ti:n wf 1..i .i . i: - " ~ _
t P r;r th; ;-rl;t; ;p;;t. w. .f . t. - 1;;h ;i:::. +he y rir-t4=> for *b" rent irr ^t "* t r" ^ * 15 *1"*y" ara **-- th =
-. 00_ s,c. The response of the containment to small steam leaks is slow, but the continued high reactor pressure results in high containment temperature, given enough time,. Leaks so small that the high drywell pressure trip does not occur will not result in a high temperature. Leaks large enough to result in a high containment temperature will be large enough to sweep air into the suppression chamber and result in significant drywell pressure increase. Large leaks will either depressurize the reactor rapidly or result in auto-relief such that steam temperatures above 281'F do not persist long enough to be of concern. Fis r; .2-7 shows me n .. 11... L.;;h s-ectrr 4 interent vu e4=> *a ca=rh - "-11 t: ;r;t... wf In ^ r ,
ef^ 7e'e~,L;The breax au.s. fw. dich 4. .11 rae.= 4e 4n .wr.-- ohi^ t:rp;ratur^ kif -- th:n 201"r ;t;-... t;.g:ratur" Er- = ,
==4h1=
f g ruffici:nt durati: : te of f e s. i ibe sell amy;;:ture, tre rf :::h
. 4: th:t th; ;per:ter ein t:h: :;r.nl ec ti;r. aithir. ;0 mm .f te r-an incidane *"A m """* *bt c^^t ri=:nt ^. .11 f r;; ;;;r ;;;;; 'in" %
M 5.2-4
- " .6 ~~
~
... .. a_.
~
. x , u.. w -
PNPS-FSAR m Activation of one of the two containment sprays r y tir: 5: Orc um
'"11 *-;n... . .dm. ;;;*r would be effective in terminating the
temperature rise because the superheat is quickly removed. The spray nozzles are designed to give a small particle size, and the heat t rar.s fe r to the subcooled spray is very effective. Since the total amount of heat in the dryvell atmosphere is low relative to the spray rate, the containment atmosphere temperature is quickly reduced to near the spray temperature. A drywell pressure condition exceeding 10 psig was selected as the basis for determining when to initiate the containment spray. See Figure 5.2-3 for time required to reach 10 psig. 0;..d e. Um
- al;; :: ri;ur: 5.2 7,*The operator will be instructed to initiate the containment sprays if containment pressure exceeds 10.psig for longer than 30 min. This procedure will ensure that the containment wall never exceeds 281*F. Depressurization of the reactor vessel can take place at the normal rate, but depressurization is not required to ensure that the wall temperature remains below 281*F. The environmental conditions considered in the design of the reactor protective system instrumentation, engineered safety feature equipment, and the qualification tests that have been conducted are described in Section 7.1.8.
5.2.3.3 Pressure Suppression Chamber and Vent System 5.2.3.3.1 General (() The pressure suppression pool, which is contained in the pressure suppression chamber, initially se rves as the heat sink for any postulated transient or accident condition in which the normal heat sink, main condenser, or Shutdown Cooling System is unavailable. l Energy is transferred to the pressure suppression pool by either the discharge piping from the reactor pressure relief valves or the Drywell Vent System. The relief valve discharge piping is used as the energy transfer path for any condition which requires the operation of the relief valves. The Drywell Vent System is the energy transfer path for all energy releasee to the drywell. of all the postulated transient and accident conditions, the instantaneous circumferential rupture of the reactor coolant recirculation piping represents the most rapid energy addition to the pool. For this accident the vent system, which connects the drywell and suppression chamber, conducts flow from the drywell to the suppression chamber without excessive resistance and distributes this flow effectively and uniformly in the pool. The pressure suppression pool receives this flow, condenses the steam portion of this flow, and releases the noncondensable gases and any fission products to the pressure suppression chamber air space. 5.2.3.3.2 Pressure Suppression Chamber The pressure suppression chamber is a steel pressure vessel in the
shape of a torus below and encircling the drywell, with a centerline vertical dia of 29 ft 6 in and a horizontal dia of 131 ft 6 in. The 5.2-5
7, , y- g.yp;. : a. ;._. . ..c:: 7 ; y. 72:7::;.: .
, .. ,,,y
_, - , ,;;w . : .: v;. ; ;xg , s.c:v.;.
.- .l Y
/c *{ 32 1
I
.'.' 2.
~
Note: The following figures, Figures 5.2-1 through 5.2-6. and Figure 5.2-8 are to replace Figures 5.2-1 through 3.2-6 4 and Figure 5.2-8, respectively, of the FSAR. Figure 5.2-7 of the FSAR should be replaced with a blank and marked as such. e e _.-mC___ - _ _ _ -_ _ _ _ _ . _ _ _ _
7- . .. ;. ::: .~. :~ ...v... --c. . . .- n .. ,, -
.c .:, , ,; x r w-U #[ M 1400 -
.' l 1
[ s a 1200
~
* ~
2 w i D= 5$
"O I 1000 '
b m 8
- w #
2& m<
- m 800 8g A 0.02 FT2 -
g' O E kb w e
-o w m -
u - 2 =" 'i g 600 A=0.05 FT E~
! O 9 m w .n e .
El]
~,..
- .>i
- 2 b
m 5,. 400 - -
'" wu A=0.10 FT 2 EE
- Er-
- C5 5t a s.
200 - 2 A=0.35 FT 9, N - g,
; w ?
1 A=0.50 FT2, g g
' 0 . e
~
O 500 1000 "~ 1500 2000 TIME (sec) a
.q i
, a -
7 ;b ,' !
- QN .
gi~ sSo E a _ 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 3 2 1 0 0 0 0 0 0 0 1 1 1 1 9 8 7 6 5 4 0 3 2 0 0 1 O
- g;3 #gc;,gO a.
,~ -
5 0 5 0 5 0 5 0 5 0 5 ; 6 6 5 5 4 4 3 0 3 2 2 1 1 5 O4 _ 0
- 1
~
N O S Y %m _c-qIg1 A R
% t P n o
S P c 3 k i I 0 a 1 e L . m a e t S g 2 T F t n
) 2 o c
e
- 0. it s 0a t I
2 ( oS l 0 t 1 E r M ee I T sw no o p P. s r ea Re l t c nu eN m
" nm ii I
0 a r t g I C I 1 8 nl
' oi
" CP 2
2 5 E R U
. - - - 3 G-I F -
0 - 1 0 0 0 0 0 0 0 0 0 0 0 0 0 0 8 3 6 3 4 3 2 3 0 3 8 2 6 2 4 2 2 2 0 2 8 1 6 1 4 1 2 1 _#[yagEE,
- . i' i !1 ;i' , .
n ;1
- *i . - / tI' : 1 ; ,j . I . e ;-..i'-!
.. ) - - ,
s* NNN g;0 hUnE a-0 0 0 0 0 0 0 0 0 0 0 0 0 3 2 1 0 0 0 0 1 1 1 1 0 9 0 8 0 7 0 6 0 5 0 4 0 3 m 2 0 1 0
- - _ - - . a '*
5 6 6 0 5 5 0 5 5 4 ya1 0 4
% eh50c8 5
3 0 3 a-5 2 0 2 5 1 0 1 5 O 0 1 k - a N e O - L m S YfgI)III A v e a -
- ' t R
P S s
)
S t o m T 2 t F a I 3 0 n T 5o 0.ti 1 0a t t oS r ee sw no 1 oP p 1 t a n sr w o )ea T c cR e P e l stc l l 20 (nu
. eN .
1 Em - Mnm Iii
' .. Tar t o ni
. ol
.' ,. \ CP 3
I l 0 2. --
. l
. 1 5 -
r E _ o t R c U a G I e F r P
- - 0
. . 0 0 0 0 0 0 0 0 0 1
8 6 4 2 0 8 0 0 0 0 0 3 3 3 3 3 6 4 2 0 8 6 4 2 2 2 2 2 1 1 1 _'o[g*&k5 ' S j;l; 1l 1 .
yp;;N'i aJ, J
- 4 . ,- . ,'j ;
E !, t
;;,;: ; P
- 6- _
. 4-Mg, S g gE- gt' n .
0 0 0 0 0 0 0 0 0 0 0 0 3 0 2 0 1 0 0 0 0 0 0 7 0 6 0 5 0 4 0 3 2 0 1 0 1 1 1 1 9 8
. - - _ - - 7 _'*
g;n Tg gSa I 0 5 0 5 0 5 0 5 0 5 0 5 4 3 3 2 2 1 1 5 0 6 6 5 5 4 0 0
- - - - - - - - - - - - 1 k
a t e. m
' a M
~
~
e t O ~ S S Y e6II 2 t A F R 3 n
- P 0 0o I S 1
- 1. ti 0a t oS l t l r a ee w sw no T t n oP o p sr c
P ) ea
. cRe e l stc 2 ( nu eN I
l 0 s 1 E m o Mnm t m I ii T ar a tg T nl oi CP , 4 2 r r 5 t o I 0 E l R I c 1 U a G I m F P
~
- - - - - - - - _ - - 0 0
1 0 0 0 0 0 0 0 0 0 0 0 0 0 0 2 0 8 6 4 2 0 8 6 4 2 8 6 4 2 1 1 1 3 3 '~ 3 3 2 2 2 2 1 o~ g g a5s-f'i< il'
0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 3 2 1 0 0 0 0 0 0 0 1 1 1 1 9 8 7 0 0 0 O 6 5 4 3 2 1
~ - - - - - - - - _
g%Seh5y3a 5' 0 5 0 5 0 5 0 5 0 5 0 6 6 5 5 4 4 3 3 2 2 1 1 5 O4 0
- - - - - - - 1 k
a N e O L 1 m S Y A IV a8 v . a e t R P S l, S 3 I 0 t 1 F n 5 o
- 3. it 0a t
oS j t j r s' , ee o T, sw no m ) oP t a c p T e sr I 2 (s ea R e I 0 l t n 1 E t c e M nu I eN pc T m nm ii ar t g nl oi CP r o t - c 5 a I - e 0 2 I r l 1 p 5 E R _ U _ G _ I F 0 8 0 6 0 4 7- 0 2 0 0 0 8 0 6 0 4 0 2 0 0 0 8 0 0 0 0 0 1 3 3 3 6 4 2 3 3 2 2 2 2 2 1 1 1 1 y~ Egg 5e i . j: ;: ?' ,~
0 0 0 0
^ i~E[uEa 0 0 0 0 0 0 0 0 0 0 0 0 0 3 2 1 0 0 0 0 0 0 0 0 0 0 1 1 1 1 9 8 7 6 5 4 3 2 1 ,0 *
- _ - - _ - - ~
- i y=3ygJj8a l 5 0 5 0 5 0 5 0 5 0 5 6 6 5 5 4 0
4 3 3 2 2 1 1 5 0 "0
- - - - - - 1 k
N a O e S Y A a' D * , L m a R e t P S l S 3 0 2 1 t Fn t o n 5.ti o c 0a P t oS t r ee sw s ) no o - oP c p t e e sr a 2 s ea ( R e l T ! 0 l 1 E tc M nu I eM T m nm ii ar t g nl oi CP 6 I 2 l ll 0 l 1 5 a E y w R U G I r F o t c a - e - p r
. - 0 I
1 0 0 0 0 0 0 0 0 0 0 0 0 0 0 8 6 4 2 0 8 2 0 3 3 3 3 6 4 0 8 6 4 2 - 3 2 2 2 2 2 1 1 1 1 _o ~ $gT5r r
~ o u c, u;; + , :::.u.s - ..es,::- n u :a..-:~a 2 .; -
. . .,n ;n::=c.,m.
l "# 3k ll i?."7 300 - .
