ML20235N364
| ML20235N364 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 03/01/1989 |
| From: | Blough A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Collins S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| Shared Package | |
| ML20235N350 | List: |
| References | |
| FOIA-88-284, FOIA-88-285, FOIA-88-286, FOIA-88-287, FOIA-88-290 NUDOCS 8903010186 | |
| Download: ML20235N364 (380) | |
Text
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MEMORANDUM FOR: Samuel J. Collins, Deputy Director Division of Reactor Projects FROM: A. Randy Blough, Chief Reactor Projects Section No. 3B
SUBJECT:
PILGRIM INTEGRATED ASSESSMENT INSPECTION (IATI) OPERAT Enclosed is the Operating Plan for the Pilgrim IATI.
A. Ra 81ough, Chief Reactor Projects Section No. 38
Enclosure:
As stated cc w/ encl:
Pilgrim Restart Panel Members Pilgrim IATI Team Me bers B903010186 890222 PDR FOIA CARREYBB-284 PDR l -f i , :, .. - in Si'. r,-y:d wu ds!:ted
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,i ENCLOSURE Operating Plan for the Pilgrim Integrated Assessment Team Inspection
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- 1. Introduction Pilgrim and has been hardware shutdown since April 1986_.to resolve long term management problems. Region I will perform an Integrated Assessment Team Inspection (IATI) at Pilgrim to evaluate the effectiveness of licen-see corrective action programs and the readiness of the plant and licensee personnel to resume power operation.
- 2. Objective pilgrim's performance history over the last several years has been poor.
When performance improvements were made, they often could not be sus-tained.
technical In April 1986, Pilgrim was shutdown due to a series of recurring problems. A Confirmatory Action Letter (CAL 86-10) was issued requiring resolution of the technical problems and approval of the NRC Region I Regional Administrator f or restart. In August 1986, the CAL was expanded to require the licensee to address other long term management and hardware issues before the NRC would approve plant restart. Additionally, the licensee was required to submit a restart program to the NRC document-ing BECo's formal assessment of the readiness for restart, including a detailed checklist for assuring that all outstanding items have been satisfactorily resolved and that plant systems have been restored and are prepared for operation.
During the extended shutdown BECo has initiated numerous management and staffing changas, implemented several complex plant modifications, and made various program improvements. During this period the NRC has per-formed improvements.
numerous inspections to determine the status and adequacy of the The objective of the IATI is to review the adequacy of any issues not previously inspected or which required followup inspection, determine if improvements made are effective and appear long lasting, and determine Pilgrim. if BECo is prepared to support the restart and safe operation of
- 3. Areas To Be Examined 3.1 The IATI will focus on the following areas:
Management Effectiveness / Assurance of Quality plant Operations Radiological Controls Surveillance Maintenance Security Fire Protection {
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Enclosure 2 3,. 2 The following
.IATI inspection: ' attributes are to. be considered and examined in the -
Development and implementation of. management goals / objectives and how they are. understood / implemented at various levels of the licensee's organization; Planning / controlling. routine activities along with effective.
program implementation; Level of understanding by. workers / supervisors of'~ potential impact of day-to-day actions on nuclear safety; Attitudes of licensee personnel with respect to: nuclear safety;.
-Involvement by senior management in day-to-day operation of the plant (including visibility of ' senior site. and corporate management);
Effectiveness of training, direction, guidance, and supervision by first-line supervisors; Adequacy of. staffing in light of planne J Accomplishments; Role of QA/QC in monitoring activities and how their reports are used by plant management; Role of licensee in working with and overseeing contractor personnel; Effectiveness of safety review committees; and Communications / Problem solving process.
3.3 For the Pilgrim IATI, the following generic /long term problem issues will receive special emphasis:
Stability and effectiveness of the management team; Overall material condition (including housekeeping and decon-tamination effort) of the plant; Timeliness and effectiveness of corrective actions (including management attention to ensure resolution and escalation to senior management if necessary);
Interfaces, communication and cooperation among operations, maintenance, and health physics personnel; i
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Enclosure 3 Status system);
of the maintenance backlog (including the. fire protection 0vertime controls in all functional areas; Validity of licensee interpretations of Technical Specifications and other regulatory bases; Quality of plant procedures and procedure changes generated as a result of modifications; Worker and management support of radiological controls, espe-cially ALARA; Worker perception of the following items:
(i) management policies (ii) management and supervisory involvement and effectiveness (iii) existance of, and reasons for, low productivity Licensee internal tracking systems and validity of closecuts.
3.4 Additional items (by functional area) requiring inspection /vsrifica-tion identified as a result of the NRC review of the restart plan or the NRC restart checklist are included as Attachment E to this plan.
- 4. Organization and Basic Operation The team will- consist of..a senior manager, inspection 4.eam leader, shift --
inspectors, and specialist inspectors (Attachment A). The senior manager will be responsible for all activities of the team. Reporting to the senior manager will be a team leader. The team leader will be responsible for assessing and developingshift dailyinspector observations, specialist inspector findings, inspecticn assignments. The shift inspectors will cover all three shifts during an early portion (3 to 5 days) of the inspection.to provide around-the-c The duration of the around-the-clock coverage will be decided by the team leader based on ongoing The plant activities, findings to date, and efficient use of resources.
specialist inspectors will provide their own initiative using the attributes noted above and followup on shift inspector observations for which they are assigned by the team leader.
All team members, including the shift inspectors, have a functional area assignment (Attachment B). All inspectors should keep in mind that the functional areas of management effectiveness, training and assurance of quality span all the basic functional areas. i i
The pilgrim SRI will assist the team in an advisory role on understanding the licensee systems and methodologies. All contacts with the press will be referred to the PA0 or the senior manager.
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Enclosure 4
- 5. Mode of Operations /Information Flow Many of the inspection and monitoring activities during this team inspec-tion will be windows of opportunity (e.g. , shift turnovers, any ongoing plant evolutions, ongoing maintenance and surveillance activities, follow-up on any events that occur during the irspection, etc.). For such activ-ities, general inspection planning and guidelines apply. As observations start to come in and be proce'ssed for followup, they become the priority for review. Inspectors are expected to focus their activities so as to be able to meet the inspection objectives as discussed above. Additional guidance is provided in Attachment C.
As the inspection is conducted, the inspector should keep in mind and refer to the intent of the programmatic inspection procedures (IP) listed in Attachment C. The inspector will be asked to give a percentage com-pleted for the IP at the conclusion of the inspection that will be for-warded to the SRI. The IP's are only for reference and guidance and it is not intended to go to 100 percent completion for these IP's.
Any safety concerns, apparent or potential violations of regulatory requirements, or items of similar significance will be promptly conveyed to the team leader after the matter has been identified to the appropriate licensee representative (shift supervisor or other similar or higher level r..a n a g e r) .
Each team member is expected to work approximately 9-hour days,
!> as to provide for appropriate overlap with team personnel on other shifts. Inspection team personnel will meet at 7:30 a.m. daily to discuss observations inspections.
and expected licensee activities and adjust ongoing planned It is expected that the inputs of the swing shift inspector will be communicated to the mid-shift inspector and documented in the log.
Team management meetings with selected team members will be conducted at the end of the work day.
- 6. Schedule and Shift Coverage j
The team will be onsite as necessary, during the week of June 6-10 for j l
badging and inspection preparation. The inspection will begin on June 13 j and run through June 24. Team members are expected to be onsite i
June 27-29 as needed to complete the report documentation.The tentative schedule is presented in Attachment D. The inspection coverage and end {
i date may be adjusted based on experience gained during the course of the inspection. Although the licensee shifts are 0800-1600, 1600-2400, and 0000-0800 hours, shift turnover is typically conducted one-half hour before shift change. Therefore, to provide for review of plar.. conditions before j licensee shif t turnover, NRC shifts will be 0700-1630, 1500-0030, and 2300-0830 hours. This provides for a nominal nine-hour work day with a 1 half-hour meal break. Other team members will generally work from li 0730-1700 hours to provide for some overlap of each shift and to corres- '
pond to station staff hours.
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Enclostre' 5
- 7. Inspection Documentation Each shift inspector should place their significant observations and activities witnessed in a common log (bound ' green book), which is to be turned over at the end of each shift to the oncoming shift inspector. It is to include:
significant events that occur during the inspector's shift and.
important observations thereof; significant information exchanges, including source (person (s));
concerns identified, including bases along with the identification of facts versus perceptions; and routine ' observations, including activities observed and applicable documents; e.g., procedure (s) by number, revision, and complete title.
It is important to note positive, as well as negative, observations. The inspector assigned as lead for a functional area will be responsible to report'on that area. Again, positive along with negative. observations are needed for proper perspective. The following major report sections /
paragraphs are expected to be a part of the final report:
report(Warren) tive); cover with inspection summary and inspection results (evalua-general overview; (Warren) operations, including event response / assessment; (Wechselberger lead, All) maintenance; (Rebelowski, Lyash) surveillance testing; (Lyash) radiological controls; (Dragoun) fire protection; (Wechselberger) security / safeguards; (Smith) training and qualifications; (Mcdonald) and assurance of quality, including safety review. (Rossbach lead, All)
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Enclosure' 6
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i The specific forinat. and content will be s with concurrence by the senior manager. pecified by theeach In general, teamsection leader should include: objectives / acceptance. criteria / reference; summary of items reviewed; significant findings (positive and negative); and conciusion.
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- n ATTACHMENT A' ORGANIZATION.
I SENIOR l.
l MANAGER -1 1 -l l S. Collins I l i I
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l-I . TEAM i l TECH. l l LEADER l- l ASST l l l. l- l l R. Blough l l C. Warren l I I I I l
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l l ADMIN ,GST l
-l - l l M.J. D Oonato l l l l l
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I I l l l l 1 l l l SHIFT INSPECTORS l l SPECIALIST INSPS l l l l l l L. Rossbach l l T.- Dragoun l
l J. Wechselberger l l G. Smith l l D. Ruscitto l I D. Mcdonald l I F. Akstulewicz ! l J. Lyash l l l l T. Rebelowski l i I I I I I
____.____--.______-----_.____.__m__ _ _ _ _ _ _ _ _ _ _ .
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l' ATTACHMENT B Pilgrim IATI Roster Senior Manager: Sam Collins Team Leader: Randy Blough Technical Assistant: Clay Warren Administrative Assistant: Mary Jo DiDonato l .
Shift Inspectors: Larry Rossbach
. Jake Wechselberger L Dave Ruscitto L Frank Akstulewicz Specialist Inspectors:
HP: Tom Dragoun Security: Greg Smith Surveillance / Maintenance: Jeff Lyash Ted Rebelowski Training / Management: -Dan Mcdonald Additional Area Assignments:
Fire Protection: Jake Wechselberger 4 QA/QC: Larry Rossbach Review Committees: -)
Inspection Report Coordinator: T. J. Kim l
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ATTACHMENT C INSPECTION GUIDANCE Administrative Guidance To Inspection Personnel 1.
Licensee shifts are 0800-1600, 1600-2400, and 0003-0800 hours. Shift inspectors will report one hour earlier (i.e. 0700,-1500, and 2300 hours0.0266 days <br />0.639 hours <br />0.0038 weeks <br />8.7515e-4 months <br />) to ' review plant status and. ongoing activities, witness licensee shift turnover, and shift; turnover with the offgoing NRC shift inspector.
2.
The 0800 daily' licensee planning and scheduling meeting will be! monitored each day by team. personnel as directed by the team leader.
- 3. i The by theteam leader wil1 receive a daily debriefing at the 0730 team meeting 2300-0830 shif th inspector, including any carry-over items from - the -i 1500-2433 shif t inspector. -In addition, they will interface daily with the 0700-1630 shift inspector during the 0700-1630 time period and rece'ive a final debriefing from team members at the close of business.
4.
Meal breaks may be taken at any time (the one-half hour time allocation will require bringing food or using the cafeteria / vending machines).
5.
Outside inquiries and contacts regarding the inspection are to be directed to tre team leader.
6.
Primary licensee contact during the 1500-2430 and 2300-0830 shifts will be the shift supervisor. The day shift inspector (0700-1630) can utilize any appropriate available member of licensee management.
Inspection / Technical-Guidance To Inspection personnel
- 1. Inspectors should maintain all relevant it crmation for significant events, persons interviewed, their title and any important facts or per-ceptions from such interviews, as well as specific activities observed
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(e.g. applicable procedure number, revision, and complete title). '
2.
Each inspector should pre plan the types of questions to be asked of licensee / contractor personnel to gather information that will allow him to address the inspection objectives and attributes. _
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Attachment C - Inspection Guidance 2 4
3.
In addition to inspecting ongoing operational, surveillance, maintenance, and other activities (windows of opportunity), each shift inspector should place special emphasis on maintenance activities with respect to observing pre-job briefing and/or any instructions given- to maintenance workers; supervisory coverage of maintenance activities; experience _ and training of j maintenance workers (that is, are maintenance tasks within " skills of the I craft");
are procedures technically adequate considering experience / I training / knowledge levels; a're procedures adhered to; are procedures developed for unique evolutions; are procedures properly changed if incor- ;
rect.; are systems ' restored to their correct alignment subsequent to the . j maintenance activity; and, is the control room aware of the activities and possible effects on plant operation. 1 f
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Individual inspectors may be requested to prepare other documentation dur-ing the course of the inspection such as: inputs to interim progress report (s); notes for the exit interviews; daily (morning) reports; etc.
- 5. Reference Inspection Procedures All areas: 71715 - Sustained Control Room and Plant Observations 42700 - Plant Procedures 40700 On-Site Review Committee !
40701 - Off-Site Review Committee 37703 - Tests and Experiments Program ;
39701 - Records Program 39702 - Document Control Program 40703 - Off-Site Support Staff
- Plant ,
Operations: 71707 - Operational Safety Verification 71710 - ESF System Walkdown 71711 - Plant Startup from Refueling Outage Maintenance: 62700 - Maintenance Program - Implementation 62702 - Maintenance Program 62704 - Instrument Maintenance 62705 - Electrical Maintenance Surveillance: 38701 - Procurement Control 38702 - Receipt, Storage and Handling of I Equipment Program 56700 - Calibration 61700 - Surveillance Procedures / Records 61725 - Surveillance Testing and Calibration Program I
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Attachment C - Inspection Guidance 3
-Training: 41400 - Training - Non-Licensed 41701 - Training - Licensed
' Assurance of Quality: 35701 - QA Program Annual Review- ..
35750 - QA Program Measuring and Test Equipment 36700 - Organization and Administration 40702 - Audit Program 40704 - Implementation - Audit Program Radiological Controls:
83000 - Series Security: 81000 - Series I
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n ATTACHMENT D Pilgrim IATI Tentative Schedule I. 6/6 Prep Week Onsite, as 'needed for individual preps and badging. As a minimum, team members should complete badging, arrange and receive a briefing from the resident ' inspectors, and attend licensee self-assessment presentations.
6/7 a.m. !
- 4 Licensee presentation on plant and program status, including results of BECo self-assessment and INP0 review.
6/7 p.m.
Detailed plant. tour in three groups (Warren, Lyash and Kim coordinate).
6/7 - 5:00-7:00 p.m.
Team meeting at either hotel or site. Adoittonal mee-ings will be held as.
needed throughout the inspection.
II. 6/13-6/24 - Inspection 1 '
6/13 - 1:00 p.m.
Entrance inte'rview onsite.
6/13 Begin specialist inspection (each inspector i s responsible for his per-sonal inspection sequencing and scheduling). Shift inspectors will have assignments of programmatic issues to evaluate when not in shift work.
6/14-15 Continue inspection. .
6/15 - 3:00 p.m.
Begin shift coverage.
l 6/16-17 Continue specialist and shift inspection.
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Attachment 0 . Ten.tative Schedule 2 6218'-8:00p.m.- f Decision will be maae whether enough has been learned to discontinue' shift inspections.
J 6/19
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Sunday 0ff (Optional).
6/20-22 Continue specialist inspection and . followup of shift inspector identified .
issues.
6/23 a.m.
Individuals organize / summarize.
. E 23 - 2:00 p.m.
Team Meeting.
6/24 - 8:00 a.m.
Exit Ory Run and Critque.
6/24 - 10:00 a.m. '
Wrap up - fill holes.
6/24 1:00 p.m.
Exit Interview III. 6/24-29 Complete Individual Documentation Onsite Each inspector will be required to provide report input, acceptable to the Team Leader, prior to being released. Inspectors may work the weekend to accomplish this or may take as much of 6/27-29 onsite as needed.
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l FIRE PROTECTION SYSTEM STATUS Date Ns..imb e r Outruta.nding Equipment MRs 1/a7- Approx. 200 10/87 42
. Total, Fire Watch Postings 1/87 40 (Non Barrier Related) 10/g7 29 Fire Barrier Repairs / Upgrades Total Scape 4162 Remaining 75:3
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DETAILS jcted on Edison Comoany (BECo) i'. _ P. Robi.rts, Nuclear Operating Manager
- R. A. Ledgett, Special Assistant to Sr. Vice President,' Nuclear
- P. J. Hamilton, Compliance Management, Group Leader
. *R. E. Grazio, Field Engineering Section Manager
- R. Wozniak, Fire Protection Group Leader
- W. M. Sullivan, Sr. Fire Protection Engineer
- R. Velez, Project Manager
- R. V. Fairbank, Licensing and Analysis Section Manager 1.2 Nuclear Reculatory Commission (NRC)
C J. Lyash, Resident Inspector T. J. Kim, Res'ident Inspector
- Denotes those present at exit interview.
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2.0 Followup of Previous Insoection Findinas
{ Closed)UnresolvedItem(86-36-01) Unimolemented Maintenance Work en
'Decraded Fire Protection Eouipment and Excessive Reliance on Fire Watches The NRC in a review of the Operability and Maintenance of fire protecticn (FP) l systems determined that the licensee did not diligently perform the maintenance and repair work on fire protection equipment. This assessment was made on the basis of a review of the maintenance request (MR) list which identified about 300 unimplemented maintenance requests on fire protection equipment. Some MRs were outstanding since 1983. The licensee in a management meeting with the NRC acknowledged that a problem with maintenance exists and committed to correct it. During this inspection, the inspector reviewed the maintenance request list and determined that only 37 MRs of relatively minor safety significance on fire protection equipment are still outstanding.
The inspector also randomly examined some of the maintenance work per-formed on various f 4 re protection systems and did not identify any unacceptatie cc-ditions. y m
N With regard to the use of fire watches, the licensee is still relyirg excessively on fire watches as an interim compensatory measure f 0r de-graded fire barriers. )
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L;censee Plo n o F Th e ih y 7: ,
Aeport t
-I ECD D. Mills Refer to " AREA CLOSURE PROGRAM" for Appx R Fire Penetr status Awaiting Constret Eng.g Paper Total Scoce Comolete To Go Releavd EnL Closure Repairs 3393 2775 618 232 106 280 Upgrades 769 628 141 83 39 19 Today Total ;d 168=- 315 145 299 Prior Total 4162 3384 778 Progress Delta 0 19 <19>
Note: " To go" includes items being reviewed for disposition as " accept as is" 8
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PILGRIM MARK-I CONTAINMENT ENHANCEMENT N. McE r i cia Februsry 1 57 l
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FCC f" : ESCFIFTION START /tENETN E6-51' .Directventfrcsthetorus'tothesainstack.Theventwill Mar 6 / 5 hksL use an existing penetration on the torus.for the standty .
gas _ treatment system (Eisi). An E-inch pipe will tap into h existing piping tetmeen too ecntainment isolation valves' and typass the SIGi. An additieral isolation vahe will
{ be it. stalled in the rew line. Nitrc;ea ' mill be ssglied to systea valves so that the vent path can te used during during a station blackcut. The vent path 4ill only te-h usedwithseniorsanagetenta;preval.
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- a Containrentsprayto::le' modification.SixofsevennozzlesL Mar 1 l'5 kks
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in each dryeell spray cluster will be plugged to reduce mater flow during an accident. This will prevent the creation of a contair.aent vacuus during acication of dry. ell spray. Flew to the rc::les will be red.u:ed iros
-five thcesand to one thcusand gallons per sinute.