]190 _
I e i 5 4 R I, f 50 - e t e l
**' a.so a
SREAK SIZE (FT8} pf I FIGURE 5.2-jf TIME TO REACH 10 PSIG IN DRYWELL , o., FOR VARIOUS SIZE STEAM LEAKS ' PILGRIM NUCLEAR POWER STATION & FINAL SAFETY ANALYSIS REPORT t 0 9 MO e
y :. ,.( ;; . g.;;- ... yqygy 3, g;;p , 3; , , _ 3 ::, . , 7 , M d }1
- PNPS-FSAR higher than the test results. The slightly V overprediction is believed to be due to a combination of:
i No condensation assumed in calculated response, Slight overpredictich of reactor vessel discharge 1 flow rates, and I Incomplete 11guld carryover into the drywell vents. As the chosen size of the vessel orifice is over increases, predicted and thethe vessel depressurization rate This trend is overprediction of drywell pressure f ricreases. illustrated on Figure 14.5-4, where calculated and measured drywell peak pressure are compared. In no case did the model underpredict the test data. 14.5.3.1.2 Containment Response . The calculated pressure and temperature responsesFigure of the4.5-5 containment shows are shown on Figures 14.5-5, 14.5-6, and 14.5-7. that the calculated drywell peak pressure is 45 psig, which After the is well discharge r I below the maximum allowable pressure of 62 psig. of the primary coolant from the reactor vessel into the drywell, the temperature 14.5-7), of the suppression chamber water approaches 130 t Most of the noncondensible. gases. psig, as shown on Figure 14.5-5. chamber during the vessel forced into the suppression are the noncondensibles soon phase. However, depressurization redistribute between the drywell and the suppression chamber.via the Vacuus-8reaker System as the drywell pressure decreased due to steam condensation,. .. The Core Spray System removes decay heat and stored heat from the core, thereby controlling core heatup and spray The core limiting metal water water transports reaction to less than 0.1 percent. through the broken i the core heat out of the reactor vessel This hot water flows rectreulation,1ine Steam flow is negilgible. in the form of hot water.into the s'uppressio l vent pipes. removed from the Primary suppression chamber water is then heat Containment System by the Residual Heat. Removal System (RNRS) exchangers. Prior to activation of the RHRS containment cooling the RHRS pumps mode (Iow (arbitrarily pressure j assumed at 600 see after the accident), coolant injection (LPCI) mode) have been adding . liquid to the reactor vessel along with the corethe spray. After the reactor vessel is jet pump nozzles, the excess flow flooded to the height ofrecirculation line break into the drywell. discharges through the This ' flow, in addition to cooling theafuel, offors considerable depressurization of the
. _s'
- cooling to the drywell and causes f
%/ containment as the steam in the drywell is condensed. At 600 sec, I I
14.5-7 Revision 6 - July 1986 ;
G=wm m; nan ..m. c xc ws , zj a . - .:...
. r .s.,_ .; , . , ;w.
.. e;
% 32 PNPS-FSAR the RHRS ' pumps are assumed to be switched from the LPCI mode to the o containment cooling mode. The containment spray would normally not be activated at all and the change over to the containment cooling b
mode need not be made for several hours. There is considerable
- time {
available to place the Containment Cooling System in operation because about 8 hr will pass before the containment design pressure is reached assuming no containment cooling. To access the primary containment long term response after the accident, and analysis was made of the effects of various containment spray and containment cooling combinations. For all cases, one of the core spray loops is assumed to be in operation. The long term pressure and temperature response of the primary containment was , analyzed for the following containment spray and cooling conditions: ) Case A - Operation of both RHRS cooling loops with two residual heat removal (RHR) pumps and two RHRS heat exchangers in suppression pool cooling mode. No containment spray. Case B - Operation of one RHRS cooling loop with one RHR pump and one RHRS heat exchanger in suppression pool cooling mode. No containment spray. -
~
Case C'- Operation of one RH95 cooling loop with one RHR pump and one RHRS heat exchanger in containment spray mode. The initial pressure response of the containment (the. first 30 sec after break) is the same for each of the conditions. During the long term containment response (after depressurization of the reactor ' vessel is complete) the suppression pool is assumed to be the heat sink in the containment system. The effects of decay energy, stored energy, and . energy. from the metal water reaction on the suppression pool temperature are considered. . Case A This case assumes that both RHRS loops are operating at design heat removal capacity. This includes two RHRS heat exchangers, two RHR$ pumps, and design values of cooling water flow to both RNR loops operating in the suppression pool cooling mode. The RHR$ pumps draw suction from the suppression pool and pump water through the RHRS heat exchangers and back into the suppression pool. This forms a closed cooling loop with the suppression pool. This suppression pool cooling condition is arbitrarily assumed to start at 600 sec after the accident. Prior to this time the RHRS pumps are used to flood the core (LPCI mode). - The containment pressure response to this set of conditions is shown as curve "a" on Figure 14.5-5. The corresponding drywell and suppression pool temperature responses are shown as curves "a" on Figures 14.5-6 and 14.5-7. Af ter the initial rapid depressurization, energy addition due to core decay heat reselts in a gradual pressure and temperature rise in the containment. When the energy removal 14.5-8 Revision 6 - July 1986 A
exas v.- .c- . _ .- Ei.a m .. -. nc_. u .; , 2 &[W.y PNPS-FSAR rate of the RHRS exceeds the energy addition rate from the decay heat, the containment pressure and temperature begin to decrease. Table 14.5-1 summarizes the peak containment pressure following the initial blowdown peak, the peak suppression pool temperature, and a
- summary of the equipment capability assumed in the analysis.
Case B This case assumes that one RHRS loop is operating at design heat removal capacity (one RHR heat exchanger, one RHR pump, and design value of cooling water flow to one RHR loop operating in the suppression pool cooling mode). As in the previous case, there is no containment spray operation and the suppression pool cooling mode is assumed to be activated at 600 see after the accident. The containment pressure response to this set of conditions is shown as curve "b" on Figure 14.5-5. The corresponding drywell and suppression pool temperature responses are shown as curves "b" on Figures 14.5-6 and 14.5-7. A summary of this case is shown on Table 14.5-1, including a sumary of the equipment capabillty assumed in the analysis. Non yen % $* w" Case C , g;w gf .g.),g. afy.p pn, n9 po,ge syn
- This case assumes the same equipmen erability as Case 8 except that & discharge from the RHR eat exchanger is routed g fo de ,
containment spray am of operatierrf It assumed that the containment spray is established at 600 sec after the accident. I The containment response to this set of conditions is shown as curve "c" on Figure 14.5-5. The corresponding drywell and suppression. pool temperatures are shown as curves e on Figures 14.5-6 and 14.5-7. A sumary of .this case is assumed shown on g the equipment capability in Table 14.5-1, including a summary ofp fhc.ANg the analysis. eu.h=%eV. The Comparing the " containment spray" Case C with the "no spray" Casa B,"M*l flow
! it is seen that the suppression pool temperature response i the came_ ,,P g e, because the same amount of energy is removed from the poo ..@ M ::: T a g th eg l ... .m ~ - ny- m.o .. ~ ---, u . tA 1 v.
k" "' '5; 5 5.h ! ' :: :: 3.fHowever, the post N4^M8" 8- M k' '5
! blowdown containment pressure is higher for the "no spray" case, as shown by Figure 14.5-5. This, however, is of no consequence since 4 he. Mme.E the pressure is still much less than the containment design pressure g* gg of 56 psig. Figure 14.5-8 111ustrates the slight effect on calculated containment leakage rate, due to the higher pressure.
14.5.3.1.3 Core Standby Cooling System Pump Net Positive Suction Head To assure proper operation of the RHRS circulating and reactor CSCS pumps following a design basis LOCA, precautions are taken to ensure that a not positive suction head (NPSH) margin is available to all above pumps at all times. V 14.5-9 Revision 6 - July 1986
,. , , - 4, , ,. . .. - : - - " " - "'
\
L7 pJ2. ! 't Note: It is recommended that curve e in Figures 14.5-5 through 14.5-8 be left unchanged. The reason is that, in the time range of6 application of the drywell spray (from 600 seconds to 10 seconds), it is expected that the drywell spray flow rate of 300 gpm and 5000 ppm will produce similar containment response. The variable that will have the largest impact is the containment pressure, but even there the impact is expected to be limited to a sh' ort period following the initiation of the drywell spray. L _ _ - - - ' - " - - - - - - ^ - - ~ ^' ' _ , ._ ---
DXdXZW= - ., ~Z . .z . : . ;. -;~ . .. : - . . .. , N f.N PHPS-FSAR Table 14.5-1 I '1 LOSS OF COOLANT ACCIDENT PRIMARY CONTAINMENT . RESPON. Case A _ Case B -- Case C _ Secondary Peak Pressure, psig None8 13.1 Peak Pool Temperature 'F 8.0 142 166 Mode of RHR5 operation 1E6 Suppression Suppression Containment Pool Cooling Pool Cooling Pool Cooling Number of RHR loops operating 2 1 1 Humber of RHR pumps operating 2 1 ) Number of RHR heat exchangers operating 2 1 1 Tota.1 RHR5 flowrate, gal / min 10,000 5,000 5,000 Core Spray System flowrate gal / min 3.600 3,600 3,600 Containment Spray System flowrate, gal / min 0 0 RHR5 heat exchanger flow- 4,COS 3 0
- l rate, gal / min 10.00 5,000 5,000 RHR5 heat transfer rate when suppression pool temperature = 165'F Stu/hr 128 x 10' 64 x 10' 64 x 10' C'
v? Number of RBCCW loops operating 2 1 1 Number of RBCCW pumps operating 4 2 2 Number of RBCCW heat exchangers operating 2 1 1 Total RBCCHS flowrite to RHRS, gal / min 5.400 2,700 2,700 Number of SSW loops operating
- 2 1 1 Number of SSW pumps operating 4 2 2 Total SSW flowrate to RBCCHS 10,000 5.000 5,000 SSH Inlet water tempera-ture, 'F 65 65 65 bressuresteadilydecreasesaftercontainmentcoolingisestablished.
.l g O u 4 0 4 * ** A i ' ' 9 , met
- - _ . _m., ,
.,,.7_--- ' _, ,,
- 7,
> T d:.'. "ow
- ne' .U !. l . -e s '%.* ,' \ ;*~4*<.**.4
% ='t
< s' d
O
. M132. 7 a
8 1 m .. .. gg. . . . .. .... [ . . . n . . u .. . .. . . , ; , . . . . . . . , , , 7_g g 4m
- s bl !
6 l=__ F_
=.
=
i p__-__
-- _ ,p y y4 ,;
y._:
,p wp _n = _= _ =_ =_
- j. . ..
j,. 33333 a% J l
.,_,,. ll, {.. ._ ,,,._ ,,,._,.,.,
B.e.s.e.t. '"~, a_
- 14. ),,
l___--_- __-_ __ . _____ j5'_4 5lst , ll s-BESSI 8- e .g' _., ,, 1 .'
- _
f
.)_____ _., .,,,.
passa
.A_u A g a? , . .
, 5 __,._,. ____,.
l i gr_ y
,! ;* 1 ,
_[-. p -2 .. i _p, ,_, . ___ partsg 7,J-.. w. . .. .,
.nsa
' ll3Hj fl l
l i
-n..., a_t . . . .._ .. , h.. . . : Si l_ A,j f,,,g,.
i
- l.
- __,_;,,).
3 n i
, i. {,,,iis p,a. ,
. l i
.. _i. y s
j,,_. e i
,, ,, . a , ,._,,
, c , cuaus ;;)
. , n ni i
r._., ,,..... . .. .: p..a is . .. ,..
- g. , ,,.,, ., ,, ., ...
r ., .._.. .
.j'
- th: Inx) I =
a....., x l'_.,_.. I o t:: fl n =_ f(, ::: ll :-:- h,!.*. ll : *~ 1,,_.,.f L._4,, -
.._a.. -
_n. , _g i se... sul f O l f ,_,, l ,.,, BSStII
; .en - ,gg,
.e 2,.
- g
] '..
yl ' j
) ',,,_,,
i
,, , ,,, .g I E:: ! E33: EE: l EE l iR: M"l -
/
l}!!S I N ll!!i! l!l!llliiii:l'RdlE! ll!N ll!Il .Nii }!!)!.!!!!! d$Ek
~
~
- a%i ), -I ,
.i . &3 ,- g - k, qD. _
%gqq . ma 8' ,I \. - QpSM I _ ,
8 r M. , it w
;Lwi, i
.a i., i
* &.l i t
(k; ia J givn-eshe; a
.;E
,,m.s.y lf;[. [.[((
g v .i 8 s 3 l i f <. *.*, .'/ , E ((l gl . 77, ,. l M . .
'. y l
e 1+ s
%("i i
l It: $
>l g:
i
.'i -
. i .
i1 i . i i . i . i .
,g
. i . i
- ?
a J ":.*
~%
s& W ,
- 43
PThi;. n .! - *iW N2dW.W:EL.:.w: :n , i G. 2 O . . n '- a
. :h...; .' % D mi : '
.s .
- MED Proposed Change Safety Evaluation No.: cDI47 1AFETY EVAulATION **Y I O N PILGRIM NUCLEAR PNER STATION Rev. No. D .
PDC PCM System Calc. Initiator: Dent: Group: No.: Name: No.: Date: T.'M.Wqcke neb FS&Mc %-52.g Fire,Wder Tie.- ah f o AHR Description of Proposed change, test or experingnt: F+e. Wa 4er-
~r, *o . n 4e G '+ Souhse % 4ev % - in +c wQ SAFETY EVALUATIM C/ACLUSIMS:
The proposed chane4, test or experiment:
- 1. C4 Does Qt ( ) Does increase the probability of occurrence or consequer.ces of an accident or malfunction of equipment luportant to '
safety previously evaluated in the FSAR.
- 2. 04 Does Not ( ) Dois create the possibility for accident or malfunction of a different type than any evaluated previously in the FSAR.
~
- 3. P9 Does Not ( ) Does reduce the margin of safety as defined in the basis for any technical specification.
RASIS FOR SAFETY EVAt tl4 TIM emcLUSIGtS: ( Go a%ula.d ba.5/s.) s 9 Change Change (>Q Recommended ( ) Not Recommended SE Performed by b M . d Date If 87 Exhibit 3.07-A Sheet 1 of 3 w-ISSUED l'OR CONSTFM.UION - l 3.07-13 Rev. 4 l l 1 l l f
<, ;+ n.;; . a y: : i:w' .: ia m;;c . . . . . . w. . x. w :: ; h w ;:u. . . .: : u :- a:- ~-- G rzv r ' Safety Evaluation - No.: 214] SAFETY EVALUATION c PILGRIM NUCLEAR POWER STATION b E 'N Rev. No. O l A. APPROVAL (g) This proposed change does not involve a change in the Technical l Specifications. l C() This proposed change, test or experiment does ( ) does not OQ involve an unreviewed safety question as defined in 10CFR, Part 50.59(a)(2).
@ This proposed change involves a change to the FSAR per 10CFR 50.71(a) and is reportable under 10CFR50.59(b).
( ) Comments: The safety evaluation basis and conclusion is: ,
@O Approved () Not Approved un J nlek7 Discipline Group Leader /Date Supporting Discipline Group Leader /Date B. REVIEW APPROVAL
( d Comments: A S M W M *
- M M L 84M4 ac c.W.J h ud % %tc " , W W h
sJw 1LJ :u M Mt - =_- k \ bc e _.'fA . NwL 6ff'? 49-gy G@p Leade'r/Date C. ORC REVIEN . () This proposed change involves an unreviewed safety question and
. a request for authorization of this change must be filed with the Directorate of Licensing NRC prior to tuplementation.
(.dThis proposed change does not involve an unreviewed safety question. ORC Chairman , Date [!M!O ORC Meeting Number
, 99 88-CC:
Exhibit 3.07-A Sheet 2 of 3 ! l 2__: = .- - - -- n n < i3sn e ...u o ro, x l 3.mtCONSip[CHON- 4
- ca nmue :.;w.w Mwa;.n ; J.:.u.u.L &.x.s ;a.a. Dun m:rm =.m ac=wr '
! ~.c . ,
Safety Evaluation No. 2/V 7 PDC 86-52B, Rev. O Sheet 3 of M. A. DESCRIPTION OF PROPOSED CHANGE The proposed modification is'as follows: A.1 Installation of a piping cross-tie between the Fire
. Protection System (FPS) and the Residual Heat Removal' System (RHRS). The cross-tie will allow fire pumps P135, P140, and the new SEP diesel fire pump-P-179, when installed, to supply fire water to the RHRS Salt Service Water injection line in the Reactor Auxiliary Bay. The cross-tie originates from' an 8" Fire Protection System header above floor elevation 23'-0"
- and terminates at the 12" Salt Service Water System injection line below floor elevation 23'-0". The-l cross-tie consists of approximately 90 feet of 8" pipe, j fittings, pipe supports, two gate valves, one. check
! valve, and a removable section of pipe. One gate valve is for FPS isolation, the second gate valve. and check valve are for RHRS isolation. The RHRS boundary begins at the inlet of the check valve. The gate valves are normally. locked closed. The ; removable pipe section is one foot in length and 1s-located between the isolation valves.-It is installed and removed with quick disconnect victaulic couplings. The FPS isolation gate valve will have a 3" bypass line around it. The bypass line consists of approximately 9 feet of 3" pipe supports, a globe pipe, fittings, valve, and an in-line flow meter with a local indicator. The bypass will be used when the water quantity required to remove decay heat has decreased to approximately 300 gpm. The bypass line globe valve will be normally locked closed. l A 1" drain line with two open gate valves will drain off any leakage past the RHRS cross-tie gate valve to a rad waste drain sump. A 3/4" vent line with a closed globe valve will be placed on the fire protection piping at a cross-tie high point. A drain will also be placed on a low point of the fire protection cross-tie to provide drainage upstream of the check valve. The cross-tie will have approximately A . feet of_. piping, ~ a manual gate valve s..e ctrecK-~vENE ! .wficT1 ce ASME Section III, Class 2 whicd QEnnEcts)tdp(TPSito t ie Salt
I!T6e.rama gg of the ServiceWaterinjectionifnY"hqf.,NFPA,.std cross-tieisdesignedl({({ psf rus, h _ _ _ . . _ - _ . . _ . . . . . . _ ___
- - : s. . e iw. . . , . v .-: ~ .a. s; a : :....~e..
. a , :.c i. 1. a. , . an zw :- T - :.: \ -
k' Safety Evaluation No.J/y] PDC 86-52B, Rev. O Sheet 4 of 14 All ASME Section III, Class 2 piping and valves will be seismically designed and supported. The ANSI /NFPA piping connected to the ASME piping Will be seismically designed and supported to a seismic boundary anchor at floor elevation 23'-0". The remainder of the ANSI /NFPA piping above floor elevation 23'-0" will be supported 5- ::iri- for seism /c 2T/,r cyNepig . ThH r)sglr7 B. PURPOSE OF THE CHANGE The purpose of this modification is to provide a non-electrical AC power dependent redundant water source for containment spray or vessel injection for extended station blackout and severe accident scenarios beyond design basis. C. SYSTEMS, SUBSYSTEMS, COMPONENTS AFFECTED This modification affects the Salt Service Water System in the following manner: o The 12" salt service water line to RHRS is affected by this modification as the 8" cross-tie line with a manual gate valve and check valve is connected to it. This modification affects the Reactor Building Area 1 in the following manner: o Civil structure of the Reactor Building Area 1 is affected by this modification due to the installation of the new 4"-GBB-10 pipes including supports, valves fittings, tubing conduit, and associated small piping connecting the 18"-GB-10 header to the RPV. This modification affects the Reactor Building Auxiliary Bay Area 2 in the following manner: o Civil structure of the Reactor Building Area 2 is affected by this modification due to the installation of the new 8" pipe including supports, valves and fittings, connecting the Fire Protection System and the Salt Service Water System. This modification af f ects += " ice-Pretenti-dn^5pted' in the following manner: g ..
.q aqiwT.-l g- '; {.' @
o The 8" fire protecti ] the QNthe.Weie o xiliary Bay is af fected by trii , diftdaM on._; Thy e- ss-tie 1 0' J
G-:: u . .G C.-- ~.- - M .v:Ymr:=~ 2 :u . , ' r ,.W: : . :, ~ .:;~12;. T :.c %::;:- : : , .. w: , , - s Safety Evaluation No. 21Y7 PDC 86-52B, Rev. O Sheet 5' of / f with a manual gate valve and a 3" bypass line with a globe valve is connected to the fire protection line. D. SAFETY FUNCTIONS OF AFFECTED SYSTEMS / COMPONENTS Salt Service Water System The safety function of the Salt Service Water (SSW) system is to provide a heat sink for the Reactor Building Closed Cooling Water (RBCCW) System under transient and accident conditions (Ref. FSAR Section 10.7.1). Reactor Building Area 1 The safety function of the Reactor Building is to limit the release to the environs of radioactive materials so that offsite doses from a postulated DBA will be below the guidelines values of 10CFR100. In addition, the Reactor Building is designed to provide protection for the engineered safeguards and nuclear safety systems located in the building from all postulated environmental events including tornadoes (Ref. FSAR Sections 5.1.3, 5.3.1). Reactor Building Auxili'ary Bay Area 2 This structure has a safety function of protecting the integrity of safety related components and equipment housed in it (Ref. FSAR Section 12.2.1.1). Fire Protection System The Fire Protection System does not perform any safety related function. The Fire Protection System is important to safety because it mitigates the occurance of equipment damage from a fire that could damage or significantly degrade the capability of one safe shutdown system in the plant. E. EFFECT ON GAFETY FUNCTIONS Salt Service Water System o The new 8" cross-tie line, the normany,.. closed-gat $ and globe valves, the check EnT *the associ~ated drain line form a part of theg EsYtIserjice Watbr sf tem pressure boundary and re brd y$ ddd ..tro i g..'i s g hp its design basis integrity.
-W~
ty
, 7. - - ,,
,, , - ,, , m , m , ,; ,, , , . . , , , , , ,
4 Safety Evaluation No. ElV7 PDC 86-52B, Rev. O Sheet 4 of M. Reactor Building Area 1 o The pipe support installation has the potential of affecting secondary containment integrity .due to the drilling of concrete for pipe support anchor bolts and due to piping loads transmitted to the concrete. , Reactor Building Auxiliary Area 2
~
o Installation of non-safety related components (connecting the- Fire Protection System to the Salt Service Water System) and the associated pipe supports have the potential of affecting the integrity of the safety related system housed in this area. The safety related systems can be affected by the non-safety related components during a seismic event if the non-safety related components suffer a structural failure. Fire Protection System The Fire Protection System does not perform any safety related function. However, the new 8" cross-tie line, the normally closed gate and globe valves and the associated vent and drain lines' form a part of the Fire Protection System pressure boundary and are required to maintain that boundary as any other Fire Protection System component. The , cross-tie between Salt Service Water and the Fire Protection System has the potential of contaminating the Fire Protection System with radioactive reactor water. F. ANALYSIS OF EFFECT ON SAFETY FUNCTIONS , i Salt Service Water System (SSWS) I o The new 8" cross-tie line, the normally closed manual gate valve and the check valve and the drain valves are i designed to ASME Section III Class 2. They are seismically designed and supported. They are subject ' to Class 2 ISI. The design ensures that .the modifications meet the design basis criteria of the system. Hence, the modifications maintain the -
- j. integrity of the SSWS and RHRS pressure bound._gries and the safety function of SSW _.iaEc:not'-affected this modification. {} p A CONSh._. s v'egoN l
. u:r. . . .% ;h. :.ii w ;; w s.m.m :.M % 2;>L ; ; .. . w . L x& c m a .. .a %..F m-~=;G-av Y
{ . Safety Evaluation No. 2JY7 PDC 86-52B, Rev. O Sheet 7 of p. Reactor Building Area 1
,,ff h jo g M hid 7 dedHb5, j o concrete drilling for pipe support anchor bolts will be in accordance with the applicable specifications and PNPS procedures This. ensures that the integrity of the structure is not degraded. Hence, this modification does not affect the safety function of the Reactor Building.