E6- :b,c,d Fire water'intertie to the residual teat rencval (EE) tar i /.6 wls
, systee. The site fire maler systes >fil te tied to the (intartiel FMR s)ates via a rese.atle spol ;ie:e. ite spo:1 piece-
, aill not be installed during ncrial cperation. A second' rar 27 /-3 .ts diesel-driven fire pac; will te added to the currently (res fire pap!
irstalled diesel- and electric-driven fire purps to ;revide
' te] s:urces of containment spray that will te inda;ctdent -
"ay I l'2 als of'ensite s.t. p:*er. Fire water can also La irdected into the (fuel trrs, :d.'
hatter vessel via the FHR systes crcsstle. %draulically c;eisted hel transfer pue;s will be installed :n the two
. dieselfirepuys.Thefirematersysterissup;iiedmater fres two onsite' fire water tanks ard free a cunicipal mater supply. The crcsstie .ill allch a fire truck to j supply .ater to the RHR syster, a third asthod that is independent af onsite e.c, po.er.
?5-53 !ach:; nitro;en sugly f:t dry ell irstrutentatica and Kir 13 / ! wh
- artain ether critical ai*-operated cuponents. A ta:kup li pid nitr: gen st; ply e:l! te pla:ed cnsite en a traile .
httled nitregen will also be installed. The nitrc;en vill te suglied to tte instrue.ents via to c:*.tainment ]
attos;here dilutice (C?t} systaa. Mar;al typass vabes will te installed arcart the C4 ctataintint isclation nahes te safe the nitregen s;;;1y irde;erder.t of esite a.c. pc.er.
Es-!i A third diesel gir.eratcr. A ce* ren-0 disse! ;ererat:r mill rar :: / E his te c nstru:ted crtite. It will te at'e te te rar.ually tied nto either of the t.o safety tLses ir:: the ::ntrtl rc:1.
It ill have at:ut 7!t :! the ca;atity of cr.e cf the carre-t diesel ;enerat e s. It *;11 n:t re: ire a separete s:.rce of :cclin; mater a^d .ill te.e an irde;erdert se.en-day h el s;;;1y.
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FOC 8 ' DESCFIFTION STAF.T/LEABiH-L36-73 ~ Autctatic depressurization systeL (ADS) logic ramifications. LMar 6 / 4 ks Four sedifications will te aide to the ALS systee: (1) .
' sanual inhibit snitches will be installed for the 120 second-tieer, (2) ranual-bypass timers will be installed,to allow dapressurization en sustained Icw water level without high drynell pressure, (3) eanual typass ticers will be installed to allow a Ic. pressure ECCS put; to start without sustained low rea:ter pressere, and (4) a typass will be installed te allow depressurization without a low pressure ECCS pus; running.
36-75 Standty ligaid control modifications. The S-10 enrichtent . Mar 2 / 7 .ns~
. in the beren salutica in the standby liquid control systec ,
ill be increased. This wi!! double the capacity of the current systee ar.d elisinate the need for heat tracing.
57-03 !ackup d.c./a.c. pcner supply. A ron-D 3 0 VDC battery nith A:r 15 /.5 ute charger and inerter mill te installed to sapply critical
, aSO VAC a d 1:0 VAC cer;cnents during a stati:n blaclout.
The p.er supply ill te eanually initiated during a blackout. The tittery charger will be tied to t'.e diesel generator in the tachnical seport center.
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, -PROPOSED GPENING REMARKS-
- f. by Dr. Murley 4- ,
Pilgrim has been shut down since April 1986. Over the past several years, significant facility deficiencies have been identified through the inspection l process and were reported in the Pilgrim SALP. We note that BECo has devoted 1
E substantial resources toward resolving these deficiencies and welcome this opportunity for BECo to present the scope and status of its restart plan.
However, we recognize that considerable effort remains.
Although the staff has been quite active in its review and inspection
~
efforts at Pilgrim, we must satisfy ourselves that programs established by BEco 1 are not only adequate, but that these programs have been implemented and are effective. We feel that our reviews will take at least eight weeks from the time we have received a complete and comprehensive restart assessment package from BECo. The conduct of these reviews will be coordinated through a Pilgrim Restart Assessment Panel. Bill Russell will briefly sunmarize the canel's makeup and activities in a moment.
Several technical areas are of particular interest. As you are aware, the staff has found a number of your Safety Enhancement Program modifications acceptable; however thret are a number of unresolved questions regarding the Direct Torus Vent. Additionally the staff will review recent changes to your emergency operating procedures. We also recognize the efforts being made by BECo to resolve the emergency preparedness deficiencies identified by FEMA.
, Although consideration of an exemption to the biennial exercise may be premature, we look fonvard to understanding your progress in this area.
1 i
l NOTE: The following 5 pages provide additional information in support these remarks.
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. , NOTES FOR REMARKS BY DR. MURLEY g
- 1. Introduction Over the past few years the NRC has had a number of concerns with Pilgrim performance.
Pilgrim shutdown April 12, 1986 due to technical issues (MSIVs,
, containment isolation and intersystem leakage to RHR system)..
Shutdown confirmed by CAL 86-10, of 4/12/86.
Issues and NRC concerns in a number of areas have expanded and on August 27, 1986, we stated our need for a formal reassessment by BEco of their readiness for restart.
SALP Report of April 8,1987 identified significant deficiencies (Category "3") in areas of:
-Radiological Controls (was "3" previous SALP) i
-Surveillance
-Fire Protection -
-Security and Safeouards l
-Assurance of Quality l
BECo has devoted substantial resources towards the resolution of many of these issues and we view many actions, such as management changes, new programs, and in-plant improvements as positive steps towards making the Pilgrim facility ready for restart. I Still much to accomplish.
Areas of focus as BECo prepares for restart, include:
l -SALP and Technical Issues
-Management issues
-Safety Enhancement ' Program
-Emergency Preparedness Issues
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, Level of NRC staff activity on Pilgrim is high and will remain high in
, the next few months.
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- , - The level of Commission interest as well as level of public intarest
- in this facility is high.
II. Resolution of SALP and Technical Issues BECo has generally been optomistic regarding their schedule for.
dealing with many of the Pilgrim issues.
BECo must demonstrate substantial improvements and progress in areas of SALP "3" and resolve the technical issues identified by the staff as necessary for restart.
We will promptly look at technical issues and items when we have confidence BEco is ready for our inspections and review.
Staff must receive results of BECo's self assessment and have time to m6ke our independent evaluation.
III. Management issues BEco has made a number of management changes and instituted programs to address NRC concerns.
NRC must satisfy itself that management changes have brought substantial improvements and that new programs are effective.
A successful Diagnostic Inspection and satisfactory completion of the staff assessment of BECo readiness are key aspects of a staff decision regarding restart.
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j IV. Safety Enhancement Prooram (SEP) l i
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Staff evaluation of SEP (submitted 7/8/87) was transmitted to REco k August 21, 1987 b
R Staff found a number of the SEP modifications acceptable under f provisions of 10 CFR 50.59 or approved related Technical
, Specification changes. These include:
4 Containment Spray Nozzle Modification-Diesel Fire Pump Diesel- Fire Pump Fuel Oil Transfer System Blackout Diesel Generator ATWS Feedwater Pump Trip ATWS Recirculation PJmp Trip
. Enriched Baron to SSLCS (Tech. Spec.)
ADS Logic Modification (Tech. Spec.)
The generic issue of torus venting is still under evaluation. We look forward to your response to our questions in this area, but expect it may require extensive deliberation before we are prepared to make a decision on this issue.
As a related matter, we expect to expand our review of your Emergency Operating Procedures and Procedure Generation Program.
The staff has not yet approved Revision 4 to the Emergency Procedure Guidelines and we must assure ourselves that BECo work in this area is consistent with our approach.
1 4
, _ , , . _ _ _ _ _ _ _ _ . _ _ . - _ - - - - - - - - ^ - - ' - - - - - - " - ' - - - " ' ' - - - - - - - - - ' - - - - ' - - ' '
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.4 V. Emergency Preparedness Issues FEMA report was transmitted to BECo on August 18, 1987
?~' .
- FEMA concluded Massachusetts offsite radiological emergency planning _
and preparedness was inadequate.
t FEMA identified six issues:
- 1. Lack of. evacuation plants for public and private schools and i
daycare centers.
. 2. Lack of a reception center for people evacuating to the. north.
- 3. Lack of identifiable public shelters for the beach population.
- 4. Inadequate planning for the evacuation of the special needs population.
- 5. Inadequate planning for the evacuation of the transportation dependent population.
- 6. Overall lack of progress in planning and apparent diminution in emergency preparedness.
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r We have received your letters of September 18, 1987, and-the staff is reviewing your progress with State and local entities.
s, Consideration of any exemption may be premature at this time.
T BECo and the State must continue to work to resolve the issues identified by FEMA.
FEMA's assessment of progress towards resolution and the conduct of
{
the December exercise will need to be considered.
VI. Cnnelusion Staff will continue its aggressive review of BECo actions at Pilgrim.
Restart Panel will coordinate staff actions and keep NRC Senior Management appraised. It is a focal point for BECo to interface with NRC.
BECo must demonstrate that changes have been made and that management is successfully dealing with issues.
NRC cannot consider restart until all the issues raised during the facility's prolonged shutdown have been dealt with to our satisfaction.
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08f 21/8 Safety Evaluation, No. oll/8 Rev. o-Sheet I of ,fo SAFETY EVALUATION PILGRIM NUCLEAR P3WER STATION PDC PCN System Calc.
No.: Date:
Ir.it iator : Dept: Group: No.: Name:
W. Riggs Engr. Mech. 86-53 Backup M 650-1 Nitrogen Supply Description of Proposed change, test or experiment: See Attachment i 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.
BASIS FOR SAFETY EVALUATION CONCLUSIONS: See Attachment II Change Change (X ) Recommended ( ) Not Recommended SE Performed by P Date M 87 u v/ n Exhibi t 3.07.d (Snget 1 c' 3) Rev. ? /p
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Attachment 1 Safety Evaluation, Rev. 0 dE M*. UlO Sheet 4 of do A. DESCRIPTION OF PROPOSED CHANGE. TEST OR EXPERIMENT The desion change covered by the crocesed modifications is as follows:
A.1 Addition of a liquid nitrogen / vaporizer trailer (X-168), and associated piping and valves. The trailor 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 accomplished by connecting to the existing fill connection located on the north wall of the reactor building. /? TE#rEN'#8 M#
VAuve. wa. ss Pbvsveo As PMYW TME PWKMt 1MhM.
A.2 Addition of two banks of ten cylinders etch, 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 n:trogen supply is not available, until the new liquid nit. ogen / vaporizer trailer is available. The cylinders will delier nitrogen gas at 110 psig (controlled by a vendor supplici 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 val e. 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 M-l, (kq (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 nitrogan path to drywell instrumentation following a loss of power.
A.3 Modification of two existing Seismic Category I supports of the l Auxiliary Nitrogen Purge Supply (H9-1-11-19 and H9 1 12 1) as specified in Teledyne Engineering Services Document No. 6520 2.
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' Attachment I Safety Evaluation, Rev. 0 (E. do.,f//8 Sheet a of Ao
-B. PURPOSE OF THE CHANGE The safety' enhancement program requires that various plant.
modifications be implemented to insure the availability of those systems needed 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.
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Attachment II-Safety Evaluation, Rev. O S.E. M. ,Af/A Sheet 4 of .ts
'O. BASIS FOR SAFETY EVALUATION CONCLUSIONS 0.1 Systems Subsystems. Comoonents Affected D.l.1 Inertina and Drywell Testinc 0.1.1.1 This modification adds a check valve and a 2 inch tee in the existing cryogenic nitrogen supply header common to i the drywell instrument supply and drywell inerting supply.
The 2 inch tee connection branch includes a normally closed glibe valve and a normally capped pipe end.
0.1.1.2 This modification adds a pressure control valve to the existing pipe which connects the drywell instrument header with the existing drywell inerting piping which is normally isolated by an existing closed valve. ;
1 The pressure control valve reduces the 125 psig (maximum) I 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 station supply pressure. p.g g D.1.1.4 This modification locks closed valve J'H0162/in the instrument air supply header to drywell instrumentation, 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 / vaporizer trailer.
0.1.1.5 This modification adds a liquid 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 .ind to the drywell instrumentation supply header via newly installed 2 inch piping when the existing liquid nitrogen storage tank is not available.
0.1.1.6 This modification changes the air operator on drywell instrument supply isolation valve A0-4356 from fail closed to fafi open. This ensures an open path to the drywell during a station blackout event . Check valve 31-CK-167 located downstream of this valve performs the containment isolation function, i 0.1.1.7 This change modifies two hangers in the seismic portion of the drywell inerting header. This is a result of a new analysis performed on this pipe line.
7
-- ;= .. .
Attachment II Safety Evaluation, Rev. 0 (E. 06. All6
- Sheet s'of ato D.I.2 Reactor Buildino This modification involves drilling holes in safety (
related reinforced concrete of the reactor building.
D.1.3 Control Room Annunciator This modification involves adding an additional alarn input to the existing annunciator to provide control room indication upon switchover to the nitrogen cylinders.
D.2 Safety Functions of Affected Systems /Comoonents D.2.1 Inertino and Drywell Testino L D.2.1.1 The inerting and Drywell Testing System is part of the Containment Atmospheric Control System (CACS), which provides the capability 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 Drywell 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 Reactor Buildino 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 doses from a postulated design basis accident will be ;
maintained below the values of 10CFR100.
D.2.3 Control Room Annunciator The control room annunciator has no safety function. i l
1
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Attachment 11 l 3-Safety Evaluation, Rev. 0- 55. A), c,7/ /$ l Sheet 4, of .10 1 1
I h- D.3 Effect on Safety Function r
D.3.1 Inertina and Drywell Testina 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 l isolation functions of. the Inerting and Drywell Testing i 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 0
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 alarm input.
D.4 Analysis of Effects on Safety Functions D.4.1 Inertino and Drvwell Testina D.4.1.1 This modification does not alter the Inerting and Drywell Testing System's purging function because 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 Testing SysteWs containment oxyc.as dilution function because the liquid nitrogen / vaporizer trailer is normally isolated and is provided to supplement the existing nitrogen supply, when needed.
D.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 nitrogen cylir.ders are connected upstream of the existing containment isolation valves. Containment isolation valve function / operation is not altered by this modification.
D.4.1.4 The hanger modifications ensure the systems ability to withstand seismic and thermal loads.
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Attachment'I!
Safety Evaluation, Rev. 0 SE.Moetjl6 Sheet 7 of .2o D.4.2 Reactor Buildino (par. sacm tu CAcc.. M4so -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 postulated 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. .
capabili to . draw down secondary containment. THis ANat/8'S SAT 85f/f8 Ty p*#c ICAT'*" ', M.M. A $KCONDAM CoNTAINM E N=f ZseLAT/0&"ConTAME.D D.4.3 Controi oImTn#InIfa9##
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, /g The control room annunciator is not adversely affected by tF modification, as such no analysis is needed.
E. SURARY 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 supplies and associated piping, valves, and instruments.
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5 Safety Evaluation, No. clllB Rev. O Sheet / of ,te r LAFEv! EVALUATION PILGRIM NUC'_ EAR POWER STATION A. APPROVAL
()() 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).
( ) Comments:
The safety evaluation basis and conclusion is:
(4) Approved ( -) Not Approved
' CDvl ANf8]
Discipline Group leader /Date Supporting Discipline Group Leader /Date B. REVIEW / APPROVAL (r) Comments: b-
? wk Sp firoup (peader/Datd C. ORC REVIEW
( ) This proposed change involves an unreviewed safety question and a request for authorization of this change must be filed with the l Directorate of Licensing, NRC prior to implementation.
( his proposed ch ge dona npt involve an unreviewed safety question. j ORC Chairrnan _ Date I[6!#9 I
ORC Meeting Number P 7- V4 . l cc:
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l Exhibit 3.07 A (Sheet 2 of 3) Rev. 3 l l
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1 Safety Evaluation, No. ,1l16 Res. o i
. Sheet f of fta ] 1 PILGRIM STATION FSAR REVIEW SHEET
References:
Support a change List FSAR test, diagra?.5, and indices affected by this change and corresponding.
FSAR revision.
Affected FSAR Revision to affected FSAR Section is shown on:
Section- Preliminary Final Fia.10.11-1 GPCD) Attachment 1 /
Fia. 5.4-1 [P6fD) Attachment 2 Sec.16. ll,3.1 Attachment 3 Se,64-5l Attachment 4 Attachment 5 .)
Attachment 6 PRELIMINARY'FSAR REVISION (to be completed at time of Safety Evaluation preparation).
Prepared by: F /Date: Y f!87 Reviewed by:MN /Date: YW&
/ i'/
Approved by: ._ 11/ /Date: _4!'7h 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 completed FSAR Change Request form (Refer to NOP).
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Exhibit 3.07-A (Sheet 3 ef 3) Pev. 3
gg . f4 Of M Safety Evaluation l No.: J //A '
SAFETY EVALUATION WORK SHEET Rev. No. 4 j
A. System Structure Component Failure and Consequence Analyses.
SH. I of 3 \
System Structure Comonnent_ Failure Modes Effects of Failure Conssents GO PE Ws06. "DUE I EMR 0ggpL.
3 ,' PAM& EMAILcL ~ TEMP G L4GWD % VALVE fRIL5 Clo6EP valve. FAILS j PR ESs eR E TEG, j MEPPAMSSUA/2[ (DfNEESCOW.r_ RAL ' F '
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i seneral aeference Material Review CALCULATIONS REGULATORY FSAR DESIGN SPECS PROCEDURES GUIDES STANDARDS CODES SECTION PNPS TECHNICAL SPECS.
Am A &curri Ote- . 128 8b-d Sgg,JD,ll 3.7. A M -6.50 -f GE SIL # y'd,Z dsc.CJ/+C,A46~.5.&
r% 6~ A ~ / d' G E SPEC CASAIc2'20F*V o s ec . 4 9 . A.
Se $* Y ~l Me /D //-l B. For the proposed hardware change, identify the failure modes that are likely for the components consistent with FSAR assumptions. For each I fsilure mode, show the consequences to the systes, structures or related f components. Especially show how the f ailure(s) affects the assigned I safety basis (FSAR Text for each system) or plant u fety functions FSAR Chapter 14 and Appedix G).
W Date . N 7 Prepared by NOTE:
u' w It is a requirement to include this work sheet with the Safety Evaluation.
Exhibit 3.07-C Rev. 2 l
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SAFETY EVALUATION WORK SHEET Rev. No. 4 j con T*lu o A T*Iod ($H olh 3)
A. System Structure Component Failure and Consequence Analyses.
i System l Structure Component Failure Nodes Effects of Failure Comments Na 0ftAIDERS V A G !n d R E b . Ov8RPfessegi-no RELIE/= VA' LVE FAL S FI Pt % ofxNs A)o Ah AvgiiA&t p Legi inpesrix)s.
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General Reference Materia Review AND E9CCTED ou) A/476-CALCULATIONS REGULATORY /=44 4 4/Ar, FSAR SECTION PNPS TECHNICAL SPECS. DESIGN SPECS PROCEDURES 3y19J5 STANDARDS CODES l
- 8. For the proposed hardware change, identify the f ailure modes that are likely for the components consistent with FSAR assumptior.s. 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 .M Date I j // /
MOTE: It is a requirement to include this work sheet with the Safety Evaluation.
Exhibit 3.07-C Rev. 2
2 4
ANSAofs2o Safet,y Evaluation-No.: o2/18
- SAFETY EVALUATION WORK SHEET R*V- NO- O 60 f~ts.)a4 nori (SH.3f.:3)
A. System Structure Component Failure and Consequence Analyses.
System i
structure component Failure Modes Effects of Failure Coments 6dPPo f T~S bJPPeKrs B PF_ UAJSOPPstT50 $ d P fe.e D /9A'S 3 FA u SzisnicAL Lf J 26/6A/E?.O 77),
f3Cc c s D E Fr4/L UA'd-General Reference Material Review CALCULATIONS REGULATORY FSAR SECTION PNPS TECHNICA1. SPECS. DESIGN SPECS PROCEDURES GUIDES STANDARDS 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 6).
Prepared by #F Date (([
U' NOTE:
V G/
It is a requirement to include this work sheet with the Safety Evaluation.
Exhibit 3.01-C Rev. 2 l
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6N a of e1b PNPS-TSAR- Sf. Ala Al18 64'o 10CFR50.44(g) requires that the_ Containment Purge and
- Repressurization _ Systems (CPRS) 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 Station utilizes for primary containment ' atmospheric control as 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 and Drywell 'can be routed to the main stack via the redundant trains of the SGTS. Makeup. of nitrongen (or air) is supplied via the 1 in-i redundant solencid valve trains. See Figure 5.4-1. _ ggg } #j,9,e Oxygen concentration. is controlled below flammability limits (5 volume percent) by a feed and bleed method (purge method as defined in 10CTR50.44). The time required before initiation of, purge (vent) of the primary containment is controlled by repressurization 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.