Reactor Building Auxiliary Bay Area 2 o The non-safety related components which .have the potential of degrading the integrity of the safety related systems houses in this. area are adequately supported by pipe supports which are analyzed for seismic forces of the saf a shutdownA' earthquake. These pipe supports are quality 'p'r. Hence, this
- modification does not degrade. the integrity of any safety related system in the Reactor Building Auxiliary Bay Area 2.. -
Tnin g[,,/,7 Fire Protection System o In order to ensure the Fire Protection System pressure boundary on the new 8" cross-tie line, the normally closed gate and globe valves and the associated vent and drain line are designated FP-Q. o In order to ensure the Fire Protection System is not contaminated with radioactive reactor water during normal operation, a removable spool piece is located in the B" cross-tie online. This spool piece is removed during normal operation. It will only be placed is in the being 8" cross-tie line when the fire water system used as- a water source during extended station blackouts or severe accident scenarios beyond design basis. G.
SUMMARY
This modification does not call for the safety equipment of the system to operate at higher pressures, temperature and more severe conditions than the existing levels, hence, the l modification does not increase the probability of malfunction of the equipment important to safety. ~% _ ,,__.: w --- - l ISSU E 9 ~'M. iv..r. L CONSTy. - . - - -
.\ON
-t__ _
l
( uw.. . . :a ;<.u n . . .> . ... . w w..e t .=i a..,-nu ., u.a.v . . = . + ::.: . = :a .- := --i=r m m r - Safety Evaluation No. Jf'[2 4 PDC 86-52B, Rev. O Sheetg of jv This modification is designed such that the failure of any noe-safety related equipment will not degrade the capability of safety related equipment to perform its function. This modification does not require a Technical Specification Change. This modification- does not involve an unreviewed safety question.
~ - '
*-*W*. .a L a..' g b v 'v. ..,*
u .. - A L ,? ( *\* CONSL.;.! TlON
, = . , = w . c .- .i . . w.- n. a. . -
... :. n . - ..=- .w., . .- . . - -
Safety Evaluation No.: al 4'7 SAFETY EVALUATION WORK SHEET Rev. No. O l A. System Structure Component Failure and Consequence Analyses. She+ 9 of I+
. Systen Structure Camoonent Failure Modes Effects of Failure Condents !
- 1. $<.4 Se er See Asc h eel dee4s l Nfe $v s fem #
_ l
- 2. m fbb.6 see A+4,ck,cl Shech Sy,[em _
3.
)
General Reference Material Review FSAR CALQJI.ATIONS REGULATORY SECTION PNPS TECHNICAL SPECS. DESIGN SPECS PROCEDURES GLIIDES STANDARDS CDDES
~ 49 3,5.A S e.c . M ' loo lo cM So no ' R
- r. . o.R-> h e c. , M *3 0 f y; . uc.7-t 2, l2. 8 A AIS C A tl. fo l- k i Fklot*l ,_
h a e_ . M-303 Aust n3t. I - W90 ApMr A Sose, M - Go o MME Ser J.A 5 WEO Arge C Sec . M-Goa IP *C. M - 50 + B. For the proposed hardware change, identify the failure endes that are likely for the components consistent with FSAR assumptions. For each failure mode, show the consequences to the system, structures or related components. Especially show how the failure (s) affects the assigned safety basis (FSAR Text for each system) or plant safety functions FSAR Chapter 14 and Appendix G). Prepared by Y A M . M d Date 5 ! 97 h0TE: It is a requirement to include this work sheet with the Safety Evaluation, t Exhibit 3.07-C 3.07-18
- ~ ~
.~~,IIev. 4 1.3 5L. o 6 uA CONST: ~;;;\oN
=
y 7, g. , - .
,.,y,,, ,
e SafetyEvaluationNo.QlV7 PDC 86-52B, Rev. O Sheetgo of l4 SAFETY EVALUATION WORK SHEET A. System / Structure / Component Failure and Consequence Analyses The following apply only to the proposed modification per PDC 86-52B
- 1. System / Structure / Component: Salt Service Water System Field Mode: Structural failure of the 8" cross-tie piping interface with the Salt Service Water System.
Effects of Failure: Loss or degradation for the
~
capability of the Residual Heat Removal System to perform its function during a design basis accident. Comments: The 8" cross-tie piping interface with the Salt Service Water System are quality category "Q". The 8" cross-tie is seismically , designed and supported to the first anchor l past the isolation valves.
- 2. System / structure / Component: Fire Protection System Failure Mode: Structural Failure of the 8" cross-tie Fire piping interface with the Protection System.
Effects of Failure: Degradation of the capability of the Fire Protection System to perform its function. Comments: The 8" cross-tie piping interface with the Fire Protection System is designated FP-Q.
- :.u ,
L'.qw Cw .! r. :l. ~ 'yy ; 3 [ w. , s r
.f v., .n.c+.;~ . . .. _ .I O N
... .- . . . , , . . < . .w , c . - . - - - - - = . .
PILGRIM STATION FSAR REVIEW SHEET NCehIIoFI4-
References:
Safety Evaluation: 0147 Rev. No.: O Date: . Support. a change for *defe'cdsoe s to Yb e. fike hofed, o g' g h J s, l* 3 erne.s Wrier Sy s +c m . List FSAR test, diagrasis, and indices affected by this change and corresponding FSAR revision. Affected FSAR Revision to affected FSAR Section is thown on: Section Preliminary Final 4 J.8 Attachment 1
- 10. 8 Attachment % 1 Q.4.E-! Attachment 3 Feh.o.7-l At;.c . ;. : M-Ela ,
-fu Sfis e7 ,
Fx .Io.1r-l . w- =- s- m - ? ls sh+.1 \ Attachment 6 PRELIMINARY FSAR REVISION (to be completed at time of Safety Evaluation preparation). Prepared by: N b /Date: ltr H, WJ /Date:
, /9,h,7 87 Reviewed by: g 0
Approved by:- 'D// /Date:((k4'l FINAL FSAR REVISION (Prepared following operational turnover of related systems structures of components for use at PNPS). Prepared by: /Date: Reviewed by: /Date: Attach completed FSAR Change Request Form (Refer to NOP). Exhibit 3.07-A Sheet 3 of 3
- - - - __ .w. ., .
ISSUE.D FOR 3oPCONSTlJ. I ~TIOft ' . . . . -
W Attachment f l Safety Evaluation No. Al4 7 PDC 86-528, Rev. 0 Sheet X of X l t. 14 RECOMMENDED FSAR CHANGES Sections 4.8 and 10.8 of the FSAR need updating due to this modification (PDC 86-52B). FSAR pages 4.8-3 and 10.8-1 have been marked with suggested updates and are included in this Attachment for your review. The following drawings will be revised as part of this PDC, but are not included here. Dwa f.D. FSAR Ficure ,Iit,1g' M-212 10.7-1 P & ID Salt Service Water System
"*E F ; I; Re,;Ja.,,geeg q,4 3 fis; r7
" 241 She d 1-
"= rial-Sy:t w M-218 Sheet 1 10.8-1 P & ID Fire Protection System e
.. .,... em B:.::c.c..s U C _,'. LVA
,. CONSTRUCTION
,, 7 ;, , ,
Attachment ?! Safety Evaluation No.y PNPS-TSAR PDC 86-52B, Rev. O Sheetfof,2 o IS 84-
- Standards of the Hydraulic Institute. The shell side ASME of the heatI Boiler and ' *g I E *"*
exchangers are designed in accordance with the I $"I $ Pressure vessel Code, Section III, Class C vessels, and the tube side
. is designed in accordance with Section VIII. The provisions of the "E5*
Winter Addenda of '.966, paragraph N2113 apply. E ," tE8-Power for four RHRS pumps (two pumps on each of two 1 oops) is 3 , ej normally provided through the two independent 4,160 V emergency g pv e buses. In the event that the normal auxiliary power supply is not cu 5 E .T { available, the 4,160 V buses serving the two loops of the RHRS are g #U powered separately from the two diesel generators. ggg % - Additional flexibility has been provided in the design by the Ig5m
*DE addition of a permanent piping connection from the station salt water
- service pumps to the RHR Piping System. This connection is sized to
- fhE(
provide 5,000 gal / min at 0 psig reactor pressure. All piping and , g3 f equipment in the Service Water System which serves Class I equipment goy, end this additional piping connection are designed to Class I seismic myyo requirements. g3 g 2 2 2 2. The interconnection of the service water and RHRS is manually Oo*M c' c. initiated. Inadvertent admission of salt water to the RHRS is e E $ .c prevented by requiring the operator to switch over a spectacle 3'o8 { flange, and to open two locked-closed valves. I.eaks from either system can be detected by periodic inspections of locked-open drains { $ ",i
\ en each side of the spectacle flangeg c gy 4.8.5.2 Shutdown Cooling Subsystem
- Ae eu s f
The SDC Subsystem is an integral part of the RHR$ and is placed in 13DE ' The inital phase of G I 5 e l f operation nuclear system cooldown is during a normal shutdown and cooldown. accomplished by = dumping g ] E steam
- i from the
. - c en
- reactor vessel to the main condenser with the main condenser acting as the heak sink. When nuclear system pressure has decreased to 50 E "o l
- l
,R psig the steam supply pressure is no longer sufficient to maintain M- El -
- vacuum in the main condenser, and the RHRS is placed in the SDC mode ~E*ECC0"!
!8 of operation to complete cooldown of the nucisar system. The SDC @, subsystem is capable oT reducing reactor water temperatures to 125'T % [# { ' approximately 20 hr after reactor shutdown so that the reactor can be 6ao l8" refueled and serviced. In the SDC mode, reactor coolant is pumped by
'OSY
- [
l the RHRS main system pumps from one recirculation loop through the 'Y { "3 RHRS heat exchanger (s) where heat is transferred to the Reactor {vev .,#c m 2 Building Closed Cooling Water System. Reactor coolant is then ) e fM vessel through connections to the'E 2 3 e returned to the reactor recirculation loop (s). Temperature control is provided by varying # " z 3 fi i" flow rate in the RHRS and by use of the RHR heat exchanger bypass .3 Y DE"UJ 3F line. 2"3uY 3 3@ l' PartoftheRHRflowcanbe.divertedtoaspraynozzleinthereactorENa*{ vessel head (see Figure 4.8-1). This spray maintains s'aturate d o en 5 Y L lT' pressure and temperature conditions in the reactor vessel head volume 5 3 "U walls, LS8vE 3 { 8'" 3 lI by condensing steam being generated by the het reactor vesselThis* permits " C 8 ' the,,r3. actor and internals during reactor cooldown. vessel to be flooded more rapidly -'--Q~ f {[7 gj IS S u; c u M r - "e [i td NSTEqON - C - i j
- ww,.a a a ?wx . n '
- x2
- ::e , P ' ' '; . . !:.wim%n% :.ws szm hauc=~---
~
Mttacnment x : Safety Evaluation No g PDC 86-528. Rev. O PNPS-FSAR Sheet / Of / wT . W I4 0-0" t 10.8 FIR $ PROTECTION SYSTEM g}S g A3EC la.8.1 Power Generation objective 4
,og, o a. e e The power generation objective of the Fire Protection System is t UN*e provide adequate fire protection capability in 'all areas of thejy*,(
station. g*e-
. a. c s. 4 eEE 10.8.2 Power Generation Design Sasis Ih e >,0 The Fire Protection System is designed to furnish water, halon, Eg carbon dioxide, and/or dry chemicals as necessary for fire 'I fde extinguishment in the station. The Fire Protection System is eo5 d2 signed to provide the following l'EE*
-*8 A reliable supply of fresh water for fire fighting EE3
- 1. g%"$
o
- 2. A reliable system for delivery of water to potential fire INej I
locations 2 c % a.
'S E o i 3. Automatic fire detection in selected areas $T A c34'
- 4. Fire extinguishment by fixed equipment acti;ated either SEEl ""##
automatically or manually for areas with a high fire risk
$EDo l
- 5. Manually operated fire extinguishing equipment for use by, EC Eu l operating personnel at selected points throughout station E ., '
the'"y*y% l I 7. g o 0 10.8.3 Description ;nt"m i e " s* 3 The Fire Protection System, Piping and Instrumentation Diagram is ' . 5 % j
- f shown on. Figure 10.8-1, sheets 1 and 2. E%@u FE, M**C -
"' 10.8.3.1 Fire Water System *$Ek 5; E5 8
lC\The carbon site fire water supply is taken from two 250,000 gal, lined , h *
- g steel water tanks which are devoted exclusively to fire 2;- T.o
(
- $l protection. The Fire Water Systemmayalsousewaterfromacity',"3g5 I
hjj T g - en ! a. water main. .. e Thewatersupplyisdeliveredbyeitheranelectricmotor-drivenpump;lvt8.n
\ 225
.E;
' (rated at 2,000 gal / min) or a diesel engine driven pump (rated 5 g " a
$1 ot 2,000 gal / min). The diesel engine driven pump is used for standby E 8 b;- and emergency use on loss of ac power. A small jockey pump (rated at w u @*C gL"E 50 gal / min) is provided to maintain a constant pressure for the water 3 r * , J i system. If the system pressure drops substantially, the motor-driven *;c y y ? 'E, fire pump will start automatically,, and if pressure continues to 3 " 3.g drop, the diesel-driven pump will also start automatically. f, ogg , ga
'5-z2d !
>E The pumps feed outdoor fire hydrants, interior hose stations.,,, g S e e b" Sprinkler Systems, and Deluge Systems fw the -slati;onr=
.r- ,
*" - u ;
, -l - a f*oo xcas ,/
yVWJ k
%-__ a j J' gm-, o f CO;nS i - .T5y"~'g '
10.8-1
-- x -~~ -- =
e n.- - . cc w :n. t w. a. : . a L. . :i::n. a; ... m >:. a .: - : - - - x . ~: w - me.: .)-/8' h aserty rve m ew Pf LfRf W F3* LED Whr r fittTfmSgEET EEM Nl oF IN13
- ter. Be. .. O.
ptC PCN System Calc. - i Initiator: Deat: Groups he.: anae: me.: Ost.e:
'I.M. Hauske ND Fs&M:: 86-525 Residual -
- Rest **
Removal Description Of Pr9 posed Sh& nee, test of etportment: FRN 86-523- si Recennect and rareute the & Reseter Head Strav Line. b) Install a stratter it_tne_ KnAritreva:er cross-tie. 1 c) Reute crain frc: 5 " GB 5 / GB t o R): 51er. Seconcary Larrfy EVALLANCP h*LtMIf>t g Containment S uttp. The proposed change test er experiment:
- 1. (4 Does Est ( ) Does increase the prabahtlity of a:curren'se er coniquences of an accident er salfunetten of equipment iMant to safety previously evaluated la the FEAR. '
- 3. (4 Does Not ( ) Dosis create the possibility for at:1 dent er en1fon: tion of a different type than av evaluated prevlously in the FSAR.
30 (x) Does not ( ) Deen reduce the for any tethatsal specifiaatten.. margin of safety as defined in the basis Batte eM RAFFTY eVAtt1&Tf5 fBlatLtffart? fas Attached Sheets ' chanse . mines (4 Recommended ( ) Est Recommended a perf. med er M L % ) M nate 7/h/e7 Erhtbtt 3.57-4 Sheet 1 ef 3 l l 3.07 13 Rev. 4
a ' a:z.v .: :n ~ u. . . . ~ m r xxw . w w ,z z w w , :m ';;w . c. . x . .. .. .w
~'
, EAFETY EVALUATION PILCRIN NUCLEAR PCNER STATION Rev. No. O FRU %- 52 %
A. APPROVAL SHEEf 2 OFM -
* ()() This proposed change does not involve a change in the Technical Specifications.
CX) This proposed change, test or experteent does ( ) don not C4 involve an unreviewed safety question as defined in 10CFR. Part k)50.59(a)(2). This proposed change involves a change to the FSAR per 10CFR 50.7)(e) and is reportable under 10CFR50.59(b). ( ) Comments: The safety evaluation basis and conclusion is: , () Not Approved [ (O Approved Dkh) 1/t$lQ Discipline Group Leader /Date Supporting Discipline Group Leader /Date
- s. REVIEN APPROVAL
. (dComments: y4-Scr o s/ra h s " - If W - b eA .v e in A hn D . s e.ti SM b-> f M .
wh 7/s/n -1 () SESA goup Leader /Date C. CRC REVIEN . ( ) This proposed change l'avolves an unreviewed safety question and
. a request for authorization of this change must be filed with the Directorate of Licensing, NRC prior to implementation.
( ) This proposed change does not involve an unreviewed safety question. ORC Chatruan Date ORC Neeting Number cc: Exhibit 3.07-A Sheet 2 of 3 3.07-14 Rev. 4
c:= w;;. .w ;s mx,a x ;w .;.:. . w .w...1. . :.x. .; u : . r., :.x. ; ...w= a,., .r . n.n ,,.w - ya 14 *s7 09:45 t617) 4I3-7027 P.2 Safety Evaluation No. El3 8 Rev. 0 . FRN 86-52B-4 Sheec ,3__ of 23 A. DESCRIPTION OF PROPOSED CHANGE, TEST, OR EXPERIMENT The proposed modifications are as follows:
- 1. Reconnection of the 4" RHR RPV head spray piping. The original design for PNPS included a RPV head spray line originating from 18"-GB-10 on "A" terminating in a spray nozzle in the top RER loop and the The head spray line consisted of piping, supports, of the RPV.
flow element, a flow control valve, a check valve, twoa containment isolation valves, a test connection, electrical power and control logic. Due head tospray water hammer line, experienced during operation of the it was disconnected header and the RPV and capped. from the RHR All valves and wiring were left intact and power isolated. To reduce the potential 4"-GBB-10 piping which reconnects for water hammer, the new = the existing pipin has been rerouted (approximately 150 ft of new piping)g In addition, . ! control valve TV-1001-58, FE-1001-85 and flow element trans'mitter close as possible to the 18"-GB-10 line. FT-1001-86 are relocated as l 1"-GBB-10 bypass Also, a line with restriction orifices is added around TV-1001-58. This will allow a maximum . . 75 gpm through the head spray line until it of pressurized. Control valve is opened to FV-1001-58 will then be , control flow to the RPV. The SQRT extractor i and flow indication centro 11er will be powered from 120 VAC vital bust Y2 which is backed up by DC power. The SB0 is required Battery / Inverter System (future modification) to provide power to RHR headspray components during a station blackout. Valve TV-1001-58 will be controlled, using nitrogen gas from the backup nitrogen supply installed by PDC 86-53, and connected to the flow control valve in this PDC. The 1"-GBB-10 bypass line is approximately 15 feet long ' with three restriction orifices. The piping includes a gate valve of the bypass upstream line fromofthe the orifices to allow isolation RHR system supply header. t The bypass line will also be equipped with a drain. 87104.1 j l
w w a :.. x z u w :a em :.w .:n m .v . ;c==.. . m .n. ~ .. ;r w ,.~...v . c. n - . - . Sgt 14 E7 09:47 (6175 4E3-7027" '
\
k Safety Evaluation No. 3882- l Rev. 0 FRN 86-52B-E
- Sheet 4- of 23 l a
- 2. The motor and actuator accessories for the RHR head j spray inboard containment isolation valve (MOV-1001-63) 1 will be replaced with qualified components. The replacement motor and actuator will be identical in speed, torque, voltage rating, and operating time to the existing one. Limitorque will supply the documentation to demonstrate the qualification of motor and actuator parts to the Pilgrim containment conditions and requirements Station of 10CFR50.49.
primary The existing motor housing will be replaced with one - that has provisions for a T-drain. The existing cable from the drywell penetration to MOV-1001-63 will be replaced with new cable. The existing cable has been exposed to high radiation and is degraded. and actuator accessories The motor valve MOV-1001-60 have already for RHR head spray outboard been replaced with qualified components. t 3. During normal plant operation, RER head spray isolation valves MO-1001-60 and MO-1001-63 position and remain in the closed if opened, will close automatically when PCIS signals are present. Contacts of relays 16A-K29 - and 16A-K53 will automatically close inboard valve MOV- l 1001-63 when one or more of the following PCIS signals exist: ! 1
- 1. High drywell pressure !
- 2. Low reactor water level
- 3. High reactor pressure l Contacts of relays 16A-K30 and 16A-K54 1
, will close l outboard valve MOV-1001-60 in a similar manner. ; i During a station blackout or severe accident beyond the plant design basis, the head spray isolation valves must be opened in order to pump water , vessel. To accommodate this change, bypass to the reactor , are added to panel C903 for each switches contacts of relays 16A-K29, K53, K30, valve to bypass l and K54 and ' override the PCIS signals. To preclude the possibility
- of overpressuring the RHR head spray piping, the existing logic is modified such that the high reactor pressure PCIS signal will override the bypass circuitry ,
if the reactor pressure is above the setpoint. This is 87104.1 l
~ " d W D .: d ?N.h d '. % '. 3 .0 E . 3 h W W N n . L E.:h ' ~;:s s l..' ' GT,2NTM ' F '
AL 14 'E7 OS:43 (617) 423-7227 p,a Safety Evaluation.No. E18 E Rev. O FrtN 86-52B-1
- dheet _j[_ of jl}_
accomplished by taking contacts of relay 16A-K28, which deenergizes a on high reactor pressure and is powered by Div. A power supply, to MOV-1001-63. Similarly, the control circuitry of ,
)
contacts of relay 16A-K50, powered by a Div. B power supply, MOV-1001-60. are used for Each of these valves can be tested if High Reactor Pressure is not present b bypass switch in the " BYPASS"y placing the respective the respective control switch position and then placing The byptnss switches are in the "OPEN" position. keylocked, maintained contact and havs two positions: " NORMAL" and " BYPASS". . 4. Presently the power to FV-1001-58 and FT-1001-86 is not backed by a DC power supply. In order to ensure the flow control valve is available during a station blackout,bypower backed will be provided to panel Y2, DC power. which is valve. Presently Note that MO-1001-60 is a DC MO-1001-63 is not backed power supply. by a DC be inoperable, in the closed position.Therefore during a SBO, this valve A reliable source of power will be provided to this valve when the DC - SBC Battery / Inverter insure opening of the valve. System, PDC 87-03B is installed to
- 5. A strainer, and local differential pressure gauge PID-4610, will be added to the 8" cross-tie.
The strainer To will be welded accommodate into the removable spool piece. the strainer, l removable segment is increased tothe length of the the differential pressure gauge are 16". The taps for , i located on the installed pipe piping segment is such that insertion and removal of the not affected. The strainer is of , conical design and is sized to trap particles 1/8" and ' larger. 6. A at a 1" drain line with one gate valve will be installed l low point on the KB piping on the 8" cross-tie. The drain line to a clean radwaste located on the GBB piping will be routed ; drain hub located in the reactor auxiliary bay at elevation 23'-0". The drain hub leads to the reactor building squipment drain sump in the HPCI room. The KB piping drain will the GBB piping drain and the be connected to drain valves will be 87104.1
75 m m u y s v g .L e:a c.a;:a.".T ~ :~:: . .: 1 M ;; m .;1 .' m n :i H ;a .+:w: = <=F:' %:. :.a '. ~ :w aw L:m : " JUL 14 ' 07 cO 8 4*.) . ( C17 ) 403 7007 r.5 Safety Evaluation No. 7.)T ?