To meet the above requirements, 16 solenoid valves are arranged to i provide redundant paths to and from the drywell and torus for N2 U makeup /repressurization and venting. N2 makeup /repressurization is provided by:
connections outside containment, a
- 1. Connecting, to hose portable nitrogen supply via truck with vaporizer 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 (gasoline driven) air compressors located on site (secondary ettergency make-up) ;
1 The solenoid valver are designed to remain closed against maximur containment pressure. to vent containment so that the maximur containment pressure will not be exceeded, and to provide a r.itrogen l flow sufficient to maintain the hydrogen concentration inside !
containment below the flammability limits. l
, 1 I
t 5.4-2 Revision 5 - July 1985 I.
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I t-L A.1 4 6 u if A liquid nitrogen / vaporizer trailer (X-168), and associated piping and valvgg. 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 ,
6 (via an on-site truck cao) 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 I 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 A1: m" n6wo banks of ten cylinders each, a cylinsier 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 j switchover to the cylinders. The existing manual gate valve,S 31-l, (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 AO'4355 will be modified from fail closed to fail open to maf ntain an open nitrogen path to drywell instrumentation following a loss of power.
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J 10.'11; INSTRUMENT AND SERVICE AIR SYSTDIS g
10.11.1 Power Generation.0bjective' The power generation objective 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 psig. 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 o
The air systems' are, in general, designed to Class II requirements, 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 h2gh 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 s in parallel downstream of the instrument air dryer X-105A and one filter downstream of dryer X-105B. The dcwnstream air filters are to ensure that no dessicant or other foreign mate ri,11 anters the instrument air system. There is also a bypass around the dryers and filters which can be opened by remote manual means for dryer X-105A and manual only for X-105B to assure a continued supply of instrument air to the essential instrument air header in the event of an air dryer failure. Normally, use of the two rotary cortpressors will maintain the air- receivers at the desired pressure for system supply. The remain 2ng compressors serve ens standby units. Actuation of the standby units is automatic and is indicated in the contro!
room.
The low pressure air supply (nominal 20 psig) is developed by two
- centrifugal air blowers. The blowers discharge for distribution 10.11-1 Revision 4 - July 1984
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' through a moisture separator _and a mist eliminator. Blower usage is .
intermittent. No dowpoint control is provided. .,
I p 'A 'normally closed pressure reducing cross-over line -is provided-i 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.
L Pressure loss in the high pressure system, sensed by several pressure
'*, switches, will .cause valves in the service air header, .the -low pressure. service- air: cross-around line, and the non-essential ~
instrument air header to close in a cascading. sequence thus 1eaving the essential instrument air _ header as the only header drawing air from the receivers in the event that supply pressure decreases.
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!/s ugl' 10.11.3.2 Equipment Description Compressors The three reciprocating' air . compressors are vert 2 cal, single stage, compressors. They are each rated to double acting 159.5 reciprocating standard ~ '/
ft min at 105 psig. The two- rotary deliver 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 range.
Each of the three reciprocating compressors is belt-driven (4 ' belts) f j
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 cowpressor discharge pressure of 102 psig. The compressor speed as 3,550 rpm.
Aftercoolers The compressor aftercoolers are shell and tube counter current coolers wi;h air passing through the tubes and water flowing around the t.tes. They have an integral moisture separator equipped with an automatic drain trap to remove condensed moisture from the cooled ear. Cooling water is supplied by the Turbine Building Closed Cooling Water System.
N 10.11-2 Revision 4 - July 1984 l
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n.e - Go 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 1 deliver nitrogen gas at 110 psig (controlled by a vendor I supplied pressure regulator), through 2 inch piping which will i tie into the existing drywell instrument supply header. The j new piping will contain a check valve, a gate valve and a i relief valve. A differential pressure indication switch with j 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 Ji-1, (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 l maintain an open nitrogen path to drywell instrumentation following a loss of power, I
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RIf SAFETY EVALUATION PILGRIM NUCLEAR POWER STATION Rev. No. O PDC /PCN System Calc.
Initiator: Dept: Group: No.: Name No. Date: 6 JJ/g7 l
g,g,.g NED FS&HC FRN 86-53 Back-up N2 Supply Description of Procosed chanae. test or exoeriment: Addition of Pressure Indication of Nitrocen Sucolv to Drvwell Eauioment.
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 re' duce the margin of safety as defined in the basis for any technical specification.
BASIS FOR SAFETY EVALUATION CONCLUSIONS:
See Attachment 1 Change
()O Recommended ( ) Not Recommended SE Performed by h af Date f 7o? /)
V Exhtbit 3.D7-A
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Sheet 1 of 3 -
3.07-13 Rev. 4 i
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No:
$_AFETY EVALUATION
_P_ILGRIM NUCLEAR POWER STATION Rev. No. O A. /PPROVAL This proposed change does not involve a change in the Technical (X)
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
'50.7)(e) and is reportable under 10CFR50.59(b).
() Comments:
The safety evaluation basis and conclusion is:
(X) Approved () Not Approved
$X -Tv th*lt?
& K W2z/n Discipline Group Leader /Date Supporting Discipline' Group Leader /Date B. REVIEH/ APPROVAL Comments: Mw AW- O N 58(Sh GroQp Leader /Date 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. i l
(X) This proposed change does not involve an unreviewed safety question.
Date ORC Chairman ORC Meeting Number Cc:
Exhibit 3.07-A Sheet 2 of 3 Rev. 4 3.07-14 fi.M i !. l1
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PILGRIM STATION FSAR REVIEW SHEET
References:
Safety Evaluation: Rev. No.: Date:
Support a change List FSAR test, diagrams, and indices affected by this change *and corresponding FSAR revision.
Affected FSAR Section Revision to affected FSAR Section is shown on:
Preliminary Final Fia. 10.11-1 Attachment ~2. X Attachments Attachment 3 Attachment 4 i
Attachment 5 Attachment 6 PRELIMINARY FSAR REVISION (to be completed at time of Safety Evaluation Preparation).
Prepared by: -
44___ /Date: 4 lUP% /
Reviewed by: l-F.} (/ / /Date:
Approved by:
o 0
o
/Date: 6!2h FINAL FSAR REVISION (Prepared following operational turnover of related systems structures of components for use at PNPS) (1)
Prepared by: /Date: Reviewed by: /Date:
Approved by: /Date:
Exhibit 3.07-A
- Sheet 3 of 3 3.07-15 Rev. 4 1.. .
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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . - - - - -- - - - - - - - - - - - - ' ~ ~ ~ ~ ~' ~ ~~
0 SAFETY EVALUATION WORK SHEET Rev. No. O A. System / Structure / Component Failure and Consequence Analyses.
)
System /
Structure /Comoonent Failure Modes Effects of Failure Comments Any & All Components Loss of/or Wrong Loss of/or wrong Hi/lo Alarm L of added Inst. Loco Sianal Indication on C7 Available:
NSR, indication; g4/t[t'/
Local Press 7
Indication Available:
Hi/lo Alarm;
- 2. Nitrocen Sucolv Ruoture/leakaae loss of oressure Local Pressure .
l Tubing Indication '
Available: System q is ItSR, '
wf6/efr7
- 3. New Hirina in Ooen/Short *Affects one Division in C7-redundency Panel C7 Circuit available.
Ground Faults
- Has no affect on C904 _ isolated by fuse.
Existing wiring in "
Failure will not Flexible C904. 2
-D"cr:::nt to C7 conduit for installation
)% g gry in panel C904.
.neral Reference Material Review FSAR CALCULATIONS REGULATORY SECTION PNPS TECHNICAL SPECS. DESIGN SPECS / PROCEDURES GUIDES / STANDARDS / CODES Sechtel Calc.
10.11 650-C300.0 Rev. O Rec. Guide 1.75 Bechtel II/I Cales.
17322-650-6200.0 B. For the proposed hardware change, identify the failure modes that are likely for the components consistent with FSAR assumptions.
l 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 8 Date !#'9 f./ / / '
NOTE: It is a req ent to include this work sheet with the Safety Evaluation.
Exhibit 3.07-C 3.07-18 Rev. 4 i k1
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- 6. Ale 1 Sheet 1 of 8 Attachment 1 A. DESCRIPTION OF CHANGE An analog pressure indicating loop / will be added consisting of a pressure transmitter, pressure indicator and a power supply. A pressure transmitter will be installed to monitor the drywell instrumentation nitrogen pressure, and will provide an. analog, 4-20 ma signal to a remote electronic pressure indicator (0-200 PSIG) installed on main control room panel C7.
The pressure transmitter will be tapped off an existing pressure sensing line currently connected to local pressure gauge PI4349. Power '
for the pressure analog signal circuit will be from Y2 via a 24VDC power supply to be mounted in the bottom of panel C7.
The added instrument loop has a non-safety related function; however since Y2 is backed by DC, this instrumentation will be available during station blackout conditions 'SBO).
B. PURPOSE OF CHANGE The addition of a backup nitrogen supply for the drywell equipment provided by PDC 86-53l necessitates a need for the control room personnel to know the status of the nitrogen system pressure. This is particuladynecessary during SB0/SA conditions to determine when the truck supply must be made available.
The pressure indicating loop will provide the control room operators information on the availability of nitrogen for the drywell instrumentation.
C. SYSTEMS. SUBSYSTEMS. COMPONENTS AFFECTED This modification adds:
- 1. Nitrogen Instrument System
- One pressure transmitter PT 4348. The pressure transmitter will be tapped off an existing pressure sensing l'ine currently connected to local pressure gauge PI4349 and local pressure switch PS4354. Both existing instruments will remain intact and operational.
- 2. Main Control Room Panels .
- Two Instruments on Panel C7 as follows:
a) One instrument power supply E/S 4348 with associated wiring.
b) Onepressureindicatorg474gwithassociatedwiring. ,5/rp
- 3. 120 VAC power (Vital Bus - Y2) !
- One power supply circuit in Panel C904. The circuit consists )
of wiring with a fuse that forms a boundary between added o;7- y- -
associated circuitsj A and Kuoc.c4gss /s/An,e,,rg ,c//pj d[i]hy j g : ;4 j
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Sheet'2 of 7 3 Attachment 1 D. SAFETY FUNCTION OF AFFECTED SYSTEMS / COMPONENTS 1- Th'e existing pressure instruments PI 4349, PS4354, and the added pressure transmitter PT4348 haWno safety function. g4fn/H
- 2. -
The two instruments added to Panel C7 have no safety function, however, they are located in the safety related. Panel C7. The panel contains controls and instruments of the Containment Ventilation, Isolation and Gas Treatment Systems.
- 3. -
The added power supply circuit has no safety related function, however, the circuit is in the safety related panels C904
& C7. The C904 panel cont ns controls and instruments of the RHCU System.
b arra [kr/
- 4. Indicator PI5031 has no safety related function.
E. EFFECTS ON SAFETY FUNCTION
- 1. -
The added primary sensor (pressure-transmitter) is connected to instrument tubing that serves two other sensors (PS4354 and PI4349). None of the instruments have safety function.
- 2. -
The added non-iafety related circuits from power supply (E5 4348) to the Pressure transmitter has been designated as associated circuits with division "B". This ensures, that any postulated failure of the new wiring or components located in control Panel C7 could affect only one division (i.e. division B), since the wiring will be run-with division B. In addition, this wiring will notaffect wiring in control panel C904 due to 3.
the protection provided by the fuse locatgpanel C904. gp,p The added wiring inside the panel C904 is associated with the division "A" due to a lack of physical separation of existing power supply for that division.
To facilitate extending this circuit M control panel C7, a fuse and Flexible / rigid conduit has been inscalled so that isolation exists between the new circuit and the existing circuit. The isolation is required from the fuse to the Power Supply (ES4348). This extension is classifed Non-Q from the fuse to the power supply (ES4348).
Flexible metal conduit will be used to provid'e separation of the new wiring downstream of the fuse to where the wiring exits control panel C904 and connects to the rigid metal conduit which runs to control panel C7. Flexible conduit will again be used in Jemb0 panel C-7 from the connection at the rigid metal conduit to power supply (E54348). The fuse has been designated as "Q" and the cable between control panel C904 and control panel C7 has been designated as M/m -q ', p i' r7
- 4. Theengr added instrument loop is powered by a..ngn-IA.CJ ,
g poweN fr power panel Y2.
p to class IE source is avoided. Thus a direct electr
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a :,o ow ..c No. 2 He9 Sheet 3 of #.5
-F. ANALYSIS OF EFFECT ON SAFETY FUNCTIONS
- 1. The added instrumentation loop has no control _ function..and no !
safety related monitoring function. The pressure transmitter,- i pressure indicator and power supply are non-safety related. The I instrument tubing and fittings are non-safety related. The l addition of these components in accordance with existing BECo Specification and vendor documents supplied with the equipment does not adversely affect the integrity of the existing system.
As delineated in the above paragraphs, the implementation of this modification interfaces structrully and/or by physical proximity, with safety related equipment or with class IE electrical power.
The structural interface is addressed in the design as follows:
- a. The installation of the pressure indicator in the control room panel is verified to be acceptable for II/I considerations per Bechtel calculation 650-C300.0 Rev. O, titled " Evaluation of Indicator to be Mounted in Panel C7 of II/I Criteria."
- b. The conduit installation for the cable run from control panel C904 to control panel C7 ed -T-rom phael 41 to rht G/d fFtWaiffer
' ' ~~
os to .Zl7.T Fri ft'rld. ' ~ _
f
- c. The installations of the power supply in the control room panel and the pressure' transmitter in the reactor building are verfified to be acceptable for II/I considerations per Bechtel calculation 650-C200.0 Rev. O, titled "Hounting Details for Instruments ES-4348 and PT-4348, PDC 86-53".
The blockwall calculations and the mounting calculations are in Bechtel Calc. 17322-650-C200.0.
Due to lack of physical separation, the non-safety related wir' of this modification is associated with safety Division "A" in Panel C9 , and f7 with safety division "B" in Panel C7. As described in previous paragraphs, the added modification is engineered to ensure that it does ;
not affect a single failure criteria of installed systems, g) yO This installation itself is non-safety related, it does not, and is not intended to meet the single failure criteria. !
I G.
SUMMARY
p The above descriptio@n be performed and analysis per tht FRN, does notdemonstrates degrade any ofthat thethe modification systems affectedto 5 by this installation, notably the Containment Ventilation, Isolation Gas Treatment Systems, RHCU, and 120 VAC Power Systems.
It does not increase the probability of accident occurance or malfunction of equipment important for Safety, as previously evaluated in the FSAR. It does not create the possiblity for an accident or malfunction of a different type than any evaluated in the FSAR. It does not reduce the margin of safety. Therefore, this modification does not I involve an unreviewed safety question. --
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' Safety Evaluation No.: FlOlp SAFETY EVALUATION PILGRIN NUCLEAR POWER STATION Rev. Nc. O PDC PCN System Calc.
Initiator: Dept: Group: No.: Name: No.: Date: Qyy MEo uSgA m esee Su-b).(. 5f6 c/s- G w
- e = s.v.t-C S&f) 5. o 7- c.
test or experiment: "Th PrepM Cb-f>
Description of Proposed changege a. m SE P che ses qa .casa +,r . -wa:pae 13;-e3 $
pm,b g.tr 4 etIit;eS bu'i*Et Oi 0i"Q Vak'*V.!*-n.c::4 ud% A*x, achvd.es% Clude b ' ed Fud oI %i%=-
.eyca ucJim -Seu _ J a.fm , 4%tch DI f o ro+ed t.% % Ia bs s h and. ,
.:w . TL. orose sed c ' '
oa vis q , carbs c::A crashed s h e & ctnend CoE., M ~.Lr Prve s.~. Y
'is .4-, spe c:1. M .44 a. u h a m m c 5 M L /sht w tel Ppc TL -54; A MXn-A*veS. \
prr L dacG2 c) e Sca%w oS + Cue C l SAFETY EVALUATION CONCLUSIQll1:
The proposed change, test or experiment:
j
- 1. ()() Does Not ( ) Does increase the probability of occurrence or l consequences of an accident or malfunction of equipment important to I safety previously evaluated in the ,FSAR.
- 2. W) Does Not ( ) Does increase the possibility for accident or malfunction of a'difforent type than any evaluated previously in the FSAR.
- 3. (X) Does Not ( ) Does decrease the margin of safety as defined in the basis for any technical specification.
BASIS FOR SAFETY EVALUATION CONCLUSIONS:
Su. Mfto e h wse nT &
Change Change
( ) Not Recosseended Q() Recomended SE Performed by _ bb . _ _ _
Date 7 i
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l Safety Evaluation No.: L.lO(p ~!
SAFETY EVALUATION .
PILGRIM NUCf. EAR POWER STATION Rev. No. O PDC PCN System Calc.
Dept: Group: No.: Name: No.: Date: t{. - $ f - 91 Initiator:
ueo c/g sg-Su o.esa sea (at..Sf6 Gwk e r9.H.s.t-C S&f) o 7-C.
~7 6 - r>roF W J C b re- 1 Description of Proposed changeg-test or experiment: i p ew,% G.f< O c.'lo tus . a. te.a 5 E P ch esex r uv & . ~T% dvaccl.k3 $ .
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SAFETY EVALUATION CONCLUSIONS: i The' proposed change, test or experiment:
- 1. (x) Does Not ( ) Does increase thw probability of occurrence or consequences of an accident or malfunction cf equipment important to safety previously evaluated in the fSAR.
- 2. (g) Does Not ( ) Does increase the possibility for accident or malfunction of a difforent type than any evaluated previously in the-FSAR.
- 3. (X) Does Not ( ) Does decrease the margin of safety as defined in the basis for any tecnnical specification.
BASIS FOR SAFETY EVALUATION CONCLUSIONS:
%u, Mfto e h wse n~f* A Change Change
( ) Not Recommended Q() Recomended N SE Performed by b f' b Date 7 l , . c- . .
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SAFETY EVALUATION No.: () pop g
.c PitGRIM NUCLEAR POWER STATION
- Rev. No. ()C A. APPROVAL g
X) This proposed change does not involve a change in the Technical.
Specifications. -
(X) .]
This proposed change, test or experiment does ( ) does not K) involve an unreviewed saf ety question as defin.J in.10CFR, Part '!
.50.59(a)(2). i
. (%) This proposed change involves a change to the FSAR per 10CFR 50.7)(e) and is reportable under 10CFR50.59(b). l l
(X) Comments: f9C Bf-flA acA.& C.wl ad s,fe cart $'t. W
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'M &ctwM aa n.ucun.G The safety evaluation basis and conclusion is:cw161. h &
(X) Approved () Not Approved
/ /AM $]*87 Discipli
~oug lead te Supporting Discipline Group Leader /Cate
- 8. REY APPROVAL
) Comments: MN -
%% lpm 4MG V i QA GroAp Leader /Date C. ORC REYlEW i-() 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.
(
his proposed change does not involve an unreviewed safety question.
, j ORC Chairman (LUk.) Date '1b 1/6'?
ORC Meeting Number 7 7-W cc:
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- ()() ~ Specifications.
This proposed change does not involve a change in the Technical (X)
This proposed change, test or experiment does ( ) does not M) 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 50.7)(e) and is reportable under 10CFR50.59(b).
(X) Conments:
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& -C,,4h Sl+ d.<stf ~T-C.a <ktctr. cJ w -nv c h a n.Q The safety evaluation basis and conclusion is:W i3 W0 " & fOC.FL-ST,8 (X) Approved () Not Approved
/ /M h]~87 Discipli ou te
~ [ Lead Supporting Discipline Group Leader /Date
- 8. REV APPROVAL
) Conments: MN
% L b n 4I5 5 l/,(@AGrejpLeader/0 ate 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. i
(
his proposed change does not involve an unreviewed safety question.
$ 8 j ORC M e g Number F '7 cc:
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References:
Safety Evaluation: P 10 (/ Rev. No.: O Date: _M - S e ']
Support a change List FSAR ' test, diagrams, and indices af fected by this change and
, corresponding FSAR revision.
Affected FSAR Section Revision to affected FSAR Section is shown on:
Preliminary Final Sectim 1. 6 - Fga l 6-l Attachment 1 X (p g, g,_z R+G+c96,^$ C Ofl, pDCL 1k~ a El Sitepl4,,) Attachment 2 O P"f*M M-T Attachment 3 -
Attachment 4 r
Attachment 5
,1 Attachment 6 '
PRELIMINARY FSAR REVISION (to be completed at time of Safety Evaluation preparation). .
repared by: Cd6 S /Date:
,, Reviewed by /88/Date: /8d7 Approved by:N 4t s /Date: I N /(/ i U $
O FINAL FSAR REY 1SION (Prepared following operational turnover of related Systems structures of components for use at PNPS). (1) :
Prepared by: /Date: Reviewed by: I
/Date: j 1
(1) Attach completed FSAR Change Request Form (Refer to NOP).