, Rev. 0 FRN 86-52B-1 .
Sheet __6_ of 23 normally open to allow draining of any leakage past the isolation valves during normal operation.
- 7. Pressure relief valve PSV-1001-59 point upstream of its existing is relocated to a horizontal run of piping, location, on a access. to improve maintenance
- 8. To comply with PNPS Appendix R High-low Pressure outboard valve MO-1001-60 Interface commitments, the power an,d c -
fire zones than the willfor be rerouted to different 1001-63. Where this cables separation the inboard valve MO-normally energized conductors cannot be maintained will be pulled in separate raceway for the outboard valve ; of the control circuit then the remainder from opening due to any cable failure.to prevent the outboard valves B. . PURPOSE OF THE CHANGE The purpose of this modification power dependent is .to provide a non-AC. for extended station blackout and severeredundant water source for vessel beyond design' basis. accident scenarios line provides an injection path to'The reconnection the vessel of the head spray of a load to the that is less station-blackout battery than the LPCI ; injection path. The strainer clogging of is being- added to prevent fire or city the containment water spray cap (PDC No. 86-52A) when is being used. provides for disposal of The 1" dra.in line building secondary containment sump. contaminated water to the rosetor C. \ SYSTEMS, SUBSYSTEMS, COMPONENTS AFFECTED . ! I 1. Residual Heat Removal System The 18"-GB-10 (RER "A" train to piping) is affected by this reactor recirculation modification. The 4"-GBB-10 RPV head spray line to the RPV is reconnected to the 18"-GB-10 piping. l The 4"-GBB-10 RPV head modification. The head spray spray line is affected by this to the RHR system line is to be reconnected and RPV and will Control valve TV-1001-58, flow element FZ-1001-85 be rerouted. and 87104.1 l l l l l l
. . I
r,~G15mm.c':G.x.T2 ccw.. ;;::::mt.w." - :. xu am .u an- an w.a, . u.w. w .:m: n:w: ~w 24 es 09:50 (617) 423-7027 .6 Saf ety Evaluation No. _3182. Rev. 0 , FRN 86-52B-E Sheet 7 of E7 FT-1001-86 are located as close as possible to the 18" line. The power supply to FV-1001-58 and FT-1001-86 is being relocated to panel Y2, which power supply. is backed by a DC backup nitrogen supply. In addition, connected TV-1001-58 will also be to the a 1" bypass line with restricting orifices is .added around FV-1001-58. The motor and actuator accessories for RHR head spray containment isolation valve affected by this modification,(MOV-1001-63) since they will be will be replaced with qualified components. existing cable in the drywell to In addition the replaced with qualified caban. MOV-1001-63 will be Pressure relief valve PSV-1001-59 the 4" head spray line is affected bywhich this is located on modification since it will be relocated to piping to improve maintenance access.a horizontal run of The 1" 8" cross-tie line is affected by the addition of a drain : containmentline routed to the reactor building secondary
- sump.
- 2. Reactor Building Area 1 The Reactor Building Area 1 is affected modification due to the installation by this of the new 4"-GBB-10 piping including supports, valves, fittings, tubing, i conduit and the 18"-GB-10 header to the RPV. associated components. connecting 3.
Reactor Building Auxiliary Bay Area 2 The this by Reactor Building Auxiliary Bay Area 2 is affected modification due to the installation of the l strainer and the 1" drain line.
- 4. Fire Protection System The Fire Protection System is affected by the installation of 8" cross-tie line.the strainer in the spool piece on the 87104.1 l
I l i 1
p m ;a m a r-a.~..w = :.m i = . w . a v n => ;z m m..e: +. ~ .r s m-- w m m ' u w. g . i.: 57 09 $0 t617) 4E3-70E7 F.7 Safety Evaluation No. 218 1 Rev. 0
- FRN B6-52B-t Sheet _j[_ of 23 5.
Primary Containment Isolation System (PCIS) The the Primary modifications Containment Isolation System is affected.by to containment isolation valves MO-1001-60 and MO-1001-63.
- 6. Control Room The control Room is affected by the addition keylock switches to override the PCIS of the
.MO-1001-60 and MO-1001-63. for valves D.
SAFETY FUNCTIONS OF AFFECTED SYSTEMS / COMPONENTS
- 1. Residual Heat Removal System Thr is tosafety provide function core ofcooling the residual heat removal system in con Core Standby Cooling Systems, junction with other containment cooling and to provide as required during abnormal operating transients and postulated (Ref. FSAR Section 4,8.1). accidents
- 2. Reactor Building Area 1 ,
l The safety function of the Reactor Building is to limit .. the that so release offsite to the environment of radioactive materials below the guideline valuesdoses from a postulated DEA will be the Reactor of 10CFR100. In addition, for the engineered Building is designed safeguards to provide protection and systems located in the building fromnuclear safety all postulated environmental events including ; Sections 5.1.3, 5.3.1). tornadoes (Ref. FSAR ' I 3. Reactor Building Auxiliary Bay Area 2 i This structure integrity.of safetyhas related a safety components function of protecting the housed in it (Ref. FSAR Section 12.2.1.1). and equipment
- 4. Fire Protection 9 I
The Fire related Protection System function. does not perform any safety The Fire protection , System is 87104.1
[TE-~ M Z Giji% @ ' DW::V M ? i ~ N T: .
" ai *.? .m ' W: ..,. . :2. =- ;n. ny . ..ic.:m .;nD .nc;,
I 7A 12 '57 09:51 (617) 423-7017 F.E Safety Evaluation No. j2ISE Rev. O { FRN 86-52B-1 ' i Sheet 9 of_d2_ important of equipment to damage safety because from a fire it mitigates the occurrence significantly degrade the capability that could damage or of one safe shutdown system in the plant. 5. Primary Containment Isolation System (PCIS) ' This systemagainst protection has the safety function of providing timely the onset and consequences accidents involving the gross release of materials from the of radioactive automatic isolation primary containment by initiating of appropriate penetrate the pipelines which primary variables exceed pre-selected containment whenever operational monitored limits. .
- 6. Control Room i Thearea au) safety function of the Control Room is to provide whera the operators can during norme.1 and abnormal control the plant conditions. a Room is designed radiological to withstand exposure k seismic event and limitThe Control to the operators j accident. .
during an ; E. ; EFFECT ON SAFETY FUNCTIONS ,
- 1. l Residual' Heat Removal System .r The new 1" and l closed valves form a part of the residual 4"-GBB-10 head spray line system pressure boundary and are required heat to removal maintain j their design basis integrity. '
l I { l The original head spray line was subject to waterhammer which resulted in damage to the piping and supports. l
- 2. Reactor Building Area 1 3
l The installation of pipe supperts I on the 4"-GBD-10 piping concrete affects secondary anchor bolts containment due to drilling of and due to piping loads i transmitted to the concrete. : 87104.1 I
n.,-n..-.-._ . . . - ~ . . . . ~ ~ . - . . _ .. - . . _ . , . ~ , --. . _- ,_.~. --
.l i
1
, , yg 1.: 'E7 O M E L61n 423* M F9 -
=
Safety Evaluation No. 2M Rev. 0 FRN 86-525-3 * : Sheet 10 of 23 3. Reactor Building Auxiliary Bay Area 2 . The installation of pipe supports on the 1" drain line affects the Reactor Building Auxiliary drilling'of concrete anchor bolts and due Bay due to to piping loads transmitted to the concrete.
- 4. Fire Protection System The strainer will be normally removed, and iswill be located in the spool piece, which the FPQ boundary. located downstream of -
5. Primary Containment Isolation System (PCIS) The Prima"y Containment Isolation signals
.Drywell Pressure and for High containment isolation Low Reactor Water level for can be bypassed through the use of thevalvesbypass MO-1001-60 switches and MO-1001-63 ,
located in the control room. , The replacement of the and drywell cabling for motor of power- and and actuator accessories MOV-1001-63 and the rerouting control cables effect on the safety function of the for PCIS. MO-1001-60 have an .
- 6. . Control Room .
The bypass of Control Room switches Panel C903 is affected by the addition for i MO-1001-60 and MO-1001-63. containment isolation valves ; F. ANALYSIS OF EFFECT ON SAFETY FUNCTIONS
- 1. f Residual Heat Removal System i i
The new 4"-GBB-10 head spray piping i normally closed and the reused 1001-63, and isolation and control valves (1001-60, [ They are seismically FV-1001-58) are quality category . designed and supported. "Q". design ensures that the The basis criteria of the system. modifications meet the design , maintain the inte'grity of the RHRS Hence, the modifications and the pressure boundary this modification. safety function of the RHRS is not affected by Fire water will only be 1 27104.1
- usedtoj
,, , f,'c g) lgfl#1 Peference5 to doc.umemY
'+h t Sens$s c. gu9}lft'cOion ennl I
wiIi he edded tothe >bc b m,ysis y nor FRN, fS teSH "fA H g lo . ls 9
&L.:s a. . L.Ca G 'sG .n.3 % & .w JP.:.u W.iG 1. D.5' .:.4 7'+. ';.N.2 M D iWi :e "WiW' :2DJT ~ 2T' JUL 1.: 'E7 03:5e tela 41-/ W O: ) Safety Evaluation No. j218E Rev. 0 . ' FRN 86-52B-d Sheet il of 2. 3 i cool the RPV during a station blackout / severe accident event which is beyond the design basis of the plant. The not. replacement affect the oforiginal the motor and actuator assembly does System other design function of the RWP-than improving therefore does not compromise its reliability and interface of the system. the safety related The valve operators will to meet 10CFR50.49 be environmentally qualified requirements. actuators will make the system more The replaced valve safety. reliable for plant The rerouting of the head
- relocation of the FCV and spray piping and the alleviate the previous assr>ciated bypass line ensure that the pressure waterhammer ' problem and thus it's safety functions remain intact. boundary of the piping, and
- 2. Reactor Building Area 1 .
in accordance with .theConcrete applicable drilling for pipe support anchor b PNPS procedures. This specifications and : the structure is ensures that the integrity of not degraded. Hence, ~ modification Reactor Building.does not affect the safety function of this the 3. Reactor Building Auxiliary Bay Area 2 Concreti in accordancedrillingwith for pipe support anchor bolts will be PNPS procedures. Thethe1" applicable specifications and drain line will also be seismically designed to meet ! ensures that the II/I criteria. degraded. Hence, integrity of the structure is This not safety function of thethis modification does not affect the . Area 2. Reactor Building Auxiliary Bay i i
- 4. Fire Protection System )
The addition i of the strainer in the spool piece on the Fire Protection System. downstream of the FPQ boun 87104.1 , u-_ _m__ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _
g;xzn un a = . - M ~: ~ "' " * ' ~ ~ ~ ' n la '57 09 52 1617) 423 02~ Safety Evaluation,No. 2l8E Rev. O TRN 86-52B-1 . Sheet 16 of 23
- 5. Primary Containment Isolation System (PCIS)
The addition of keylock bypass switches does not affect the safety function of the Isolation System and meets the design Containment Primary criteria of Section 7.3.3 of the FSAR. Under normal operation valve MO-1001-60 and MO-1001-63 will isolate on a group 3 isolation signal. In order to bypass the isolation signal the operator will' be bypass switch into the bypass required position. to place the procedures will be put in place Station simultaneous operation of the bypass that will only allow design basis water sources are unavailable. switches when The replacement of the motor, actuator accessories and drywell cabling for MOV-1001-63 with qualified components improves the reliability PCIS. The rerouting of power and MO-1001-60 improves the PCIS by control cables for Appendix R, High-Low Pressure complying with PNPS Interface commitments. This prevents the simultaneous opening of both inboard and outboard containment isolation ~ fire. from a common area
- 6. Control Room qnol oAev ekss IE eppeat I w>k The addition of bypass switches to panel 903 does not 7/Ulf 7 affect the safety function o. the modifications to panel C903 are seismically Control Room. The by BECo and are to be qualified The switches have been added to the PDC by FRN later. '
Evaluation evaluated in Human Factor i Study 287-010 to address NUREG 0700 guidelines for control room design reviews. Specific calculation references to document the seismic qualification by minor FRN. analysis results will be added to the PDC , I G.