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Safety Evaluation: > 10 (/ Rev No.:
Support a change O Date:
_f-S-e]
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 6 A"' b by Fg l 6-1 attachment 3 .
y pyc t, 3g, a 3; tePlan) Attachment 2 O P"f'*M A*"f Attachment 3 Attacf w ; 4 Attachment 5
.! j Attachment 6 PRELIMINARY FSAR REVISION (to be completed at time of . Safety Evaluation preparation). .
repared by 016 $
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Approved by:M M M w /Date:
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FINAL FSAR REVISION (Prepared following operational turnover of related .
systems structures of components for use at PNPS), (1)
Prepared by: /Date: Reviewed by:
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SAFETY EVALUATION WORK SHEET Rev. No. O l .
A. System Structure Component Failure and Consequence fr.alyses.
[ System t Structure Component Failure Nodes Effects of Feilure Conenents
- 1. Ofci+P idb l PN N h (sea Ew h;b.t 3.o7-2 2.
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General Reference Material Review FSAR CALCULATIONS REGULATORY SECTION PNPS TECHNICAL SPECS. DESIGN SPECS PROCEDURES GUIDES STANDARDS CODES
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- 8. For the proposed hardware change, identify the failure modes that are likely for the compor.ents consistent with FSAR assumptions. For each failure mode, show the consequences to the systen, 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 6).
Prepared by Date 3[3' NOTE: It is a requirement to include this work sheet with the Safety Evaluation. I Exhibit 3.0 W Rev. 2 h .
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System Structure Component Failure Nodet, Effects of Failure Comnents
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- 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 6).
Prepared by _ 24 Date 3b~'
NOTE: It is a requirement to include this work sheet with the Safety Evaluation.
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Attachment A to Safety Evaluation # PIDV gLt a <> :.
6 and PDC 86-56A 1
.t A. DESCRIPTION OF CHANGE This change describes the preparation activities required to prepare the site yard area for the installation of a new diesel generator facilities:
The preparation activities include the following:
- . (a) Excavation for the fuel oil tanks, the halon tank (s)/ neutral resistor foundation, the switchgear foundation, the diesel generator foundation and radiator foundation.
(b) Compact subgrade below diesel generator. !
(c) Placement of reinforcement and concrete for the foundations. l (d) Installation of the fuel oil storage tanks and associated piping.
(e) Installation of concrete duct banks for electrical cable.
(f) Installation of grounding for diesel generator, neutral grounding resistor, switchgear and duct bank. '
(g) Backfilling the fuel oil tanks with pea gravel.
(h) Structural backfill under diesel generator.
(i) General backfill and compaction.
(j) Placement of reinforcement and concrete for protective slabs above the fuel oil tanks, q i
(k) Installation of cathodic protection for all underground piping.
]
(1) Installation of paving, curbs, finishing grading and installation of crushed stone and landscaping stone ground cover. ,
1 The final result will be a facility ready to receive the diesel generator I mechanical and electrical components, j l
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Attachment A to Safety Evaluation # vic k gLt: 0:
I b and PDC 86-56A A. DESCRIPTION OF CHANGE 2 This change describes the preparation activities required to prepare the site yard area for the installation of a new diesel generator facilities:
The preparation activities include the following:
(a) Excavation for the fuel oil tanks, the halon tank (s)/ neutral resistor foundation, the switchgear foundation, the diesel generator foundation and radiator foundation.
(b) Compact subgrade below diesel generator.
(c) Placement of reinforcement and concrete for the foundations.
(d) Installation of the fuel oil storage tanks and associated piping.
(e) Installation of concrete duct banks for electrical cable. -
(f) Installation of grounding for diesel generator, neutral grounding resistor, switchgear and duct bank.
(g) Backfilling the fuel oil tanks with pea gravel.
(h) Structural backfill under diesel generator.
(i) General backfill and compaction.
(j) Placement of reinforcement and concrete for protective slabs above the fuel oil tanks.
(k) Installation of cathodic protection for all underground piping.
(1) Installation of paving, curbs, finishing grading and installation of crushed stone and landscaping stone ground cover.
The final result will be a facility ready to receive the diesel generator mechanical and electrical components.
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&acSw! A $6 VW gist 2.cb L B. Puroose of the Change The purpose of this modification is to prepare the site yard area located south of the Turbine Building and adjacent to the Relay House for the installation of a new diesel generator set. This new diesel generator set will provide an additional source of back-up electrical power and reduce 4 the probability of a complete station blackout.
C. System. Subsystem. Comoonents Affected The changes implemented under thir, PDC package have no impact on any existing safety-related buried commodities nor do they modify any existing safety-related electrical / mechanical systems and/or civil structures.
D. Safety Function of Affected Systems Since this PDC affects no existing plant systems it has no effect on any safety functions.
E&F Effect on Safety Functions /and Analysis See D,above G. Summary This modification describes the preparation of a yard site for the '
installation of a new diesel generator. Since these activities do not affect any existing plant systems, it has no effect on any safety related system of safety function This modification.does not involve an unreviewed safety question.
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B. Purcose of the Chance The purpose of this modification is to prepare the site yard area located south of the Turbine Building and adjacent to the Relay House for the installation of a new diesel generator set. This new diesel generator set will provide an additional source of back-up electrical power and reduce ;
l the probability of a complete station blackout. '
C. System. Subsystem. Comoonents Affected l
The changes implemented under this PDC package have no impact on any l existing safety-related buried commodities nor do they modify any existing l~ safety-related electrical / mechanical systems and/or civil structures.
D. Safety Function of Affected Systems Since this PDC affects no existing plant systems it has no effect on any safety functions.
E&F Effect on Safety Functions /and Analysis See D,above G. Summary This modification describes the preparation of a yard site for the installation of a new diesel generator. Since these activities do not affect any existing plant systems, it has no effect on any safety related system of safety function This modification.does not involve an unreviewed safety question.
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No. ,;t l24 Sheet i off SAFETYEVdLUATION PILGRIM NUCLEAR POWER STATION
Initiator: :: Dept: Group: No.: Name: No.:
i L.R.' Namer Date: l NED S&SA 86-73 Automatic 3/2/87 D. Gerlits Depressurization ,
System
)
Description of Proposed change, test or experiment: Addition of high drywell 4 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 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.
BASIS FOR SAFETY. EVALUATION CONCLUSIONS: See attachment 1 Change Change (X) Reconnended ( ) Not Reconsnended Q.
SE Performed byQL. . Namer / . GerlitsMKMe Date ! j /Y/
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PILGRIM NUCLEAR POWER STATION-j A. APPROVAL j n
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.4' ( ) This')roposed change does--not; involve.a change in the Technical j Specifications.
(X) This proposed change, test or experiment does ( ) does not (X) involve an unreviewed safety question ;as defined in 10CFR, Part b c 50.59(a)(2).
' 1 (X).This proposed change involves a-change to the FSAR per 10CFR 50.71(e)- {
and;is reportable under 10CFR 50.59(b).
( ) Connaents:
The safety evaluation basis and conclusion is:
(X)' Approved () Not Approved- )
dWk kciplineGroupLeader/Datr
- 4/7 /o w
. Supporting Discipline Group Leader /Date B. REVIEW /APPROV AL
( ) Conraents:
abfS&SA Group Leader /Date C. ORC REVIEW-
-( ) This proposed change involves an unreviewed safety question and a request for authorization of this change must be filed with the i Directorate of Licensing, NRC prior to implementation. l
( ) This proposed change does not involve an unreviewed safety question.
ORC Chairman Date ORC Meeting Number C C ',
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References:
Support a change l
List FSAR test, diagrams, and indices affected by this change and corresponding FSAR revision.
E Affected FSAR Section Revision to affected FSAR Section is shown on:
j _ Preliminary Final
- . N/A a.tt
- :h.;;at i 4.4.5 Attachment 2 y 7.4.3.3 Attachment 3 V Figures 7.4-6,7,9,11; Attachment 4 6 7.3.6 Att:dser,t 5 Att::bant 5 PRELIMINARY FSAR REVISION (to be coupleted at time of Safety Evaluation preparation).
Jh. Id Prepared by: L.R. Namer / Date: h Reviewed by: h D. Gerlits
/Date: M27/P7
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Approved by: $b. /Date: Y/rf/P7
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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 completed FSAR Change Request Form (Refer to NOP).
(
Exhibit 3.07-A Rev. 2 (Sheet 3 of 3)
g y ; . -- 3,7;7 77 v c5 f ';;,L .., . ;- . y . 3, . - %er Safety Evaluation No. 2lM 1 Sheet 4 of t; SAFETY EVALUATION WORK 5 HEEL A. System / Structure / Component Failure and Consequence Analyses.
Systhm/ ' Failure Effects of q Structure / Component Modes Failure Comments a
- 1. High Drywell Pressure Timer Contact No ADS Channel Redundant Bypass Timer Logic Closure Failure Initiation Channels plus Signal Hanual Capability f
Available 1
(
- 2. ADS Inhibit Switch Switch Contact ADS continuous Redundant channel In Normal Position Open channel inhibit plus manual ADS initiation -
available
- 3. Low Reactor Pressure Timer Contact No RHR or Core Redundant channels Bypass Timer Logic Closure Failure Spray initia- plus manual capa-tion signal bility available General Reference Material Review FSAR CALCULATIONS / REGULATORY SECTION PNPS TECHNICAL SPECS. DESIGN SPECS / PROCEDURES GUIDES / STANDARDS / CODES 4.4.5 Sheet 49 GE Document NUREG 0737 II. K.3.18 (Table 3.2.B) 24A1719, Rev. 0 7.4.3.3 Sheet 73b- NE D0-24951 NUREG 0700 7.4.3.4 Sheet 73 (Sec. 3.2) GE Report Sheet 73a EAS 154-1286 (Analysis) 7.4.3.5 Sheet 109 (Sec. 3.5.E) 505-C200.0 670-C100.1 670-C100.2 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).
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Prepared by L.R. Name D. Gerlits Date b) kh NOTE: It is a requirement to include this work sheet with the Safety Evaluation.
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' Attachment I- No. M24 Sheet 1 of 10 Bases for Safety Evaluation Conclusions
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1 A. Description of Change: J The proposed change to the Auto Depressurization System contai.ns ,
these new features: 1 a). allowing initiation of automatic vessel depressurization on- !
sustained low RPV water level for 11 min. (nominal . setting of the new bypass timers). This supplements present logic by expanding '
p 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 geaucing the need for operator-actiorh
.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 i 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 capability of the ADS, allowing auto initiation, if required, for events such as breaks external to the drywell or a stuck open SRV. <
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A new 11 minute timer.-incorporated into the. logic system' design '.is
- initiated on the same low' water level trip signal currently used for ADS. An alarm alsrts the operator:that the bypass timer has been activated. After the set. time delay, the relay contacts for the. ADS high drywe11' 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' '
sec timer times out, the logic system' confirms tha_t 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 DQ1 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 305 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 Automatic Depressurization System (ADS) actuation logic should be modified to eliminate the need for manual actuation to assure adequate core cooling. A feasibility and risk assessment was required to determine the optimum approach. One possible scheme which was considered is ADS actuation on low reactor vessel 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 BHR Owners Group evaluation of j
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g Safety Evaluation Attachment 1 No. 2124 Sheet 3 of 10 this resulted in GE report NEDE-30045 (February 1983). The NRC has accepted either addition of a bypass to the drywell pressure trip plus a manual inhibit (Option 4 on NEDE-30045), or elimination of the high drywell pressure trip plus a manual inhibit (Option 2 on NEDE-30045) in their safety evaluation report on II.K.3.18. BECo 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 signal 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 a9to 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 FSAR Section: 6.4.2 Safety Function: In case the capability of the Feedwater System, Reactor Core Isolation Cooling (RCIC) System, and HPCI System is not sufficient to maintain the reactor water 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.
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- 2) System: Core Spray i 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.
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 l 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. This i 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.
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Safety Evaluation Attachment 1 No. 1.124
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- c. Addition of the " ADS Inhibit" switches will allow 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 c 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:
t 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 3 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 i degraded by a stuck open relief valve.
NEDE 30045 then refers to NED0-24708A " Additional Information Required for NRC Staff Generic Report on Boiling Hater Reactors" Section 3.5.2.1, and says that 1 the operator has at least 30 to 40 minutes to initiate !
ADS and prevent excessive fuel cladding heatup for j both the classes of events listed above. This 30 to
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Safety Evaluation Attachment i No. 1/ 24 Sheet 6 of 10 40 minutes was based on starting from full power with equilibrium core exposure and complete failure of all the 1 high pressure makeup systems. '
The addition of the high drywell pressure bypass timer is one of the two modifications proposed by NEDE-30045 that the NRC accepted. The analysis of the effect of adding the high drywell pressure bypass timer is contained in "8ypass 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 eventually require ADS to depressurize the reactor vessel. The limiting case for this class of events is a Reactor Hater Cleanup line break outside primary containment. This event also assumed a loss of all high 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 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 ano actual depressurization is 6 minutes.
The actual time delay setting of 1122 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 the GE Specification 22A3034. i l
This modification provides further improvement in i electrical separation for RHR and 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.
l 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.
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Safety Evaluation Attachment 1 No. 312.4 Sheet 7 of .10 Moreover, the bus "B".is protected from faults which may occur in the "Divison I" panei 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),
switches (GE type CR2940), and indicating lights (GE type and ET-16). The switches and indicating lights 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 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 manufacturer's test response spectra (TRS) for the relays to the TRS curves generated in Calculation 505-C2000.0 for panel C932. Likewise.
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 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 1.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 ADS blowdown. If all these conditions are satisfied, it is GE opinion that vessel depressurization should not be interrupted.
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Safety Evaluation i Attachment 1 No. III4 l Sheet 8 of 10 .f 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 1 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 l 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 provide capability to conveniently disable the automatic logic for starting vessel depressurization. The E0Ps, provide instructions to deliberately disable the ADS logic for the following two conditions only:
l
Safety Evaluation Attachment 1 .No. 2s24 Sheet 9 of 10
- 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, E0Ps 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 E0Ps 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 rnd Core Spray Modi fications 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 BHR 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 retains 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.
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- -In conclusion, the addition of the proposed modifications to the existing logic meets the requirements of NUREG 0737.
!. 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 evaluated 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.
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I For large breaks, the vessel depressurtres rapidly through the break without assistance. The signal for the relief valves to open and remain opeis based upon simultaneous signals from: (1) drywell high O n ss y (2) reactor vessel low-low water level. (3) adequate discharge pressure on one of the LPCI or core spray pumps, and (4) 120 sec delay timer completes timing cycle. Further descriptions of the operation of the automatic depressurtration feature are found in Section 6 Core Standby Cooling Systems, and Section 7.4, Core Standby Cooling System Control and Instrumentation. The Automatic Depressurtration System is designed &s Class ! equipment in I accordance with Appendix C.
A manual depressur17ation of the nuclear system can be effected in the event the main condshier is not available as a heat sink after reactor shutdown. The steam generated by nuclear systes sensible and -
the core decay heat is discharged to the suppression pool. The core is reflooded by the low pressure CSCS. The relief valves are individually operated by remote manual controls from the main control room to control nuclear system pressure. l The number, set pressures, and capacities of the relief valves and safety valves are shown on Table 4.4-1.
4.4.6 Safety Evaluation l ,
(
The ASME Soller and Pressure Vessel Code requires that each vessel' ,
designed to meet Section III be protected free pressure in excess of the vessel design pressure. A peak allowable pressure for upset '
conditions of'110 percent of the vessel design pressure is allowed by .
the code. The code spectf tcation for safety valves requires that.- '
(1) the lowest safety valve be set at or below vessel destga -
pressure, and (2) the highest safety valve be set to open at or below i
105 percent of vessel design pressure.
! The reitef valves are set to open by self actuation (overpressure safety mode) at 1,104 to 1,126 psig and the safety valves are set to l operate at 1,227 to 1,253 psig. This satisfies the ASME Code specifications for safety valves since the valves open below the 1,250 pstg nuclear systes design pressure and below 1,313 psig (105 percent of nuclear systes design pressure).
Safety valve capacity is determined by analyzing the pressure rise accompanying the main steam flow stoppage resulting from an MSIV closure with the reactor initially operating at 1.998
- t. The analysis hypothetically assumes the reactor is shut down by an indirect flux scraa. Reference 1 describes the reasons for choosing this event, the conservatism of applying upset condition 11alts to the event analysis, and models a.nd methodology used in the evaluation of this event. The analysis is repeated for each reload cycle and the results are given in Appendix Q. The analysis indicates that the design capacities of the safety valves and reitef valves are capable -
of maintaining adequate margin, approximately 75 psi below the peak l ASME Code allowable pressure in the nuclear system (1,375 psig). The C) 4d5 'tYi5fon 6 - July 1986 l
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- Attachment-3 Safety Evaluation No 4-124 pups-rsam Sheet i of /3 purpose in that they relieve pressure by inherent mechanical (overpressure) action or by action of an electric pnematic control systak.' The s
. an overpressure relief by mechanical action is initiated inherently by '
condition in the nuclear system. The depressurization by automatic actien of the control system is employed to rekee nucitar system pressure so that the core spray and LPCI when the systems can inject water into the reactor vessel during a Loch NPCIs is inoperable. The automatic control and instraentation equipment for the automatic depressurisation mode of {
relief valve operation is described in this section. j The control " system. which is functiemally 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
,. s valve. . The selenoid operated pilot valve controls the pneumatic )
l pressure ' applied to a diaphragm actuator which controls the relief
.' l.cvalve'.directly.
'*'" An accumulator is included with the control !
I j .. ,' [',; edulpen't . :for each relief valve to store pneumatic energy for relief
" i 3
. :;, ;. valve?eperaties.
a The acc oulators are sized to provide. sufficient
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of twenty pilot ,.actuations fo
- air' supply to the accumulator. : Cables from the semeers lead
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.? cabinetsMoy.fbe','l control Troom.. 'where - the logic arrangements'are form I
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. fo11ewinge,ghe electrical centrol circuitry is powered.by cia ithe I
.> , . ygthout ' automatic transfer. The equipment of ads Logic ' B ' is en ZT,Ba'ttery'97with:,aa automatic' transfer to Battery A
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'"- l "BattTorf B.YTherefore, ' loss of any battery affects emir ,upaa ene 120 losssecof
'1 '. . . .". timing', circuit. Electrical elements in the control system energise ..&
,D * '.[tB cause opsning of .the relief valve. E&ch relief valve tis '. powered ,.
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- 4-j Two initiation . signals are used for the Automatic Depressurisation )
system l 1. Reactor vessel low-low water level
- 2. Primary containment (drywell) high pressure comic *rkmy a re fe NJh*W
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- Ha, /g open. After these .72- :.g
..; :--" there is a
, 120 see time to delay to permit the NPCI to restore water level before the relief valves are actuated. Reactor vessel low water level indicates that the fuel is in danger of becoming 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.
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3 After receipt of both initiation signals, and after 1Ame 2 min dela rewff l provided by timers, the solenoid operated pilot air valve is val *7 '
energized provided that at least one LPCI or core sp,ray pump is' I* 4 4-10 Revision 2 - July 1983 .
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turs-rsAR Sheet 2, of /3 i
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l f M '"*#es 68 running'. An interlock is provided in each trip system in order to 6 give reassurance that low pressure core coolant is available before the ADS actually permits depressurisation of the reactor vessel.
These pressure permissive interlocks are desiped to meet the -
l I
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 ene low pressure pump shows 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 sipal of the presence of pressure in the low pressure systems, an annunciator immediately alerts the operator so that the malfunction can be ,
N ae t,he..k.in=de vm M 1o.r.rM.t.e.d..'ss..ee ne w '
.e s na sI s tLa too%seel ., u .ns's o~evntests.ss*sa"uned r a r.-rs ~ns.s .T'ke m5d4 8 b E2ergization of the solenoid operated pilot valves al ows pneumatie l
pressure from the accumulator to act on the diaphragm actuator. The !
diaphrap actuator is an integral part of the relief valve and espands to hold the relief valve open. Lights in the main control room inform the main control room operator whenever the solenoid operated pilot valve is energized, indicating that the relief valve is open or being opened. ,
, A two position switch is provided in the main control room for the remote control of each relief valve. The two positions are "open" .
and " auto". In the "open" position the switch energises the solenoid-operated pilot valve, which allows pneumatic pressure to .be O. applied to, the diaphrap actuator of the relief valve. This allows the main control room operator to take manual action independent of the automatic system. Appropriate numbers of relief valves can be ]
manually opened in this manner to provide a controlled nuclear system under conditions where the normal beat sink is not cooldown available. In ' auto" position, the valve is controlled by the ads t logic. Manual reset circu ;
i low-low water level and yd -- jts ag provided
- hifor the reactor essure vessel initiating l )
signals. ty manually resetting these sipal pr h ce use recycled. The operator can use the reset swit_ches delayftimers are add % h'S h to delay or N A" H mM autosaus v 4 es == rulius valves? if such delay or,drW h f'a ts p%,
f,'dt ef a merecycles nIN,, fe$[/.Tr@emoj the gn is,gYimer mfor,,u,,alfe,,tuationonone ads " Reset" button y one o(,Ahe
/
, o tr ussstems. The second timen s must be reset in
= Reset" order for button resetstothe the operator de secon(lay eagpget automatic activation of these valves. .
The logic s'cheme used for initiating the ads system is a single trip system containing two trip system logics as shown on Figure 7.4-6.
Each trip system logie can initiate automatic depressurizations when the logic in,that trip system is satisfied. Each trip system logic includes a timer that delays the openihg of the relief valves. This B Sfd H allows time for the HPCI to restore water level before the relief valves are actuated.f The ads trip system is de powered.
St / l' _
- m Instrument specifications and settings used in the original plant d safety analysis are listed on Table 7.4-2. The current instrument ,
7 4-11 Revision 2 - July 1983 w-I
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, No 212 4 PMPS-FSAR Sheet 3 of I3 settisgs are listed in the Technical Specifications referenced in Appendia B. The wiri.ng.for the trip systems is routed in separate conduits to reduce the probability that a single event will prevent .
automatic opening of a relief valve. Pump discharge pressure switches are used to sense that the core spray and LPCI pumps are ,
running. j
)
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 LPCI systems following a LOCA in which the other makeup systems, Feedwater, RCICS, HPCIS fail to maintain vessel water level. The 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 Instrianentation The pressure and level switches used to initiate the ADE are common to each relief valve control circuitry. Reactor vessel low water level is detected by four switches that measure differential pressure. Primary containment high pressure is detected by four pressure switches.
pge (o/ brh & tM f 4 9rippysthee layfrght fe*yad T.W) ,
s Two timers, are used in the control circuitry for each relief valve.
The delay time setting before the AD5 is actuated is chosen -to be long enough so that the HPCIS has time to start, yet not so lang that 4 the core spray system and LPCI are unable to adequately cool the fuel '
if the NPCIS fails to start. An alass in the 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.
I \
7.4.3.3.4 Automatic Depressurisation System Alarms N
& temperature element is installed in a thernovell in the relief valve discharge piping several feet from the valve body. The temperature element is connected to a multipoint recorder in the main control room to provide a means of detecting relief valve leakage during station operation. When the temperature in any relief valve discharge pipeline ,emceeds a preset 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 low enough to give early indication of relief valve leakage.
Additionally available are individual 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-203-31, 3, C, and D, and also the status of the safety valves tv-203 4A and 48. The open/close indication is made possible by the installation of acoustic transducers on the discharge piping of the relief valves RV-203-31, B, C, and D, and on the bodies of the code safety valves RV-203-4A and 3. When the valves are open, indication L
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FNPS-F5AR Sheet 4 of /3 Sheet g of /3 lcf ' indicating lights en the safety and relief
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cudible' alarm will also sound if any of the is and Engic 3 j
tro 10 indicating lights -for each relief valve, entielly to give an indication of valve opening
- ystem is illustrated on cad vibration induced by the steam flow through e initial plant safety
' rent instrument settings eferenced in Appendiz R.
sid2 the cor4 trol room, are also available to *ing the receipt of an i relicf valves.
Depressurisation System Environmental interlocked to prevent ions islenoid valves, and relief valve operators are ' ' **"
22 control and instrumentation equil ment of the t tel inside the primary containment and must remain rallable, th mee symy ironment resulting from a IACA. These items are
, , , ,gg,, , ,,,,,
. litics that permit proper operation in the most sculting from a design basis EACA. Gammt and clos considered in the selection of these iteas, to a preselected inlue.
csted outside the drywell, is selected. in times allowing water to be normal and accident environments in which it ystem Control and Instrumentation 7, ',*Ig,',I' ,", r*ssure and sure. Either laitiation tion and Physical Arrangement I
ses consist of two independent spray loops as ht tk more".Is" 'in danger 7.4-8. Each loop is capable of supplying te owlast.- lDrywell high ster to the reactor vessel to fool the oor* i nuclear.[ system 3peacess o design basis LOCA. The two spray .toops are The reactor'vesse10. low A -
.rically separated so that no ' single physical ,
,,,,,,,, settings and..ths .
ps inoperable. Each loop includes one ac motor
- ,gg,,3, ,,, sel^ectedtead e ate vs1ves, and the piping to route water from geration without Lad'cing u '.
1 to the reactor vessel. The controls and '
' .~. #
the core spray systems include .the sensors, -
1 w,1ve operating mechanisms used to ' start, ch system. Except for the check valves SA and 93 -l
, permit blockage',ofIthe
- 1. TMse switches ;4re '
\
which is inside the primary contalament, the. ma to permit shutdown of
> sing mechanisms for the coce spray systems are out affecting the reactor ter Building. Cables from the sensors are routed l
- room there the control circuitry is assembled in ach core spray pump is powered from a different spray system is a trip le af receiving standby power. The power supply :ircuits. A typical core in each loop is from the same source as that wn on Figure 7.4-4. The ny pump in that loop. Control power for each of can initiate core spray s comes from separate de 'rxses. The electrical ip systems are powered by control room for one core spray loop is located
- from that used for the electrical equipment for 1
spray pumps are shown on I ly started without delay if 7.4-13 Revision 5 - July 1985 Revision 2 - July 1953
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Attachment-3 Safety Evaluation PNPS-FSAR No 2 12. 4 Sheet 6 of /3 the preferred (offsite) ac source is available. Each pump can be ,
manually ' controlled by a main control room reacte sultch or thu 8W automatic control system. A pressure transmitter on the discharge pipeline from each core spray pump provides a signal in the natn control room to indicate the successful startup of a pump. If a core spray inttlation 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 with overload protection. Overload relays are !
applied so as to maintain power as long as possible without lamediate I damage to the mtor.s or emergency power system.
Loss of voltage trips are provided with time delays sufficient to permit automatic transfer from the unit auxillary transformer source to the startup transformer source (preferred offsite) without tripping the' pump power supply breaker open.
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 ammeters. Current or voltage is measured with fleid standard ammeters and voltanters.- ,
.o-The motors 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. .
^
The long tisie, high-set, and groun_d sensor elements d general accordance with recommendations in the IEEE. Induction.Motorw.
s hh ' .
- Protection Guide No. 288, Novenber 1968. The setting of..the l low-set t. --
element is not covered in the Guide. .- . '(7 ;g. j- s The long time element is set at 115 percent to 125'percen't motor current with a time delay set about twice rated actor;of 90ed Vf.
starting'-
time. The long time element is used for overcurrent annunciation.and .
in series with the low-set instantaneous element, set at 'about twice ."'
rated motor current, it is used to. trip the motor circuit breaker for overload protection. This design permits continued motor -operation .
under emergency loading conditions while alerting the operator to a nominal overload condition.
The high-set instantaneous element provides short circuit protection and is set at.about ten times rated motor current which is compatible 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. .
I I
The ground se'nsor relays are instantaneous relays operating from ground sensor current transformers. The relay setting typically provides a 30 to 1 margin of maximum 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 l ground faults when operating on the diesel generator source.
7.4-15 Revision 2 - July 1983
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Attachment-3 Safety Evaluat. ion PNPS-FSAR No 2124 Sheet 7 of /3 Flow measuring 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 otherwise, 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. A11. motor-operated valves are equipped with Italt and torque switches to turn off the valve motor when the valve reaches the limits of movement. Each automatic valve can be operated free the main control room.
Upon receipt of an initiation signal the test bypass valve is interlocked shut. The coie spray pump discharge valves are automatically opened when nuclear systes pressure drops 'to a ,
preselected value; the setting 15 selected low enough so that the low pressure portions of the core spray system are not overpressurtzed, yet high enough to open the valves la time to . provide':. adequate . ,.
cooling for the fuel. .- 5 nuclear systes pressure. Two pressure switches are,.used ,.te@r monitor.:
Either switch can initiate openl 'of the%sG4, '
discharge valves. The full stroke design time of ,the' pusqWdischarge.1EC'+ ':
valves is selected to be rapid enough to assure proper. delivery ~of WG.f ' ^. 91 '
water to the reactor V stroke design operating timesvessel are as follows: in a design , ;4 basis accidentyLThe4f
.% r".,p.g ,.i.gg.pi.
, ,,,,."- Q Test bypass valve 45 sec. 5 i."c.b.T El Pump suction valve 90 sec .
4.,,.
Pump discharge valves 18 sec , . .,:
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Core spray system pressure between the two pump discharge I .
monitored by a pressure switch to permit detection 'ofile '.' ~7.rce - @'" 5 the nuclear system into the core spray systes outside :.the Qrtaar.y p,4'[.
containment. A detection system is also provided toll. con't}nu,ously, w ;.j a l confirm the integrity of the core spray piping between; the. loilde,W,7.,/,'" ' ]
the reactor vessel and the core shroud. A differential : pressure i.:m. / l switch measures the pressure difference between the top of the ' core ' - ;' .
{
support plate and the inside of the core spray sparger pipe just ~ ' -
outside the reactor , vessel. If the core spray sparger piping is .
sound, this pressure difference will be the small drop across the e o core resulting from interchannel leakage. If integrity is lost, this ; ,. -
pressure drop will also include the steam separator pressure drop. .
An increase in the normal pressure drop . initiates an alara In the main control room. Pressure in each core spray pump suction and
~ ;
I discharge is monitored by a pressure indicator which is locally >
mounted to permit determination of suction head and pump performance. f 1
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Attachment-3 Safety Evaluation NPS-N No 2.12.4 Sheet 8 of /5 7.4.3.4.6 Core Spray System Environmental Considerations 9 Therr are -no control and instrumentation components for the core spray system that are located inside the primary containment and that must operate in the environment resulting from a 14CA.
components of the core spray system that are required for system
'All I
g 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 Low Pressure Coolant Injection Control and Instrumentation 7.4.3.5.1 Identification and Physical Arrangement Iow Pressure Coolant Injection (LPCI) is an operating mode of the Residual Heat Removal System (RHRS). Because the LPCI system is designed.to provide cooling water to the reactor vessel following the design basis IACA, the controls and instrumentation for' it are discussed here. Section 4.8 describes the NHR5 in detail. - 9
.I I
Figure 7.4-10 shows the entire residual heat removal system incfuding
' the equipment used for IJCI operation. The following list af equipment itemises essential components for .which.. control..or'
~
instrumentation is required to operate in the IJCI mode: . . ,.. ' d; . f '; -
.,r Q:T. ?,y,g: .
Four NHRS pumps q '.J.* M*!2.;.& . .
8 .' . .;-.'
Pump suction valves (from suppression'yool) C' LPCI-to-recirculation loop injection valves -
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The instrumentation.for LPCI operation provides inputs. ?; ,,
circuitry for other valves in the RHRS. (this is meges.sa.ifte ensure ~ ' ' "
d'd g* ,'
the water paped from the suppression.. pool. .by',;the , p F-j,t
- thatrouted directly to a reactor recirculation .1oop.?.These25'ateel ag = .<?,.t k features are described in this Section. The . actions 'o'f'ihid^irsae).,orN?CM -
recirculation loop valves are described 16 this .Section,becaisse 4 U bese$AIT.
actions are accomplished to facilitate LPCI opefatics[ 'T ,.;}[,'[
?.: <r.t .
..,-i..
LPCI operation uses two identical' pump , subsystems, ' each ' subsystem #,'? ,
with two pumps in parallel. The two subeystems are.larranged- to VF '
discharge water into different reactor recirculation loops. CA cross -
connection exists between the pump discharge lines oLanch subsystem. E-Tigure 7.4-10 shows the locations of instruments, control equipment.
and LPCI components relative to the primary containment. Except fair' check valves 1001-68A, 1001-485 and the reactor l the !.PCI '
recirculation loop pumps and valves, the components pertinent to LPCI
.- operation are located outside the primary containment.
The power for the RHR system pumps is supplied from ac buses that can receive standby ac power. Each pair of pumps in each subsystem receives its powt r fr om a different bus. Motive power for the injection valves era bo".h sides used during IJC' operation comes from a cocoon bus whics ca be automatically connected to either of two i Control power for the LPCI alternate standby poser sources.
-- coeponents comes ,from the de buses. Redundant trip systems are 7.4-17 Revision 5 - July 1985
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Safety Evaluation Mment.33 _
pyyg y No 2.12.4 Sheet 9 of /3 i powered frass different de busses. :The use of consnan buses for some of the LPCI components is acceptable because the core spray systems 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. ,
LPCI is arranged for both automatic operation and remote menuU
operation from the main control room. The equiseent provided for manual operation of the system allows the operator to take action independent of the automatic controls in the event of a IACA.
7.4.3.5.2 LPCI, Initiating Signa".s and tagic The overall operating sequence for LPCI following the receipt of an initiation signal is as follows: ,
i 1. If the preferred (offsite) ac power is avellable, 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 maintained open so thEt no l
automatic action is required to line up suction f 2. If the preferred source of ac power is Imot[* available, ;one ,-
i pump in each subsystem starts after a 5 .sec' delag.after (the', . .f-1 - '
, 'psup .in, ] eachOclg
) standby power source is operating. The ,'s subsystem starts after a 10 see time delay J,. w .M F m Nd'W. .'..
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If the accident has not resulted from, rupture eif'_a['rea'e
- 3. g t." .
i recirculation line, the LPCI instrumentation geleists' " 3.Qryl.:. a e .
for water injection .
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- 4. If the accident has resulted from , ,
recirculation "
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. reactor identifies the damaged loop lines,c..thepC'i .
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- 5. The recirculation pump discharge'N .
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ed 842 loop ' autoimatl 11 . anR?
t reactor recirculation ,, .,M,, ,r..
recirculation pumps are tripped 'l C 2','.D. , Mff. % i M
- 6. Valves in the LPCI system respond au a tJ, . ., d. *.
water pumped from the suppression chamber >1s! routed to'the .;.].,.,f' - ~
undas. aged loop -
' # dA**$'9".N5M'- ' #.
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- 7. hn nuclear system pressure has dropped to 'a- predetermined '
value, the LPCI injection valves to . j the ? undamaged ... .
recircul'ation loop automatically opent allowing the' LPCI 7 p'mpstoinjectwaterintothepressurevessel*.),.
u .: ;, ;.'..
- 3. The LPCI system then delivers water to the reactor vesse1' ' .
via that recirculation loop to restore water level and
. pcovide core cooling Tigure 7.4-10 shows the locations of sensors. Figures 7.4-11, 7.4-12, and 7.4-13 show the functional use of each sensor in the 7.4-18 Revision 2 - July 1983
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Attachment-3 Safety Evaluation No 2.114 PNPS-FSAR Sheet /0 of/3 4
control circuitry for the various LPCI components. Instrument characteristics :and settings used in the initial plant safety analysis are given on Table 7.4-4 The current instrument settings 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 wtth low reactor ;
pressure and primary containment (drywell) high pressure. Reactor vessel low water level indicates that the fuel is in danger of being i overheated because of an insufficient coolant inventory. Primary '
. containment high pressure is Indicative of a break of the nuclear na drywall.
.nSe ff \ system process barrier indda '
- The instruments used to detect reactor vessel low-low water level \
6: l coincident with low reactor pressure and primary containment high i pressure are the same ones used to initiate the other CSCS. Once an l Initiation signal is received by the LPCI control circuitry, the signal is sealed in untti manually reset. The seal-in feature is shown on Figure 7.4-11. .
Keylocked control switches have been installed to peralt b1o'ckage of the drywell high pressure initiation signal. These switches are primarily for use under post-LOCA conditions to permit shutdown of
, the applicable RHR pump actors witicut affecting the reactor vessel
. Iow 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 7.4-6. .
- Elther of the two decision making logies can inttlate the LPCIS. .The tri,n systes is powered by de buses.
7.4.3.5.3 LPCI Pump Control The functional control arrangement for the pumps is shown ce Figure 7.4-11. The reaction of the pumps to an inttf ation signal I depends on the availability of power. If the preferred (offstte) ac !
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 (offstte) ac source is available, the four pumps
- start simultaneously with no delay. Only three of the four RHR 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 provjded in the LPCI circuitry for the RHR pumps, as well as those used for the LPCI injection valves are capable of adjustment over a range of 1.5 times the specified setting listed on Table 7.4-4.
Pressure indicators installed in the pump discharge pipellnes upstream of the pump discharge check valves, provide indication of 9 proper pump operation following an initiation signal. A low pressure v ~I . 4-I 9 SevDn '2. ~ Jut.y 1995
- 7. A.l g Dewieim 2 1n1y IQR4
- - _ 'F . . , . . .
Attachment-3 Safety Evaluation No llA4 PILGRIM FSAR U O 2
Insert "A" hage 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) i 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 light for each switch is illu:ninated when the switch is in the " inhibit" position.
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 I
will not terminate an ADS blowdown once it has begun.
l 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 logics, 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 automatically 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 l
reactor low pressure or high drywell pressure, after a selected time delay (0 l to 30 minute adjustable). The timing function starts when the low-low water l level setpoint is reached The timers are reset automatically if the water i
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 the two ADS trip systems, are depressed.
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. Attachment-3 Safety Evaluation
/
" No ; LIM
~
Shaet 12 of 15 PILGRIM FSAR Insert "E" - Page 7.4-19 (New Paragraph)
LPCI 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 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.
Insert "F" - Page 7.4-11 (New Paragraph)
Keylocked switches have been 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 SRY do not necessarily cause a High Drywell Pressure signal.
Insert "H" - Page 7.4-11 l
Each logic channel also contains a bypass timer, which allows automatic 1 depressurization with low-low water level only, after a predetermined time has I passed. An annunciator indicates that the bypass timer is running and that a low-low water level signal is present.
_ _ _ . _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . . . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ . _ . _ _ _ _ . _ _ _ _ _ . ._.m.._________
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~ Sheet 13 of I3 "LWLblWLd lHDPbIIDIb]l d-c BUS I C s DECISION MAKING LOGIC LWL::LWL::: (TRIP SYSTEM LOGIC) ptDP[::}DP::]
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HPCI TRIP SYSTEM ACTUATION LOGIC (1) q k
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ACTyTg LO0g i, _ ACTUA~ ,J i _,_ L._. -
LPCI TRIP SYSTEM ACTUATOR LOGIC (2)
LWL.1, 1 LWL. 1 --
lHDh.HpP,1,;',, l d-c BUS I 9ECISION MAKING LOGIC
_ = sLWL:; # (TRIP SYSTEM LOGIC)
LWL ,; m hD{::.__ HDg: ] l RLP4 {$80T $80T ) 4RLPi I m~
1 u_---- a CORE SPRAY SYSTEM ACTUATION LOGIC (2)
. .......... . ......... m LEMND n LwL-LOW WATER LEVEL KP-HIGH DRYWELL PRESSURE ADS-At/TOMATIC DEPRESSURIZATION SYSTEM CSP-OPE CORE SPRAY PM RLJN!M LPcIP-OPE LPCI PW RUPNIM FlGURE 7. 4*6 RLP-REACTOR LOW PRESSURE TYPICAL CORE STANDBY COOLING IS-Inhibit Switch SYSTEMS TRIP LPSI-Low Pressure Pump Seal-In SYSTEMS ACTUATION LOGIC BDT-Bypass Delay Tirner PILGRIM NUCLEAR POWER STATION FINAL SAFETY ANALYSIS REPORT
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Attachment- 4 Safety Evaluation.
Nu 282 4 Sheet / of I OTHER FSAR FIGURE CHANGES Figure 7.4-7 (ADS FCD)
Figure 7.3-6 (NBS FCD) See 24A1719 (ADS MOD SPEC)
Figure 7.4-9 (Core Spray FCD) for details of .the changes .
Figure 7.4-11 (RHR FCD)
I 4
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2 13;
, . Safe 8y Evaluation No.: 2. l3 I 1
SAFETY EVALUATION PILGRIM NUCLEAR POWER STATION Rev. No. O j PDC PCN System Calc.