SUMMARY
This modification does not call Residual for the operation of the Heat Removal System at higher pressures, temperatures or more severe conditions than the existing 87104.1
Wmmm:g2w..acGw .c: . a: . m t. n mi t..: ~ u.:.v o nra:a.:.w n un : . x . . xv w . . p.n. 14 '87 09:54 (617) 423-701~
~
Safety Evaluation No. 2182. Rev. 0
- FRN 86-52B-i Sheet 1 of 23 design basis, therefore the modification does not increase the probability safety. of malfunction of the equipment important to The isolation bypass switches have lighted position indication, are designed single failure proof, and are administrative 1y controlled and-as such do possibility for an accident or malfunction not create the type than any evaluated previously in the FSAR.of a different I The isolation bypass capability be limited to Station Blackout or severe accident cond so the marginisofnot specification safety as defined in the basis of technical reduced.
This modification 'is designed such that the failure of any non-safety related components q capabilit of will not degrade the ' function y safety-related equipment to perform its ! s - Change.This modification does not require a Technical Specification ' This modification does question. not involve in unreviewed safety , h i i t l f a l 87104.1 1 6
E !;.H % .c Gr;:s.% W a Am.u: :..m: . . : . zw! .;;' N .' :::: " .. m ..: ; ' : v . x ..Er z. 1 l Safety Evaluation No.: 21% 2. EMETY EVALUATION WK SWErT pgge.528-Rev. No. O
$MAET A. Systee Structure Component Failure and Consequence AnaTyses. if OF 23
. System
- Structure Comoenert Failure 14 odes Effects of Failure % nts Residual Heat See Attached Sheets 1.
Removal System *. 1
- 2. Primary Containment See Attached Sheets l Isolation System 3.
s General Reference Material Review FSAR cal.mt.ATIONS REGJLATORY SECTION PNPS TECNNICAL SPECS. DESIGN SPECS PROCEt11RES EJIDES $TANDARDS CDOES 4 2r 3,7 Spec. M-300 10 CFR 50
/C,7 Spec. M-301 ANSI B31.1-W80
/0,9 Spec. M-600 ASME Section III-W80 l 7, 3 -
- 3. For the proposed hardware change, identify the failure modes that are
. likely for the components consistent trith FSAR assumptions. For each failure mode, show the consequences to the system, structures or related components. Especially show how the failure (s) affects the assigned safety basis (FSAR Text for each systse) or plant safety functions FSAR ".
Chapter 14 and Appendix G). Prepared by - Date 7!I3 !i7 NOTE: It is a requirement to include this work sheet irith the Safety Evaluation. - Exhibit 3.07-C 3.07-18 Rev. 4 .
CLU:V ; % 2.a..L:~i:G a i:5.G C u . : ww::: ,: M.::.. z : 4 ';;; ;;i
. . v L.wi.r r:2:2.m- =-. w . n x ?.:. x Ja 1 : 'E7 O M 5 561~' 4 0 ~70 U
- c Safety Evaluation No. ElbE-Rev. 0 .
FRN 86-52B-1 Sheet 15 of 23 SAFETY EVALUATION WORK SHEET A. System / Structure / Component Failure and Consequence Analyses The following apply only to the proposed modification p&r FRN 86-52B-1:
)
- 1. System / Structure / Component: Residual Heat Removal System Failure Mode: Structural failure of the 4" Head Spray Piping Effects of Loss or degradation of the capa-Failure: bility of the RHR System to perform its function during a design basis.
Comments: The 4" head spray piping and components are designed and 5 installed to the requirements of !' the ASME BPVC Section III , Subsection NC for Class 2 i
, components g Failure Mode: Structural frilure of the 8" cross- I tie piping tjue to the strainer or ,-l 1" drain lire. . j Effects of Degradation of the capability of Failure: the RHR s(stem to perform its function d .2 ring a design basis accident.
Comments: The 8" cross-tie piping is "Q", seismically designed and supported to the first anchor beyond the isolation valves. The- 1" drain ! line is ASME BPVC Section III Class 2 up to the drain valve. Tn 5Johb',,Ae.rdocdon of Fv-lool-5r A in hll.h s of +ke. i" byprs ITne) and d e_ in s-% lk h'on o f + h s ven+ w n l re Jae th e. l 67104.1 Possibl1,'f7 of strudurei foii /vre km vsfer ! bWner, pg ffg/g
LTT:,=.=5.5%2ex zu ux ww . ~ :w. .x . ,;a. :.; u. .:a .2 . : .:. .v -= r . -
~
= w.e r. .w: ...>
)
n it. 'E7 09:55 (617) 425-7017 Safety Evaluation No. D N Rev. 0 . { j FRN 86-52B-A Sheet _,]f _ of 23 i I
- 2. System / Structure / Component: Primary Containment Isolation System Fa'ilure Modes Inadvertent bypass of containment i'
isolation signal for valves MO-1001-60 and MO-1001-63. Effects of Possible loss of vessel water in-Failure: ventory during accident or transient th'at initiates containment isolation signal. Comments: Keylock switches are installed to [ bypass containment isolation t signals for valves MO-1001-60 and MO-1001-63. Use of the keylock switches are indicated (white light ; on board). Requires two failures - to open line (both valves must be . opened) and high reactor pressure will still cause valves to close. i Additionally, a check valve is located between the vessel and ; valve MO-1001-63. 2 sp+e,,/s w k ,Jc.,p a,,1: Res*, dud k,+ kon i Sp % Fu kre Med <. : Fa,\ce. of LLRT (Age 8 Jag r ([,
- F F;il*e of % ,* J ;J a*8 o 9-u : am cou_w,u.ae, Leihs useol.
M*e Mode i Falce. of s),% v&e ,ms,n EFre.chs of Fdice ; h '.lu ve o f Coh k eem4 hoh Woh l 87104.1 bo 0d, , t e
W;- .yc 7% .e .. w.r. :; i , a> ; y. . .; .;,:. . . . : .: -z a: . , . . ~ . . ;. , , ,. riusulO i f ATIC3( F11A REVI h 1HEET FRv %-5'2B -
References:
SHGE.717oP 23 Safety Evaluation: DE Rev. No.: O Date: ~/ 87 f4 7 Support a change for modification to the Residusi Heat Re:noval S/eten. List FSAR test, corresponding FSARdiagrams, revision. and indices affected by this thange and Affected FSAR taettan Revision to affected FSAR Section is shown en: Prettsinary Final Fie. & _ R -1_ , N M-241 $heet 1 Tit. 10.7-1 m M-212, Fit. 10.8-1 AMeehagabS M-218 sheet 1
~7. 3 Attachment #1 S le. 7. 3 - i ,
Attachment 9 i Attachment 8 ' Mft.1MIRARY FSAR RfY111GI (to be completed at time of Safety gyaluation preparatten). Prepared by: I d N /Date: to: 7/d; 13ff7 Revleued M Approved by:, ,
# Dater 7[/M&7 ,
gpr 7j Q,, . . f,. ! e '.
. . . . . . i.' . .
. t. . ..:
FINAL FSAR RfV111fal (Prepared fbilerf ag operstianal turnover'ef'tgitted ' ' j[. - systems structures of components for ese at PWS). . .
.~ .n..
. 1.
Prepared by: #Date:. Revissed ky
. #Date:
. ~
Eshtb1t 3.07-A
- But 3 W 3 ,,
4 3.07-15 Dev. 4 -
~ :'.' l e
, 'l t,*
- _ - _ _ = _ _ _ _ _ _ _ _ _ _. ._
pxcai z+.r : ra.s. i:.u. 'De::m.nsm..' . s w .x w ,. :v, v ::.. u x. ~ r:mm .:w :~ n .: e ~ mm--:--~--- AL 14 '67 09:56 (617) 423-7027 F.1! i Attachment 1 Saf ety Evaluation No. 2 3I2. AfV.O l
.FRN 86-52B Sheet of 4 33 RECOMMENDED FSAR CHANGES .
Section 7.3 of the FSAR needs updating due to this modification (FRN 86-52B). FSAR pages 7.3-12, 15, 15a and Table 7.3-1 Sheets 2 and 11 have been marked with suggested updates and are included in this Attachment for your review. The following drawings will be revised as part of this PDC, but are not included here. Dwg. I.D. FSAR Figure Title M-212 10,7 1 P & ID Salt Service i Water System l
- M-241 Sheet 1 4.3-1 P & ID Residual Heat Removal System M-218 Sheet 1 10.8-1 P & ID Fire Protection System 8
. j I
1 i i h 1
}
87104.1 l 1
n52,2,2.;. s.i.nn .%w n 1:4 ti.c w w ' ' :a n x = .. m:: .:m: .. a 'i. 'v . : .. a r M ATT'CH!n541J. A PNPS-FSAR NEW EVALV477ev h' Asv, o D & 2. Traversing incore probe D RHR reactor shutdown cooling supply WEETl9 oF r.3 RHR reactor head spray Reactor water sample lines Reactor water cleanup i Drywell purge inlet and makeup gas *, ** Drywell main exhaust Suppression chamber exhaust valve bypass *, ** Suppression chamber purge inlet and makeup gas *, ** Suppression chamber main exhaust l Drywell txhaust valve bypass *, ** RHR-LPCI supply RHR to Radwa'ste containment atmosphere sampling lines . i
- Containment makeup and ventilation valves are also provided for use following an accident condition.
These are remote manual operated. Refer to Section 5.4.3.
** The reactor water low level isolation signal can be by-passed. These valves may be opened anytime provided the low-low water level signal is not present.
. The second and lower of the reactor vessel low water level P
- Ct !(' By ress sw, & $ isolation settings was selected low enough to allow the
-g g,,y f % . g , removal of heat from the reactor for a predetermined time
. 11 wing the scram, and high enough to complete isolation M *hC Whol VooM ['in time for the operation of CSCS in the event of a large YO Over'rlde Me, break in the nuclear system process barrier. This second I low water level setting is low enough that partial losses of Pascto* W ate.P feedwater supply would not unnecessarily initiate full
/c /.7vglgnej isolation of the reactor, thereby disrupting normal shutdown P % A 7 co n b, ne.,f'or recovery procedures. Isolation of the following b6h p ,sso pipelines is initiated when the reactor vessel water level falls into this second setting.
*IOEob Ss'y NS, All four main steam lines HPCI, RCIC, and main steam line drains O 7.3-12 Revision 2 - July 1983 I
A __ __L_L
[HB XZ nii3 5.B ::ac a m n.: w .~ 3 z L :: ME
. L3: .....t 'i;'x' cn. " e ' r/ .
1
) . ATTncHmEar L TAfTTPEvoLvoTc4 en 87.
PNPS-FSAR SIE Y C) Ftw'86-525 included here to make the listing of isolation functions 57 DEET 20 OF 23 , complete. l
- 6. Primary Containment (drywell) High Pressure ,
High pressure in the drywell could indicate a breach of the nuclear system process barrier inside the drywell. The automatic closure of various Class B valves prevents the ] ' release of significant amounts of radioactive material from the primary containment. Upon detection of a high dryuell pressure, the following pipelines are isolated. See Title 7.3-1, Signal F. Traversing incore probe RHRS shutdown cooling su 1 RHRS reactor head spra # Reactor water sample lines Drywell equipment drain sump discharge Drywell floor drain sump discharge i Traveling incore probe tubes Drywell purge inlet and makeup gas *, ** Drywell main exhaust ' suppression chamber exhaust valve bypass *, ** Suppression chamber purge inlet and makeup gas *, ** Suppression chamber main exhaust Dryvell exhaust valve bypass *. ** l RHR to Radwaste containment atmosphere sampling lines The primary containment high pressure isolation setting is selected to be as low as possible without inducing spurious isolation trips. See Table 7.3-1, Signal F.