Initiator: Dept: Group: No.: Name: No.: Date: 5%/g7 IfG-75 tpg67 S@DS T"M.Ne645kc NED FS & M C.
$)y,) 4 87-504-Description of Proposed change, test or experiment:
En nd ec) B a r-on M ed if o *c s s , *n u 4e SLc a n ot assoc m 4 e d
% L+,',e I Speci ff e. + ten l charses.
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 malfunction of equipment taportant to safety previously evaluated in the FSAR.
- 2. (>d Does Not ( ) Does create the possibility for accident er malfunction of a dif f erent type than any evaluated previously in the FSAR.
. 3. (>Q Does Not ( ) Does reduce the margin of safety as defined in the basis
-. for any technical specification. '
(
( BASIS FOR* SAFETY EVALUATION CONCLUSIONS:
.Ser A+4a ck ed %e + s 1
Change Change
$<) Recossended ( ) Not Recommended (X ) SE Performed by Nw 90, d Date 4!G[P7 Exhibit 3.07-A Rev. 3 '
Sheet 1 of 3 I
l i
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i Safety Evaluation l c .
No. 2I3r j
%.c. Y 4 157 A. Description of Procosed Chance. Test or Exceriment:
This modification replaces the Standby Liquid Control Systems's (SLCS) <
existing sodium pentaborate solution (natural boron with 19.8 atom percent B10) of 9.4 to 16 weight percent concentration with an enriched sodium u pentaborate solution (bo$i 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 sensirig
~
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 Chance The enriched sodium pentaborate solution is being added to the SLCS to comply with the NRC's Anticipated Transient Without 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, i
1 l
- q. : ia, s ,, A;. y % . 7 . , 2.. , L c-. . ngr '- -- 7-----
Safety Evaluation i , N o .' k 'b '
R w . c - 6/ L ' 57
- The' response time of the system is improved by this modification due to
' higher rate of B10 injection into thr 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.
O. Safety Function of Affected Systems /Comconents -
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 hea;er 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 concentration of 8.42 percent and ,
l aBIO enrichment greater than 54.5 percent provides a total margin of 136 1 l
percent beyond the amount needed to shutdown the reactor.
]
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Safety Evalua% ion
. . No. W .%
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The upper limit, 9.22 weight percent, concentration of enriched sodium pentaborate'has a saturation temperature of 38'F. To preclude E precipitation, the minimum solution temperature will be maintained above E 48'F, which is !0*F above the saturation temperature-of the maximum L 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
- the room minimum design temperature is 60*F, and any temperature change 1 would be gradual. In addition, the daily monitoring will be bancked up by the tank heater, heat tracing, and low temperature alarms. 'If the l solution' temperature in either the tank or pump suction lines' reaches 53*F, the tank' heater or heat tracing will commence operation.. If.the 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 percentofBgv. The54.5 technical specifications re that the B10requl0 will enri enrichment be greater than atom percent. If the B 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 crite ia can be met, Technical Specifications'will require that the B{0 enrichment be returned to greater than 54.5 htom percent within seven days. If at the end of this seven day period, Bl u enrichment is still less than-or equal to 54.5 atom percent, Technical Specifications will require that the NRC be notified within seven days with BECo's plani to bring the enrichment into compliance.
This ensures that if the B ul enrichment is less than or equal to 54.5 atom percent, the operator will shutdown if the original shutdown criteria cannot be met, or bring the enrichment into compliance if the original shutdown q1[r)teria can be met. In' order to ensure a B10 enrichment grea than 54.5 percent the Technical Specifications will require that the Bgr enrichment t,e. determined prior to restart from a refueling outage or any time boron is added to the storage tank. This frequency is considered adequate because B10 is a stable 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 B10 enrichment be determined as part of the receipt inspection before release the matrial for use. Technical Specifications will also require that o{0enrichmenttestresultsbeknownwithin30daysofsamplingthe 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 to the station procedures.
- .c - .sx, ,- - -- . ,, ; '
. .s Safety, Evaluation-
, No. * ' bi P.J.e .wa ni7
, 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 excee.ded. 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 occu.r 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 l 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.
4430)
- Hiah LevelAlarm Alarm (9.22) - (3HO) (9.22) = 10.3 weight percent Low Level ( l
( G. Summary The SLCS by itself cannot cause an 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 l from rated power. Hence, the modification does not increase the l consequences of an accident. )
This modification does not call for the safety equipment of the system to 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.
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 GPM; available 78 GPM) and for minimum volume of solution requirements (required 2068 gallons at a mid-range concentration of 8.82 percent; available 3960 gallons). l
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Safety Evaluation No. 24 3 '
i2,..;, c - s ta is-1 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 question.
p.
References
- 1. General Electric Company Letter, Pilgrim ATWS.SLC System Modification, R. G. Ferguson to R. N. Swanson, dated 2/2/87.
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D 3 ,, gN w * * @WM@ * @ %O l
Safety Evaluation SlG (
SAFETY EVALUATION No.: ()
4 PILGRIM NUCLEAR POWER STATION Rev. No. (d )
A. APPROVAL
- () This Specifications.
proposed change does not involve a change in,the Technical
@ This proposed change, test or experiment does ( ) does not N 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(e) and is reportable under 10CFR50.59(b).
(4 Comments: N uebite N. $lhM -t' The safety evaluation basis and conclusion is:
% Approved () Not Approved iscipline 6hMap LeaderfDete
~ to PTA c)V/y7 AY Shf87 Support 1~ng Bisciplies Group Lander / Bate
- 8. REVIEW APPROVAL
( ( Comments: la cra PaA 2l a.pphes.k h modt h 7h Dre cun-M e f borvn i i -hih o
MenM h usa $roup Le.derenate Y
S/9r7 94/e7 C. Osc REVIEW -
() This proposed change involves an unrevisued safety geestion and a request for authorization of this change must be filed with the Directorate of Licensing, NRC prior to taplemsstation.
() This proposed change does not involve an unreviewed' safety question.
ORC Chairman Date ,
ORC Meeting Number cc:
Exhibit 3.07-A Rev. 3 Sheet 2 of 3
Safety Evaluation No.: ks.
SAFETY EVALUATION IdORK SHEET Rev. No. O
~
System Structure Component Failure and Consequence Analyses.
')
A.
- System Structure Component Failure Nodes Effects of Failure Consnents
% P 4 3s'F SLCS L.6;llb h $ce A m eLed SL,.e4 2.
%+down Ot Reac4&
$7SLC General Reference Material Review CALCULATIONS REGULATORY FSAR SECTION PNPS TECHNICAL SPECS. DESIGN SPECS PROCEDURES GUIDES STANDA G E SPM. 2 5 7 N A ITA 10GR 5'O. (,2 12 3. 4/ 4. 4 O 2,3 r*
( 3,1 5.L
- 14. 2.
A PPs*MG
- 8. For the proposed hardware change, identify the failure modes that are likely for the components consistent with FSAR assug tions. 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 6).
b -
d Date 4!G !&7 Prepared by NOTE:
It is a requirement to include this work sheet with the Safety Evaluation.
Exhibit 3.07-C Rev. 2 I
1
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' Safety' Evaluation No. De%'
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' SAFETY EVALUATION WORK SHEET A. System / Structure /Comoonent Failure and Consequence Analyses
- 1. System / Structure /Comoonent: Standby Liquid 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 determined daily.
- 2. Tank heater and heat tracing commence operation when the solution temperature reaches 53*F.
- 3. Solution temperature of 48'F will alarm in control room, reactor wust be placed into Cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of alarm.
- 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 i
j
)
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ ___ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ .__________..________J
p . . -
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PliGR'Im STATTON FSAR REVIEW SHEET
' Referer es:
Safety Evaluation: Pbc B6-75 Rev. No.: O Date: f/4 /t 7 Support a change List FSAR " test, diagrams, and indices affected by this change and corresponding FSAL revision.
Affected FSAR Revision to affected FSAR Section is shown on:
' Section Preliminary Final
.SEc.7, OMS 3.8,3 3 3.B.4,3.s,5,&3.8f, Attachment 1 fig, 3.3.1 M2413 Rev. E7 Fic 3 .ir.E M - 1 F - 2.3 FIG . ~3.9. 3 E J. %. (;, Attachment 1. I I
Attachment i
Attachment PRELIMINARY FSAR REVIS!ON (to be completed at time of Safety Evaluation ;
- preparation) . .
Prepared by: h. N ,
/Date: 4!(!87 Reviewed by: Date: .,F/I7 l Approved by: /'74 /Date: M/M ;
FINAL FSAR REVISION (Preparet to11owing operational turnover of related systems structures of cogonents for use at PNPS). (1) ;
I" Prepared by: /Date: Reviewed by: /Date:
(1) Attach completed FSAR Change Request Form (Refer to NOP).
Exhibit 3.07-A Rev. 2 Sheet 3 of 3 t
PDC 86-75, Rev. 26h I Safety Evaluation,(Rev. O Attachment 1 SheetIof IG AITACEMENI 1 RECOMMENDED FSAR CHANGES v .
The pages of the following sections, Wab es & figures of the FSAR that need a to be updated due to SLCS modification (PDC 86-75) have been marked with suggested updates and included in this attachment for your review.
FSAR Sections: 3.8.3, 3.8.4, 3.8.5, 3.8.6 Ono wm- i."-2. 3.;-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 FIGUPE TITLE -
M249 3.8-1 P&ID SLC System M-lF-213 3.8-2 SLC System Process Diagram 4
PDC 86-75, Rev. 0 Safety Evaluat on, Revo 0 Attachment 1 Sheet 2 of /6 N G/
PNPS-TSAR I
~ "
3.8 STANDBY LIQUID CONTROL SYSTEM 3.8.1 Safety objective The safety objective of the Standby Liquid Control System (SLES) is to provide a backup method, which is independent of the control rods, l
to maintain the reactor suberitical as the nuclear system coals in the event that not enough of the control rods can be inser.ted to counteract the positive reactivity effects of a colder moderator.
- 3.8.2 Safety Design Basis i
- 1. Backup cepability for reactivity control shall be provided, independent of normal reactivity control provisions in the nuclear reactor, to be able to shut down the reactor if the s nonnal control ever becomes inoperative. ,
- 2. The backup system shall have the capacity for controlling the
, . . . ~J re, activity difference between the steady state rated operating ,
condition of the reactor with voids and the cold shutdown '
.~ condition, including shutdown margin, to assure complete shutdown 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
(- ,
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 l
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 system; the possibility of unintentional or accidental, shutdown of the reactor by this system shall be I sinimized. '
3.8.3 Description The piping and instrumentation for the SLCS is shown on Figure 3.8-1.
Tigure 3.8-2 is a process diagram for the system. The SLCS is manually initiated from the main control room to pump a boren neutron absorbercannot solution into the reactor if the operator believes the reactor be shut down or kept shut down with the control rods.
? . n. a.
l PDC 86-75, Rev. O j
, S fety Evaluatipn, Rev. O !
I Attachment I b 0f f 6 /
PNPS-TSAR f, 1 -
i
'$ However, insertion of control rods is expected to always assure
- l l
prompt shutdown of the reactor should it be required. The boron
{d absorbs thermal neutrons and thereby terminates the nuclear fission '
jN[
j chain reaction in the uranium fuel.
l 4 l 8 The SLCS is needed only in the improbable event that not enough core to accomplishg l
4 }s control rods can be inserted in the reactor The SLCS therefore is c ,g 3 j shutdown and cooldown in the normal at manner.
a steady rate within the'$J'e sized capacity onlyofto shut the reactor downthe Shutdown Cooling Systems, and keep the reac
"?'$ a
- A going critical again as it cools, i
boron solution tank. the test water tank, the two positive h t
y
- g #i'r fe Thedisplacement pumps, the two explosive valves, valves and controls are and associated local mounted in the Reactor Building outside the g The liquid is piped into the reactor vessel and ,g y(
g 4 primary containment.the bottom of the core shroud so that it mixes with j o- dy discharged near See Section 3.3, Reactor l i the cooling water rising through the core. ~
-p" f / Vessel Internals Mechanical Design, and Secpon 4.2, Reactor Vessel *y g
-a f and Appurtenances Mechanical Design. @ ft wh ,
The specified neutron absorber solution is aisoditanpentaborate n: m -
inu ;
.j $ $ solution. 4,It bri; is prepared byfessousan.ng sinr -
..: in demineralized water. An air sparger is f _::= =r To ' prevent system plugging, the
(
provided in the tank for mixing. tank outlet is raised above the bottom%..of the f
a strainer.
- gg '
\
f' At all times when it is possible to make the at reactori1.iti:1 least O galcoreof k D critical, the SLCS shall be able to deliver -
e.t?. Q percent # sodium pentaborate solution ^ %=GnN or equivalent Win into.2the Q s*e reactc y W "h jeko& stan G.V liqui oy control tank and lling vit 1
~
m apo ate in clume. .hw "I tg demin alized wdesign east th low .le 1 alarm er to a oncentra d on -at th low level lars po' t and 5 ac sol ion is ed up o the erflow vel vol to all v _foM - { *b $g a be di
$ the sjy .QLort tercerpun# - s .; .p 5
veo ai 1 sstLs _r to lov-e !
j
_ ion temperature of the specified solution is M FF so the The u saturat equipment containing the solution is installed in a room in
'" which the ^^"
a air temperature is to be controlled E tank, and a l 2 -~ - '
- -3.
temperature controller An electricm '; inenersion mis,gesse heater
- !: rused i.t in thet f ir d::+" N 7 i to elevate g :.. ,.., c;. N n _ q the !
p and assure that the boron dissolves when first added or to I
/ mtemperature
~*4--
the water. High or low temperature, high or low liquid level, a
I STET shorted heater causes an alarm in the control rooE bt- 4.mb
's sized '"' the solution into to inject j l
Each positive displacement pum __ _: ' : solution level in the the reactor in 50 to 125 inin, ,
The pump and system design tank, at all reactor operating pressures.The two relief valves are set to exceed the I pressure is 1,500 psig. suf ficient margin to avoid valve f reector operating pressure by a flooded f leakage. The relief valves are installed with the discharge l
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PDC 86-75, Rev. 0 Safety Evaluat on, Rev.0 Attachmen2 1 4/3j
'. GENERAL ELECTRIC CO. sheete/ of /6 Nuclear Energy Business Operations ENGINEERING CALCUL4.10N SHEET DATE ,
NUMBER -
- SY SHEET or SUBJECT -
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I PDC 86-75, Rev. O
' Safety Evaluation, Rev. O Attachment 1 Q Sheet S of 16 21 h PNPS-FSAR To to prevent evaporation and precipitation within the valve.
prevent bypass flow from ore pump in case of relief valve failure in' -
the line from the other pump, a check valve is installed downstream i, of each reitef valve line in each pump discharge line.
The two explosive actuated injection valves provide high assurance of opening when needed and ensure that boron willThe not valves leak into the have a reactor even when the pumps are being tested. Each explosive va.1ve firing reliability in excess of 99.99 percent.The plug is circumscribed is closed by a plug in the inlet chamber.
with a deep groove so the end will readily shear o'f when pushed. by the valve plunger. This action opens the inlet hole through the l
! plug. The sheared end is pushed out of the way in the chamber, and -
I 15 shaped so it will not block the ports after release. l The shearing, plunger is actuated by an explosive charge with dual
- ignition primers, inserted in the side chamber of the valve. .
Ignition circuit continuity is monitored by a trickle current,Indicator and an I I
alarm occurs in the control room if either ci,rcuit opens.To service a valve.
lights show' which channel primer circuit opened. f af ter firing, a 6 in length pipe - (spool piece) must be removed
- l
~ isrnedtately upstream of the valve to gain access to the shear plug. L The SLCS is actuated by a three pos'1 tion keylock switch .on the This assures that switching from the "off" .
control room console.
~
position is a deliberate act. Switching to either side starts one
- injection pump, opens an explosive valve, and closes the Reactor j ,
' Cleanup System tsolation valves to prevent loss or dilution of the h (
boron. h$ I A green light in the control room indicates that power 1s available J to the pump noter contactor, but that the contactor is open (pump not g running). A red light indicates that the contactor is closed (pump 6 h
running).
4h
. ~
L.iquid flow is confirmed by a decrease in reactivity, storage ,4 tanka C red lig, drawdown and pump running indication. j }f# l switch turns on when valve 1101-1, downstream of the explosive valves j 1s open. If the pu;np lights or explosive valve light indicates that the 11guld may not be flowing, the operator can lamediately turn the J.t
- keylock equipment.
switch to the other side; this switch actuates The chosen pump will start fgthe altern either even though its local pump and either explosive valve. switch at the pump is in the jv "sNp"J position
~
Pur.p discharge pressure indication is also y
,for test or maintenance.
provided in the control room. lj1
- l l
Equipment drains and tank overflows are piped not to the .Waste {g but to separate containers (such as 55 gal drums E _.
. : :- : System c J Gn .M to prevent any trace of boron from #
inadvertently reaching the reactor. Tvcw Arw
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e.u k. n.w..wam. ul J
%.u .4 m&gmina4%,er-ceA l l
Instrumentation is provided locally at the standby liquid control tank and consists of solutinn temperature indication and control, l
PDC 86-75, Rev. 0 y - Safety Evaluatipn, Rev. 0 Attachment 1 yI Sheet (o of /(, (,_,7 / 3 f PNPS-FSAR s
tank level, and heater status. Instrumentation and control logic is
-C ,
c presented on Figure 3.8 4. '
3.8.4 Safety Ev41uation l
e.1 The 51.C5, although not necessary for plant operatton, is required to I< be operable when the reactor is in other than cold condition.
3 ag f Despite this precaution, the system is expected never to be needed for plant safety because of the. large number of Independent control rods available to shut down the reactor. To ' further assure this a' ,"
availability, two sets of the components required to actuate the % er ,
{ pumps and explosive valves are provided in parallel redundancy.
, 1$ r
. p.
The system is designed to bring the reactor from rated power to af a # Y
$ cold shutdown at any time in core life. The reactivity compensation g ,4 y3 l
to zero and allow 2
g
.a g provided will reduce reactor power from ratedcoolingIt the#,INuclear Syste ly control rods remaining withdrawn in the rated power pattern.
includes the reactivity gains due to complete decay of the menon v i AB ,f inventory. .It also ' includes the positive reactivity effects fron %
4 eliminating steam volds, changing water density from hotMj. to cold,t -
l reduced Doppler effect in uranius, reduction of neutron leakage fresh t ,
dg bolling to cold, and decreasing control rod worth as the moderator (concentration cools. The. specified minimum final 4 , . . ' . l.
g f l
- g. reactor core provides a reactivity worth of approximately -0.12 A kb l .e#p i l
($g g plus a margin of -0.05 A k for calculational uncertainties ane M l
- e-U g e , assures a substantial f tso shutdown too margin. f ,s, .3 JS The specified minimus ave ge concentration of natural boron in theT J reactor, to p ovide the s ecified shutdown margin af ter operation of Nj y
the $1.C5, Is G Cp ppm, .
The alntmum quantity 4
-4 fesodlue pentaborate to be injected into the reactor 15 calculated , _ J based on the required ppe average concentration la the reactor n0 f
- I y of reactor coolant in the-:-reactor ^vestel 'N
$ coolant, and the quant and rectreulation loops = - -
for taperfect string, result is increased by 25 percent to al
-.--low- - -
leakage, and volume in other small piping connected to the reactor.
W Cooldown of the Nuclear System will take several hours as a minimum, -
to remove the thermal energy stored in the reactor, cooling water, and associated equipment and to remove most of the radioactive decay heat. The controlled limit for the reactor vessel cooldown is 100*F/hr, and normal operating temperature is about 550*F. Usually, shutting down the plant with the main condenser and various shutdown cooling systems will take 10 to 24 hr before the reactor vessel is opened, and such longer to reach room temperature 00*F) which is the condition of maximum reattivity and therefore, the condition which requires the martmum boron concentration.
5.b%a The teemo injection rate ts 1tmited to the range of 39 to 79 gal / min.
The lower rate assures that the boron gets into the reactor in about 1 1/2 hr, considerably quicker than the cooldown rate. The upper
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PDC 86-75. Rev. O Safety Evaluati , Rev. O Attachment 1 o Sheet 7 of fS' ,zm / ;
PNPS-FSAR 11mit injection rate assures that there is sufficient alzing to the boron does not recirculate through the core in uneven concentrations '
which could possibly cause the nuclear power to rise and fall -
l
. cyc11cally. -
l l
The SLCS is designed as a Class I seismic system. The system pfptng and eculpment are designed, installed, and tested in accordance with f l
Nonprocess equipu.ent such as USAS B31.1.0 Section I and Appendix A. ,
{
4 the test tank is designed as Class II.
5, The SLCS is required to be operable in the event of a station power l failure, so the pumps, valves, and controls are powered The from the
. pumps l 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 are. system.
poweredoperation.
from the l
essential instruments and lights 45 The -
$ ., 120 V ac instrument power supply. 10G 5 ;
8
,o -The SLCS an# pumps have sufficient pre-e margin. Up to the system 3 reitef valve setting of 1,400 psig, to assure solutto injection , psig in into .the .
the reactor above the normal pressure of .aboutThe nuclear system reitef and safe
.l J .
bottom of the reactor. '
E f begin to re leve pressure above about 1,100 psig;
- positive d ad' therefore,-
neap deo 4 dad. A a.
- - g
'37 The system \1s aM to provide 1 concentration l of amtmani boron in l
' the reactorAcf 700 ppm The shutdown s'arcin from this concentration J Pligrim's Supplemental Reload License Submittal in 3" J can be foundThe Appendix Q.
in analysis and models for the reload core are described in the GE Standard Application for Reactor Fuel."'
(i
.er 3.8.5 Inspection and Testing l
Operational testing of 'the SLCS is performed in at least_ two parts _to j avoid inadvertently injecting boron into the reactor.jy ope ing EEI s u o m neethe valves tr[ ate y 1 Met [n$ i its lo d ch. to and from opened thetosolution and from tank the Closed test tank, and the i ~three the valves (two locked deminerallred waterclosed) in the test tank can be recirculated by d 3 locally. mp a n ogs fSt me p to plu . mt he i a turning _ on either pump ,m,a _
nv mee vetr u unctional testing of the injection portion of the system is' ,
accomplished by closing the locked open valve from the solution tank, l
' opening the locked glosed valve from the test tank, and actuating the in the control room to either the A or B circuit.
keylock switch and blows open the injection valve in that This starts the pumpThe lights and alarms in the control room indicate that the circuit.
system is operating.
l
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1 j
PDC 86-75, Rev. O Safety Evalua%, n, Rev. 0 Attachment 1 PNPS FSAR Sheet 6of16 2/3/
local locked open valve to the reactor in the j
- By closing aleakage through the injection valves can be detected at containment, 3 a test connection in the line between the containment isolation c valves.
closed for tests, or open and ready for ,
A the local valve is from the reactor through the first check valve sj operation.) 1.eakage
. can be detected by opening the same test connection whenever the e reactor is pressurized. -
3 injection valves and explosive j l After the functional tests, the & ;
charges must be replaced and all valves returned to their normal
~ l A
positions, as indicated on Figure 3.8 1. 3
~
The test tank contains demineralized 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 raf tlling 4 or fl.ushin the system. *
- ~d1 g Should the boron solution ever be injected into the reactor, either < f"h.,-
intentionally or inadvertently, then, theafreactor ter making certain subcritical, thethat
+ the y j
". '. {.5 normal . reactivity controls will keepboron' 3 is rem gross dilution followed by operation of the Reactor Cleanup System.J e There is practically no effect on reactor operations when thepr P -
concentration has been reduced below approximately 50.. ppe. 3
(
~
The concentration of the sodlum pentaborateiln the solutton' tank f s M .-
determined by chemical analysis periodically. %s e.sichact d A 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.
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Revision 6 - July 1986 3.8-6
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PDC 86-75. Rev. O
[ Safety Evaluatjon, Rev. O
' PNPS-FSAR Attachment Sheet 9of/6 1 L '2/? f 3.8.7 Current Operational Nuclear Safety Requirements Ilmiting condition for optration, sarvet11ance
- current The requirements, and their bases are contained in the Technical Specifications referenced in Appendix 8.
3.8.8 , References NEDE-24011-P-A, General Electrical Standard Application for
- 1. .
Reactor Fuel. Applicable revision.
P, Stand t9 Lihuici Control Sp l es G N l Cerecity in i
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FIGURE 3.8-3 -
SATURATION TEMPERATURE OF SODIUM PENTABORATE SOLUTION
$ PILGRIM NUCLEAR POWER STATION 1
b FINAL SAFETY ANALYSIS REPORT
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l SOLUTION VOLUME j
CONCENTRATION REQUIREMENTS !
PILGRIM NUCLEAR POWER STATION FIN AL S AFETY AN ALYSIS REPORT l
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Safety Evaluat on, Rev.' 0-
! Attachment 2 .
.Sheetiof) 113/
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'. GENER AL hELECTRIC-
.c .
STANDBY LIOUID CONTROL SYSTEM .,
i CONTROL CAPACITY
~
EQUIVALENCY REPORT
' i PREPARED FOR THE BOSTON EDISON COMPANY ,
PILGRIM NUCLEAR POWER STATION
[~ , JANUARY.29, 1987 i
1 PREPAPED BY: R.T. EARLE E Eda.r 8 l/29d?
YF.RIFIED BY: J.K. SAWABE I 24 M 7
).'A. Q Verification Material in DRF C41-00095/2, Section L4.
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.' Safety Evaluati n, Rev. 0 l
- ,
- Attachment 1, Sheet 2,of 9 1;g j l
L DISCLAIMER OF RESPONSIBILITY .
This document was prepared by or for the General Electric Company. l Neither the General Electric Company nor any of the contributors to
" )
this document: .
A. Makes any warranty or representation, express or implied, with 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, I
B, Assumes any responsibility for liability-or damage of any kind i
(
which may result from the use of any information disclosed in 4 this document. ,
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Attachment 2 Q , Rev. 0 j Sheet 3 of aj 211; j d
i ABSTRACT ,
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1 This document was prepared for the Boston Edison Company to address ,
l the requirements of the Standby Liquid Control System (SLCS) at-the s !
Pilgrim Nuclear Power Station for compliance with the NRC ATWS Rule 10CFR50.62. The plant specific values used to demonstrate compliance with the NRC ATWS Rule are the same as the minimum values provided in the system Technical Specifications.
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PDC'86-75, Rev. 0 ,
Safety Evaluation, Rev. 0 Attachments Sheetfof9.{~?.I3l-j TABLE OF CONTENTS
~~
11 Abstract 1
- 1. Introduction , .
1 ,,
- 2. Discussion 1
2.1 SL,C System Design Basis ,
. 4 1
2.2 NRC ATWS Rule *
(_ l'
- 3. Analysis .
2 3.1 Equivalent Control Capacity Definition 3
3.2 Equivalent Control Capacity, Calculation 4
- 4. Summary 5
- 5. References .
4 4
O
- 111 - .
b.
- : % :w: : . . . . . . .- . . .. . . m , ., . 8 PDC 86-75, Rev. 0 Safe 9y Evaluati@, Rev. 0
. Attachment 1, SheetSofp . L > 12 l 1.0, INTRODUCTION I
Boston Edison has requested an evaluation of the minimum required
! concentration (weight percent) of sodium pentaborate for the Pilgrim Standby Liquid Control system to comply with the NRC ATWS rule requirements in 10CFR50.62 (Reference 3.). The minimum concentration is to be based on equivalency to the minimum 86 gym, 13-weight percent l 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 j required. This is based on the assumptions that the minimum -
f the s dium pentaborate decahydrate exceeds 54.5 bf5 B
n enrichment atom percent, one pump is required to operate and the actual capacity of each pump exceeds the required minimum pump flow rate. ~
. 4 2.0 DISCUSSION l( ;.1 SLC System Design Basis The generic design basis for the SLC system is to provide a specified cold boron shutdown concentration The to SLC thesystem reactor was vessel as described typically designed in NEDE-24222 (Reference 4.).
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 solution to the generic S',C System design basis. Changes to flow rate, concentration or boren enrichment, to meet the ATWS Rule, must not invalidate the original system design basis.
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PDC 86-75, Rev. 0 Safegy Evaluati n, Rey
- o :
Attachment 2. b Sheet (o of c, Qg,
.2 NRC ATWS Rule
~
Paragraph (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 baron content equivalent in control capacity to 86 gallons per minute of 13-weight percent sodium pentaborate solution."
The NRC Staff has provided clarification of equivalent control capacity (Reference 5.) as follows:
(1) The " equivalent in control capacity" wording was choosen to allow flexibility 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 were 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 f act that diffierent values would be equivalent for smaller plants was recognized in NEDE-24222.
"The flow rates given here are normalized from a 251-inch i.e., !
diameter vessel plant to a 218-inch diameter vessel plant, the 66 gpm control liquid injection rate in a 216 is equivalent to 86 gpm in a 251. This is done to bound the analysis....(pp.
2-15 INEDE-24222))."
(3) The important parameters to. con' sider in establishing equivalence .
are vessel b'oron concentration required to achieve shutdown and the l
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 .
l 3.1 Equivalent Control Capacity 1
The NRC equivalent, control capacity concept of the ATW5 rule is a very simple, direct criterion that does not require consideration of the
. mixing efficiency or to account for plant-specific core nuclee.r characteris .ics. Consequently, Pilgrim can demonstrate incompliance with
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PDC 86-75. Rev. 0 Safety Evaluation
. Attachment'1, Q , Rev. 0 2.g g ,
, Sheet J ofc) the-equivalency requirement if the fc11owing relationship is shown to
~
be true:
>= 1 (Equati,on 1) g EU251
- C
- E. .
86 M 13 19.8 I
where the plant-specific parameters are defined as:
s s
.j Q = minimum SLCS flow rate (one or two pump operation l appropriate), gpm. .
l
/
M = mass of water in the reactor vessel and recirculation system j
at the hot rated conditions, lbs.
- k
}
C = minimum sodium pentaborate solution concentration, weight percent. '
10 isotope enrichment (19.8% for natural E = minimum expected B , -]
I boren), atom percent. .
(
The value of M (the mass of water in the reactor vessel and . i recirculationhstematratedconditionsintherefer'encepla 628,300 normal water level, control rods ,
temperature, rated void content, fully withdrawn, expected minimu ,
internals dimensions.
3.2 Equivalent Control Capacity Calculation The NRC requires the use of minimum' plant-specifi pump operation, 54.5 atom percent boron enrichment, Pilgrim can ;
I demonstrate compliance if the following relationship is true: .
(Equation 2.)
C >= 13
- M.
- 86
- 19.8 ~
M 251 9 i ht l where C is this case is the minimum allowed concentrationllowed (we g percent) of the sodium pentaborate solution, Q is the minimum a individual pump flow rate, M is the mass of water in the reactor an recirculation lines, and E is the minimum allowed boron'. enrichment
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l PDC 86-75, Rev. O l
. Safety Evaluation, Rev. o ]
i Attachment L Z Sheet gof g g l l
l 1 level. The water mass is based on the same conditions as the reference j plant water mass. The minimum allowed individual pump flow rate was I obtained from Reference 1. The minimum allowed boron enrichment level
' will become part of the system Technical Specification and Design F Specification (Reference 1). .
l
~-
Q = 39 (minimum rated) gpm M = 507,850 lbs E = 54.5 % i -
Using the current Pilgrim plant-specific values (in Equation 2) gives a required minimum concentration of 8.42 weight percent sodium .
pentaborate.
13
- 507,850
- 86
- 19.8 (Equation 3.)
C >=
54.5 628,300 39 ~
e
( -
C >= 8.42 4.0
SUMMARY
Whentheconcentrationofenrichedsodiumpentaggratedecahydrate ) is equal to, or (enrichment exce nding 54.5 atom percent boron B greater than 8.42 percent, Pilgrim meets or exceeds the NRC ATWS rule-equivalency requirements. ,
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~
Safety Evaluati n, Rev. 0 Attachments.
Q;3, Sheet c) of c) !
. s.O REFERENCES
- 1. Doc. No. 257EA169, Rev. O, Standby Liquid Control System Design Specification.. . .
- 2. Doc. No. 257HA169AV, Rev. 2, Standby Liquid' Control System Design Specification Data Sheet. .
l
- 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, Clarification Control capacity of forEquivalent Standby Liquid Control Systems, January -
- 28, 1985.
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Attachment 3 Sheet- of 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. .
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-LIHfTVNG CONDITIONS FOR OPERATf0N SURVEILLANCE REOUTREMENTS
. If 3.4 STANDBY LIOUID CONTROL SYSTEM 4.4 STANDBY LIOUID CONTROL SYSTEM l
Aeolicability:
Aeolicability: J Applies to the operating status of the Standby Liquid Control Applies to the surveillance i System. requirements of the Standby !
Liquid Control System. !
Obiective: -
l To assure the availability of a Obiective: l i
system with the capability to To verify the operability of the shutdown the reactor and maintain Standby Liquid Co,ntrol System. l the shutdown condition without '
the use of control rods. ;
Specification: ,
1 Specification: l A. Normal System Availability A. Normal System Availability ;
- 1. During periods when fuel is f in the reactor and prior to startup from a cold The operability of the Standby !
condition, the Standby Liquid Liquid Control System shall be '
verified by the performance of Control System shall be the following tests:
operable, except as specified '
in 3.4.B below. This system 1. At least once per mo' nth p need not be operable when the each pump loop shall be L reactor is in the cold functionally tested by condition and all control recirculating ;
rods are fully inserted and demineralized water to the ,
Specification 3.3.A is met. test tank.
- 2. At least once during each operating cycle:
- a. Check that the system relief valves trip !
full open at pressures less than 1800 psig, and reseat on a falling pressure greater than 1275 psig.
- b. Manually initiate the system, except ,
explosive valves. !
Pump boron solution through the recirculation path and l back to the Standby l Liquid Control l Solution Tank. Check j that each pump flow ;
rate exceeds 39 GPM l against a system head i of 1275 psig.
1
____________________1.__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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9.:. .'. ; Y ~ ; , :..,. '=,VTE L= ~ =- ;;;,.L.T-, T LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUfREMENTS __
, 3.4. STANDBY LIOUID CONTROL' SYSTEM 4.4 STANDBY LIOUID CONTROL SYSTEM O
- c. Manually initiate one
' of the Standby Liquid Control System loops
~ and pump demineralized water into*the reactor vessel.
This test checks explosion of the charge associated with the tested loop,.
proper, operation of the valves, and pump-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 InonetAhle B. Surveillance with Inonerablo Comoonents: Commone m -
- 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 until the fulfilled and continued inoperable component is operation permitted repaired.
provided that the i component is returned to !
an operable condition within seven days.
96
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. LfMTVXNG CONDITXONS FOR OPERATLON StjRVEILLANCJ REHJIREMENTS 3.4 STANDBY LIOUID CONTROL SYSTEM 4.4 STANDBY LIOUID CONTROL SYSTEM C. Sodium Pentaborate Solution C. Sodium Pentaborate Solutiqa At all times when the Standby The following 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.
L 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 6 3.4.1. least once per scnth.
Also check concentration I
- 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.
- 3. The enrichment of the 4. Enrichment: Check liquid control solution Boron-10 enrichment level shall be maintained at a by test anytime boron is boron B10 isotope added to the solution and
( D.
enrichment exceeding 54.5 atom percent.
If specification 3.4.A B, or prior to restarting from each refueling outage.
Enrichment analyses shall be received within 30 days C.1 or C.2 cannot be met, the of test performance. 4dinen reactor shall be placed in a S:t n-2: ---:: ,
Cold Shutdown Condition with F a check shall be maae to 7 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 specif6 tion / enrichment, concentration 3.4.C.3arenotmet,[he6ag and volume to established
& the Boron-10 Isotonic " criteria. 'N the See enrichment top 4.5 Atom ^ 3 ' fevels clo hof moet If +he Bo* percent within seven days th9n i levels Ynce;t from thejime of enrichment *be. or' Shm I cles #*Jh CNYevib J recess.4submitareportto fhe Pegetor shyll (pe f accol l m. 9 g" l /
Colcl 6hddown Conc}, hon with the NRC.and advise them of des.i9h plans to bring the solution fdl Cdtenc4) up to a demonstratable 54.5 gli operable (chMI rods * ' O '" 24 h o es , y/ /
D5"d*d b viq atom percent Boron-10 ff 4Y4et- Yb 5 klm A- f CH o oI Isotopic Enrichment. g g gy y y
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BASES:
. -3.4 & 4.4 STANDBY LIOUID CONTROL SYSTEM A. The requirements for SLC capability to shut down the. reactor are identified via the station Nuclear Safety Operational g Analysis (Appendix G to the FSAR, Special event 45). If no i more than one operable control rod. is withdrawn, the' basic shutdown reactivity requirement for the core is satisfied and e
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 systqm is L required. The design objective of the standby liqutd control
! system is to provide the capability of bringing the reactor from full power to a cold, xenon-free shutdown condition 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 ppm of natural boron in the reactor core. The.700 ppm equivalent concentration in the reactor core is required to bring the reactor.from full' power to a three percent Ak subcritical condition, considering the hot to cold reactivity difference, xenon poisoning etc. The system will inject this boron solution in lesh than 125 minutes. The maximum time requirement for inserting the boron solution was selected to override the rate of reactivity L
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-Water-Cooled Nuclear Power Plants). The Standby Liquid Control system must
- a (Y 4'#
have the equivalent control capacitygbf 86 gpm 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 B.42% concentration and 54.5% enrichment of Boron-10 isotope atf39 GPM pump flow rate
- q satisfies the ATHS Rule (10CFR 50.62) equivalency requirement.
. Because the xadeed concentration / volume curve has been yevisdeedsetsQto reflect the increased B-10 isotopic enrichment, e*-
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M&5 hh:555Sh~e Experience with pump operability indicates that the monthly 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.
9 b1380% l FP jEe>neni yo eva lMc. ]be.
Amendment No. 15 f,y,,n g c.9.p,jg7p, meet +be.orihlff3',$n 6hdc}cWh c.rHeviq w h e rj eve r-4he- B-io ewcheneet PehhWeWI is hok kneih% been added,
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0 ' BASES:
3.4 & 4.4 STANDBY LIOUID CONTROL SYSTEM (Cont'd)
The minimum limitation on the relief valve setting is intended to prevent the loss of sodium pentaborate solution via the lif ting 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 Jocps is needed for operating the system. One inoperable pumping circuit does not immediately threaten the shutdown capability, {
and reactor operation can continue while the circuit is being l repaired. Assurance that the remaining system will perform its l 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 beyond 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 cuard against boron precipitation.
Minimum solution reem-teit.prature is 48'F.
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 4a4y 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.
- 101
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NED Pr0 posed Change Safety Evaluation
' ~ '~
No.: 2/33 SAFETY EVALUA110N PILGRIN NUCLEAR PG4ER STATION j Rev. No.
PDC Pol System Calc.
Initiator: Dest: Group: No.: Name: No.: Date: 1 T.M.Hauske NED +64Me 86-52A Residual j NA G Heat {
Removal j Description of Proposed change, test or experiment: -4 I
Replacement of the RHR Containment Spray Caps SAFETY EVAL 11ATION CDNCLUSIGIS:
The proposed change, test or experiment:
\
- 1. (x) Does Not ( ) Does increase the probability of occurrence or 1 consequences of an accident or malfunction of equipment laportant to ,
i 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. CX) Does Not ( ) Does reduce the margin of safety as defined in the basis for any technica? Jecification.
BASIS FOR SAFETY EVAltl4TIGI PN11tEIGli-
+
Ree Attached Sheets
{ _
i
~
t . Change Change
! (X) Recommended ( ) Not Recommended SE Performed by h Date 5!I Exhibit 3.07-A Sheet 1 of 3 3.07-13 Rev. 4
L Safety Evaluation No. 'l f 3 7 Page2of32, SAFETY EVALUATION OF REPLACEMENT OF DRYWELL SPRAY HEADER CAPS A. Description of Chance t
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 1-1/2"-7G-25 Fogjet caps has all seven nozzles open.
B. Purcose of Chance g4 The replacement of the RHR containment spray header caps wi'll result in' ( ,$
e e hea4ee . This flow reduction will minimize the possibility of 4,.au 4
~
damaging the drywell structure by sudden decompression following the Erg 44, 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. (SEE~ ATTwen 3r" T_ }
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 i suppression pool return valve needs to be open during containment spray so that rated flow through the RHR heat exchanger will be J maintained. l 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 torus-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 W.at Removal SystgLm The RHR cools the suppression pool 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 spray mode of the RHR operation (FSAR Figures 5.2-2 to 5.2-7).
l
y -
.y , , ,;,
' Safety Evaluation No. A If 3 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 l vacuum breakers. This delay is included in the analyses of Section F and has no deleterious consequences.