- containment makeup and ventilation valves are also provided for use following an accident condition.
These are remote manual operated. Refer to Section 5.4.3.
~
7.3-15 Revision 2 - July 1983
2am./.: .# u.;;cw a.=,..;, aa m . ~ w .x..mz -. w, n ef wm .w- = - - - - SAFETTEVol.U$7lod W. El%I PNPS-FSAR W. O MN 86-52 B SHEETzioF 23
** The reactor water low level isolation signal can be by- "s, passed. These valves may be opened anytime provided ,
the low-low water level signal is not present. ,
- 7. RCIC System Equipment Space High Temperature .
I High temperature in the vicinity of the RCIC System equipment could indicate a break in the RCIC steam line. The automatic closure of certain Class A valves prevents the excessive loss of reactor coolant and the release of significant amounts of radioactive material from the nuclear system process barrier. When high . temperature occurs near the RCIC System equipment, the RCIC turbine steam line is isolated. The high temperature isolation setting is selected far enough above anticipated normal RCIC system operational levels to avoid spurious operation but low enough to provide timely detection of an RCIC turbine steam line break. See Table 7.3-1, Signal K.
- 8. RCIC Turbine High Steam Flow
~
RCIC turbine high steam flow could indicate a break in the RCIC turbine steam line. The automatic closure of certain
~
4 *
- 8 pass 7 s-,+ches hue, been prog,d,cf;, e e_ @ i Co'dvol Yoom 10 O Y'2YhN E. O T- YG9C koy W6SCY l low level '
oincd pN freSS#e. tiolclY)on mcy 5 sej % 5
/
, containment }Qh V
7.3-ISa Revision 2 - July 1983
n u 7 et k p u s o o o s 1 b t nel d o (
)
e h ~ s g g eef c r e s l T nc e lc e c n) in) h r wne e n r r i 2 ) ) 2 ) I leII
#b o o e o 2 2 .
6 l( 2 ess ( p ri t t i t l( t ( ( G t ( s a s bI s s n u t e h t e ept a* e a ol e e ot r o e s mn e s t r o t t C O t o pe t o teb o o o o hN
) ) hp D Cloui N
( M n ( M s T N h N N T
- T e
lew1 o2) d d d d d d d d d d d d st( e e e e e e e e e e e e
- r a s s s e s s s s s n s s e ot) o o o c o o o o o e o o o l l l l p l l l nS9 l l l i l C C C C O C C C
( C C C C C l e g n b d d n e le l d d r d r d r d r d r a r r t) e e r a e a a a c a e is a7 / n f n d a d d d d d d d lod C ( i i n a a n n e a n n a n a t ilp n a n e 0 0 t t t t t t u g, t t
. . S S S S S B o S S
$ 5 S n w ol o e l it ne s r E.
f a ta F. F. F. F. ew A. v v o e U. U. S. S. S. U. U. M el t D e lo S_A t A C G G A A G G R ng I D
)
3 e) ( rs6 eo( e e wl oC) s s e P 5 e c c o( o o
) t c c c e c c c r c r
- t e c c a a d o o o P o P n d o a o R o A C
( S F 1 S 3 P e e e N 7 no e e e e e e e e P ot 1 d e d d d e d d d d d d l d d l l d i l l l l t 1 l le e e i s e e e e e i t.t t e n a s t t t t t t s B t s n t u t b ts n u u s u r u n A c e O O O O O O I T lerM O oh I O O O I s s a I. I. I. I. I. I I
- I. I.
B B B A A B B B B A A l A A C n) re6 e p( wO ) ) b I o3 c 3 t( e c c e c e e e c t e d e o s o s d s o a . ( ep e k y e e e e e e c T) e e b e e i e 6 t o t t t t o t t t e h t e n a a e l a a k e( n l C U G G C C D D c C C e C i t e 0 0 _0 h l e 0 l e 0 0 1 e 09 09 09 C 9 A v 9 lM 9 t 9 9 9 n o 8 5 8 8 l i 9 B. B. B. lea t 4 B. 4 4 4 4 A A A 4 0 0 0 0 wr 2 1 1 9 9 7 ? 1 1 1 1 1 5 1 5 yt 2 2 2 2 ir e n 1 k 2 2 3 3 1 I 3 I I 3 I I I I I f e I I I P r y c c
. e a e
. e d
t n t n u e e t t r . . p r r d u u . a e n e h h e e m e e e ow ow s g s eh n n r r e er e e t o t a n r ia , o o r p r pb e a n s t o d ri r ps le t s t y m e t e ce c e pl pe t w t I t l t lo ol o pe t n u e t o c o t c ue sg r n o nr op ar e p a r. e g uo s op ue sh il n o Kh Kh 1 e t scs e a c cs r s re c ef ie L e n r e e e l J o g o o - ,u - P ep t p t t n - e s .d . d t n t n . y . w n w o r n e a o o s o rt R o t Rl M oe D uus i L R u R l N p R e R e R e R Rl R ae M t e M e Mh M h N i e H E R M f Ct r M d s P H R N e B M e R R R D R 9e t F D T F
. I
L l - -
.. j 8TfAc4MEAlf ,1 i 9%TYEVM.Va&O 1]Q 22L PAN 86-52 8 N'Y * \
l l SHEET2.3 of ?.3 l PNP 5-Malt Tanle 7.31 (Cont) 15 saanual svitetes override all autoestic signals on see sem11er valves that typnas the suppressico chamber and dryvell exhaust valves.
- 16. Signal *A* er *r* eauses autoestic withdreval of TIF probe. When probe is withdratm. the valve autoestically closes by mechanical action.
- 17. las.htico signal *A" is permitted to close LPCI valves but only .
vtan the Rai7 shutdovo cooling supply valns are not fully closed
- or resi, tor pressure (sipal U) is below 75 peig. Innate positten todicat163 lagets are act required at the isolatico salve display panel.
- 18. By re ss tw,ickes hwe. keen gov, ele of ik O'- Lonk"0) MR fo overyo de Iso)nhos S&m ls ,A & "F .
' m s .
e 9
'a g- I d.$
- 9. %
- a ;
P.', t a
*9 5
o==- h m' _m.__ _ _ _ _ _ _ _____ m .. m .._ _' = ' F
=
O< . i.. 62 %:H;:.%. s . ~ - '
~ i:- < '- ' 'A ~
" ~ ' ' " '
301 492 7921 iM8 mC-M&M-MNBB NO.003 001 M'h0'*7 g g ' m Q ,p m (A te N M ! J *M S , NCMORE HARDCOPY R 1 2 A' '# 4 r-vfW'f _. From: NEWS Posted. e 19-M y-E G. Sy ' (39) Subj ect : NUCLEAR REGULATORY COMMI O :nu- r at or y. - ami s si on as drag
- UPI STATE Wire (MASSA S) ,
w Dukchis assails NRC delays ) h' By JERRY BERGER UPI Statehouse Reporter '*",f
's '
BOSTON (UPI) _ The Nuclear Regulatory Commission has draqqing its -
>b, foot in dealing with a call for the permanent shutdown of the Pilgrim, nuclear power plant, Gov. Nichael S. Dukakis has chargod. ' V / 44s4 In the letter to NRC Chairman Lando Zech, released Tuesday, Duk(LOs_ I, quoctioned the 10-month delay in acting on a proposal that calls on Ecoton Edison Co, to prove the troubled Plymouth f acility should not be Wr**~#
cloced permanently. '
"The lack of a meaningful response *from the NRC after 10 months is totc11y unacceptable," Dukakis wrote. "Public confidence in nuclear pewsr plant operation will only deteriorate if the commission does not rocpect its own mechanisms for inviting public input and participation."
The Pilgrim plant has been closed for more than 13 months to deal with management and saf ety inadequacies uncovered since the initial chutdown caused when the emergency saf ety system was triggered twice within eight days. The NRC's own assessment teams have concluded the plant is not yet roody to reopen and a special Ingislative commission has been f ormed to 1Cok into the issue. Dukakis questioned how the NRC could avoid dealing with the the Pilgrim show cause order in the face of its own internal r evi ew.
"Despite the preeminence of the fedoral government in nuclear powir licensing, state governments cannot exercise the important responsibilities assigned to them under federal law if the Nuclear Regulatory Commission refuses to even respond to our legitimate concerns or act on a duly filed show cause petition," he wrote.
The position is similar to the one Dukakis adopted in refusing to $ approve state evacuation plans for the Seabrook, N.H. tiuclear power plant. A Democratic presidential candidate, he has contended a safe-ovecuation is impossible in the event of a Seabrook accident. . That stand has delayed licensing of the New Hampshire plant, , although the same argument could not be used to prevent the re-opening of Pilgri m, which has already received its operating license. j
.-e ---m-(
c {, 7 l> Q @N [$ e 1
- .
- . . ~c. w : w :.. w ;:.,.+. :;. ..:e 2 -. i.. w - a~. : * . c. .:- ? = = => - ~b < ..:r *-: " ' - -- - '
)
JUNE 1987 OVERVIEW REPORT PLANT NAME: Pilgrim i PROJECT MANAGER: R. H. Wessman MONTHLY OPERATIONAL OVERVIEW FOR JUNE 1987 The plant continues to be down for an extended outage for refueling, resolution of equipment and operational difficulties, management issues and emergency planning concerns. l Restart not expected until Fall 1987. Monthly Licensing Action Summary: TRIPS THIS MONTH 0 TRIPS LAST 12 MONTHS 0 NUMBER OF REPORTABLE EVENTS 0 l LICENSING ACTIONS Opened in June 1987 - 5 Closed in June 1987 - 3 Outstanding 06/30/87 - 42 SIGNIFICANT ISSUES AND EVENTS Significant Activities Durino June 1987 On June 10, 1987 during implementation of a planned system modification, the licensee discovered the presence of corrosion products in the primary containment spray headers and nozzles in the drywell. Up to i inch of rust was found in the upper drywell header. The Events Assessment Branch (EAB) has reviewed the informaiton and recommended that a long-term review be undertaken by the Plant Systems Branch. In addition EAB recommended that the need for a generic communication be determined. Commissioner Asselstine visited the site on June 24, 1987. BECo, Region I, and NRR met in Peoion I on June 29, 1987 to discuss the status of the Pilgrim outage and BECo plans for making the facility ready to restart. O T/f7
c=v- ri=:mn7mx .::=i;. s: :nwn::. u:vu ::. = ;,w :c: r:a v:;u n ..: m a. m :.v - - -- ' - c.z
~ i . .: v;: w- -
t u Significant Meetings Scheduled During July 1987 (NRR, Region, IE, etc.) None Scheduled Items for Management Attention
- Utility is still evaluating restart schedule; informal estimate from BEco is for restart about mid-September. This appears to be optomistic.
- The response to Gov. Dukakis of Massachusetts regarding Senator Golden's 2.206 Petition was signed by Chairman Zech June 25, 1987.
- Number of licensing actions to be completed before restart may require extra resource connitments.
- There is a Commission status briefing scheduled for August 10, 1987.
Status of Items Previously Identified for Management Attention Much of the staff work on the response to the 2.206 Petition of July 15, 1986 from Senator Golden is complete. The package has been forwarded to OGC for concurrence. OGC has identified several concerns which may require upper management's attention. Next Scheduled Refueling Outage February 1989 cc: V. Nerses . R. Wessman PD 1-3 R/F M. Rushbrook O w________}}