E. Effect on Safety Functions The proposed change would affect the safety functions identified in Section 0, as discussed below:
- 1. Effect on Drvwell Resoonse The proposed 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 Drvwell Resconse In Reference 2, GE reported the result of their evaluation of the L reduced containment spray. GE reanalyzed the FSAR containment response for break sizes ranging from 0.0? to 0.5 ft.2 assuming the reduced containment spray was initiated J0 t 9utes after containment pressure reaches 10 psig. This assumption is consistant with the present
)
'
- Safety Evaluation No. 1 / O i Page/ofg
- 1. Effect on Drvwell Resconse (cont.)
FSAR requirement. It was determined that a containment spray flow rate of 300 gpm is sufficient to reduce the airspace temperature 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 operating, has been calculated (Reference 1) to be 543 gpm when the suppression pool bypass valve (1001-36A,B) is open with total RHR flow 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 Exchancer Efficiency When operating the containment spray, the 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 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) requiring an operator action rhich will be incorporated into the revised operating procedure 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 other potential safety issues were found during~tnis safety evaluation.
Thus, no conclusion of the FSAR is affected and there is no reduction in margins of safety. No new accident is introduced not is the probability or consequences of previously analyzed accidents increased.
H. References
- 1. Bechtel Flow Analysis Calculation 17322-H-660-1.
- 2. General Electric Letter G-HK-7-157, dated 4/20/87, " Safety Evaluation of Proposed Capping of Certain Drywell Spray Sparger Nozzles".
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4 Safety' Evaluation No. A i 3 1 Page5ofy ATTACHMENT 1 RECOMENDED FSAR CHANGES ,
The pages of the following sections, tables and figures of the FSAR tnat need .
i 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
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- Safety Ev'aluation
. N3.: 3113 i SAFETY EVALUATION WORK SHEET Rev. No. D A. System Structure Component Failure and Consequence Analyses. M 0 of M' Systen Structure Connonent Failure Modes Effects of Failure Comments O
- 1. SHR_Conta,inment DW Temo 2,Q F _See Attached Shee_t f
Sorav Header Caos .
- 2. RHR Containment glevated buo- See Attached Sheet.
Spray Header Caps pression Pool Temp.
i 1
' I Gentral Reference Material Review l FSAR CALQJLATIONS REQH.ATORY I
$1CHDR PNPS TECHNICAL SPECS. DESIGN SPECS PROCEDURES ELIDES STANDARDS CDDES 4.8 NA/ M /7312 af-#0-/ 10CFR50 s3 GQa,. & ik-7-o.r v 14 A
- 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).
Prepared by N Date Sfrfs 7 NOTE: It is a requirement to include this work sheet with the Safety Evaluation.
Exhibit 3.07-C l 3.07-18 Rev. 4
- . , 9_ c ; 3 , _ --
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Safety Evaluation No. 113 3 Page / of 32
?
SAFETY EVALUATION HORK SHEET A. System / Structure /Comoonent Failure and Consequence Analyses The following apply only to the proposed modification per PDC 86-52A:
- 1. System / Structure / Component: RHR Containment Spray Header Caps 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 rate.
l
- 2. System / Structure / Component: RHR Containment Spray Header Caps !
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 exists for operator '.o 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.
I Comments: RHR water is maintained free of particulate and i 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
- _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ l
1 Safety Evaluaticn l SAFETY EVALUATION No.: () 4 fj3 PILGRIM NUCLEAR power STATION 4
Rev. No. ()
A. APPROVAL 597 8 oF 32 QQ This proposed change does not involve a change in the Technical Specifications.
(k) This proposed change, test or experiment does ( y does not Od involve an unreviewed safety question as defined in 10CFR, Part 50.59(a)(2).
. (M This proposed change involves a change to the FSAR per 10CFR 50.71(c) and is reportable under 10CFR50.59(b).
() Comments: fidreenw) ewdr hoar-J ualn us /.e
& A ,, g cew. ee sognr *Mm The safety evaluation basis and conclusion is:
6d Approved () Not Approved Ju P in fb'Yn Discipline Grdup Leader /Date k S/~S%P7 Supporting Biscipline Group Lander / Bate i
l B. REVIEW APPROVAL
() Comments: -
YW 5654 W'a**r/****
U C. ORC REVIEW
() This prope' sed change involves an unrevisued safety questian and a request for authorization of this change must he filed with the Directorate of Licensing, IEC prior to implamsstation.
() This proposed change does met involve an unraviewed' safety l question.
ORC Chairman Sata ,
OSC Meeting Nuidwr cc:
Exhibit 3.07-A Rev. 3 Sheet 2 of 3
, . . .. .. .. . e .
.. iz: . .::
PILGRIM STATION Shaat '7 of JL FSAR REVIEW SHEET
References:
PDC 86-52A Safety Evaluation: 1/f3 Rev. No.: O Date: NJ /
Support a change to the RHR Containment Spray Caps List FSAR test, diagrams, and indices affected by this che.nge 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.
- Table 14.5-1 Attachment r Fig. 4.8-2, ,
.5.2-1 to 8,. -
Attachment 1 6.4-3, &'14.5-5 to 8 Attachment i
~
PRELIMINARY FSAR REVISION (to be completed at time of Safety Evaluation preparation).
Prepared by:N d od'/Date: T fi /Date:d ~7 i 87Reviewedby[:
Approved by de A r//Date: J /
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).
l Exhibit 3.07-A l
Sheet 3 of 3 I
1 .
3.07-15 Rev. 4 i
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j Safety Evaluation No. d /33 Sheet te of_,_]2_
ATTACHMENT 1 RECOMMENDED FSAR CHANGES l
l, 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, i t . 5 ~' , 14.5-S I
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l 'Coolent flow to th2 RHRS heat exchangers from ths R2ector Building Closed Cooling Water System is n:t required immedictoly af ter a during is not necessary LOCA
- because heat rejection f rom. the containment ,/, l 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 Subsystem is an integral part operation to remove heat from the pressure suppression This pool system to is
- reduce pressure in the primary containment following a LOCA.
also operated as required during planned operations to control suppression . pool water temperatures within the limits assuried in the
- Station Safety Analysis.
With the RHRS in the SPC mode of operation, the RHRS pumps are 1 aligned to pump water from the suppression pool through the RHRS heat j 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 SPb mode functions to transfer heat from the primary ;
- - contairunent to the Reactor Building Closed CoolingItWater System is concluded thereby lowering the primary containment pressure.
safety design bases 2 and 3 are satisfied by this mode of RHRS operation. In the event of reactor vessel isolation, the RHRS in the SPC mode is capable of maintaining the torus water temperature It is concluded that_ below 130*F for at least 2 hr of RCICS operation. of RHRS power generation design basis 2 is satisfied by this mode
- A operation.
.. . l
- 4.8.5.5 Containment Spray Subsystem
/
i The Contairement Spray Subsystem provides containment spray capability
- J N, h o b%as anTheQwater .
alternate method pumped forthe through reducing RHRS beat containment exchangers canpres be 1 P diverted to spray headers in the drywell and above the suppression M.
RHRS M
'"'*d+h C pool. - Thehspr exist in te drywell ygneauersgin thereby loweringthe drywell containment condense any steam pressure.
in the bottom of the drywell until the water level he CAchgew spray collectslevel of the pressure suppression, vent pipes where it
{j,y g g rises to the and overflows, 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 The Contaisunent Spray in the above the suppression pool.
free volurne remove energy from the drywell by condensing steam,l Subsystem will thereby making available the drywell voltane to accommodate additiona quantities of gases frors any postulated metal water reactions above that which the containment can inherently accommodate without spray.
The containtnent spray mode of the kHRS 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.
The. Fama mm > Pod'oh t'ebs fo the. Swpphe.Ss'8 oh poo
- v. 4.8-5 Y al the. Suppgsg',oh pool 67 pen
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-PNPS-FSAR i- - The Reactor Building Closed ' Cooling. Water System (RBCCWS) piping supplying the drywell coolers will be revised to seismic Class I to p maintain the pressure boundary integrity of this piping under seismic a loading. Refer to Section 10.5.5.1. The .drywell coolers were originally purchased as seismic Class I equipment to serve as pressure boundary only.
The PCS design . loading ~ considerations are given in Section 12 and Appendix C. The Station Safety Analysis presented .n Section 14 demonstrates the effectiveness of the PCS as a radiological barrier.
p In addition, primary containment pressure and temperature transients-from postulated DBAs are also presented in Section 14.
h 5.2.3.2 Drywell l
I. The drywell is a steel pressure vessel with a sph'erical lower L portion, 64 ft in diameter, and a cylindrical upper portion 34 ft 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 j ASME Code,Section III, Code Case M-1312-(2), the maximum drywell
[m pressure is 62 psig. Thermal. stresses in the steel shell due to
{
temperature gradients are taken into account in the desip.
t Special , precautions not required by codes were taken in the fabrication of the steel drywell shell. Charpy V-notch specimens were used for impact testing of plate and forging material to give assurance of proper material properties. 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 defomation 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.
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 door may be operated unless the other door is closed and locked. The 5.2-3 i
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ . . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ . - 1
/4 j y PNPS-FSAR drywell head and equipment hatch covers are bolted in place and sealed with gaskets. l t
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 containment 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 dryvell.
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 Far tha covering the'same size range. *h: *i-- it ta-- -t:'-
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. we. 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.*aF**="h 5 ; :.;-i daws de no vi im.. L.;;h
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, mii *mam. . ..a. R."7 would be ef fective 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 transfer to the subcooled spray is very effective. 'Since the total amount of heat in the drywell atmosphere is low relative to the spray rate, the- containment atmesphere temperature is quickly reduced to near the spray temperature.
l A drywell pressure condition exceeding 10 psig was selected as the l basis.for determining when to initiate the containment spray. See Figure 5.2-3 for. time required to reach 10 psig. L.d ;;. G.,
l ;u:_ 1 f ri; = :.: ?,'Trie operator will be instructed to initiate the containment sprays if containmer.t 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 ~1.1.8.
5.2.3.3 Pressure Suppression Chamber rnd Vent System
. 5.2.3.3.1 General The pressure suppression pool, which is contained in the pressure suppression chamber, initially serves 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.
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 releases to the drywell.
4 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 dryvell 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 l l
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5-7. Figure 4.5-5 shows 45 psig, which is well ig. After the discharge
' el into the drywell, the approaches 130*F (Figure e stabilizes at about 27 the noncondensibis gases.
er during the vessel noncondensibles soon :
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/'k e RHRS pumps (low pressure
- ding .itquid to the reactor FIGURE 5,2-jI er the reactor vesse nozzles, the eXC855 TIME TO REACH 10 PStG IN ORYWELL i break into the dr IIf.
g FOR VARIOUS SIZE STEAM LEAKS fuel, offers cons PILGRIM NUCLEAR POWER STATION E,,,, , ga \on of the FINAL 8AFETY ANALYSIS REPORT ts condensed. At 600 sec, Revision 6 - July 1986
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FNPS-FSAR l
the RHRS pumps are assumed to be switched from the LPCI mode to the 8 containment cooling mode. The containment spray would normally not be activated at all and the change over to the containment cooling
(
mode need not be made for several hours. There is considerable time available to place the Containment Cooling System in operation l 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 priaktry containment was analyzed for the "ollowing containment spray and cooling conditions:
I 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 i mode. No containment spray. -
t Case C - Operation of one RHRS 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 see after break) is the same for each of the conditions. During the long h
term containment response (after depressurization of the reactor vessel 1s complete) the suppression pool 1s assumed to be the heat h sink in the containment system. The effects of decay energy, stored energy, and .tnergy. from the metal water reaction on the suppression f 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 RHRS
- pumps, and design values of cooling water flow to both RHR loops
! operating in the suppression pool cooling mode. The RHRS 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 s cooling condition is arbitrarily assumed to start at 600 see af ter 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 "a" and on suppression pool temperature responses are shown as curves Figures 14.5-6 and 14.5-7, Af ter the initial rapid depressurization, energy addition due to core decay heat results in a gradual pressure 3 and temperature rise in the containment. When the energy removal b 14.5-8 Revision 6 - July 1986
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PNPS-FSAR N p O rate of the RHRS exceeds the energy addition rate from the decay d 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 The assumed to be activated at 600 sec after the accident.
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 cur.ves "b" on Figures 14.5-6 and 14.5-7. A summary of this case is shown on Table 14.5-1, including a summary of the equipment capability assumed in the analysts.
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Case C q gl,73 gM o vi e SvfFtW86 This case assumes the same equipmen erability as Case 8 exce that & discharge from the RHR eat exchanger is routed de ,
j- containment spray m e of operattorrf It assumed that the containment spray is established at 600 sec after the accident.
j ~
The containment response to this set of conditions is shown as curve "c" on Figure 14.5-5. The corresponding drywell and suppression. poolA -
' temperatures are shown as curves c on Figures 14.5-6 and 14.5-7.
summary of tMs case is shown on Table 14.5-1, including a sumary ofJhcANg - g the equipment capability assumed in the analysis. c emah[. N e, Comparing the " containment spray" Case C with the "no spray" Case B Mflow it is seen that the suppression pool temperature response i the name_ y g e,
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l blowdown containment pressure is higher for the "no spray" case, as 4 he.
shown by Figure 14.5-5. This, however, is of no consequence since wmc the pressure is still much less than the containment design pressure g* gg I'
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 Systes Pump Net Positive Suction Head To assure proper operation of the RHRS circulating and reactor CSCS f pumps following a design basis LOCA, precautions are taken to ensure that a net positive suction head (NPSH) margin is available to ail above pumps at all times.
O 14.5-9 Revision 6 - July 1986
c.- = ;- v.m .: , . , ,_ .
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It is reconrnended that curve c in Figures 14.5-5 t,brough 14.5 8 be left unchanged. The reason is that E
time range of6 application of the drywell spray, in the(from600 seconds to 10 seconds), it is expected that the drywell l spray flow rate of 300 ppm and 5000 gpm 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 short period following the initiation of the drywell spray.
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PNPS-FSAR Table 14.5-1 I LOSS OF COOLANT ACCIDENT PRIMARY CONTAINMENT RESPONS,E SPRY Case A __ Case B 1 Case C -_
Se_condary Peak Pressure, pstg None 13.1 Peak Pool Temperature, 'F ' 8.0 142 166 166 Mode of RHR5 operation Suppression Suppression Containment Pool Cooling Pool Cooling Pool Cooling Number of RHR loops operating 2 1 1 Number of RHR pumps operating 2 1 1 Number of RHR heat exchangers operating 2 1 1
Total RHRS flowrate, gal / min 10,000 5.000 5,000 Core Spray System flowrate sal / min 3,600 3,600 3,600 Containment Spray System flowrate, gal / min 0 0 RWR5 heat exchanger flow. ' CS^- 3 e
- 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' '
Number of RBCCN loops operating 2 1 1 Number of RSCCW pumps operating 4 2 2 Number of RBCCW heat exchangers operating 2 1 1 Total RSCCNS flowrate to RHRS, gal / min 5,400 2,700 2.700 Number of 55W loops operating 2 1 1 Number of SSN pumps operating 4 2 2 Total 55W flowrate to RBCCWS 10,000 5,000 5,000 55W inlet water tempera-ture, 'F 65 65 65 ressure steadily decreases after containment cooling is established.
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ShKET 3 1 o r- 3 2 JUSTIFICATION FOR CAFFING 6/7 DRYVELL SFRAY N0ZZLES There are two primary issues related tc the magnitude of the dryvell i spray flovrate. These are
- 1) The integrity of prikary containment and operability of dryvell equipment following the dryvell pressure-temperature I transient due to dryvell spray initiation. This issue relates j to the use of the large-capacity RER systen to supply the dryvell sprays.
- 2) The capability to effectively scrub airborne fission products that may be present when or after the sprays are initiated.
This issue relates to the use of the relatively low capacity fire water pump to supply the dryvell sprays. These two issues vill now be discussed. Dryvell shell and equipment response to dryvell spray initistion is a sensitve function of the dryvell spray flovrate. The current config-uration of the dryvell spray system results in a maximum dryvell spray
- flovrate of approximately 10,000 GPM, using the RER systaa.This 1srge flovrate can produce a very severe depressurisation transient in the dryvell when the sprays are initiated, depending on the dryvell conditions at initiation. This transient response in tne dryvell may jeopardise the integrity of the dryvell and /or the equipment in the dryvell, as well as affect the operability of the dryvell equipment. Although the proposed initiation limit curve is designed to protect the integrity of the dryvell pressure boundary, nominal shell/ equipment condition is assumed.
Consequently, if the condition of any equipment or the shall is degraded, it is possible that the spray initiation transient could challenge it. Accordingly, reduction of the maximum dryvell spray flovrate provides additional safety margin to that in the spray initiation curve. One way to reduce this flow is to increase the resistance in the flow path by reducing the flow area. l The use of the fire water system to supply the dryvell sprays does not have the same type of concern associated with it. Here the concern is that the large flow area provided by the 7 x 104 nossles does not allow the fire pump to pressurise the header sufficiently to ensure atomization of the flow. That is, the flow from the dryvell spray header would be a trickle rather than a spray. This type of flow from the spray nozzles vould greatly reduce the capability of the sprays to l scrub any airborne fission products that might be present in the dryvell atmosphere. Although staan genersted by water cooling core debris can carry airborne fission products and deposit them in the suppression pool, this mechanism is only effective provided the l containment is intact, whereas sprays effectively scrub regardless of containment condition. Consequently, it is desirable from I this standpoint to increase the pressure drop across the nossles to ensure atomization of the spray flow when the fire water pump is its
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source. Again, a way to increase the pressure drop across the spray nozzles is to reduce their flov area. For both issues related to the sprays, reducing the spray flov area has been shown to reduce the potential concerns. This justifies the reduction in spray flov area caused by capping 6 out of 7 nossles.
'The rationale for espping 6 of 7 nozzles on each spray head, rather than 2 or 5, or all the norries on 6 of 7 spray heads, or etc. vas to preserve the symmetry of the spray pattern ao much as possible. Although significant flow reductions occur for this configuration, these reduced flows have been shown te provide satisfactory cooling of core debris (should it be expelled from the vessel) and scrubbing of airborne fission products.
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paxw7:4 m a y w w .::: a w m + " m : , n ic.,. m ;' n.:~ .: .: L ; ~ m= T ;; [: o . y. l' i; < l p AGENDA W DECEMBER 10, 1987-f- .>
- 1. E0P STATUS UPEATE L
- 2. ' DESCRIPTION OF-PILGRIM E0Ps
- 3. QUESTIONS AND DISCUSSION TWN h/h621TY - [#CD l
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UPDATE OF PNPS E0P STATUS SINCE AUGUST 8, 1987' SIX o COMPLETED VALIDATION ON PLANT SPECIFIC y TOR INTERIM VALIDATION ISSUED VR2 IMT740 /befn* A&r.6w/40 M cLU OfA16- fietna axe PApossvM o COMPLETED OPERATOR TRAINING , 40 HOURS OF CLASSROOM Wpf 40 HOURS OF SIMULATOR l l 0 . COMPLETED AN UPDATE OF THE DRAFT E0Ps INCORPORATED SIGNIFICANT VERIFICATION AND VALIDATION FINDINGS, AND OPERATOR COMMENTS REVIEWED, APPROVED AND ISSUED 0 COMPLETED AN UPDATE OF ALL E0P SATELLITE PROCEDURES
^
- twn.sgeir-ic U PGPNE TD M TCH-
__.h0 -k syrewin Paawv43 o PROVIDED AN UPDATE OF ASSOCIATED UM NTATION TO NRC 0 FUTURE PLANS l 1
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t o FUTURE: PLANS REVISE PSTG AND E0Ps FOLLOWING FORMAL NRC APPROVAL 0F REV. 4 EPGs FINALIZE E0P VERIFICATION AND VALIDATION FINALIZE PROCEDURE GENERATION PACKAGE AND SUBMIT TO NRC FOR REVIEW V* / h,*D(s' af g #,h f phsk,pf/ yp## l l h-____m__ --__a .__ - _ _ _ _ _ _ _ _ . .
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o CRITERIA FOR DEVELOPMENT OF FLOWCHART PROCEDURES o Procedures must accurately reflect the technical. content and intent of the PSTGs. o Format must accommodate sequential, concurrent, and oyerride actions. LP#Mp/y eV o Text of instructions and decisions should be simple and unambiguous. o Procedures must include all cautions, figures, and tables required for execution of operator actions. o Overall arrangement should be clear, easy to follow; crossover of lines should be minimized. 1 o Text must be readable, o Sheet size must be no larger than 36" x 60". o Entries to and exits from each procedure must be clearly , and unambiguously presented. I
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NUREG-0899 Sections PNPS EOPs
; (Writers' Guide & Busan Factors) (Writers' Guide Section) 5.2 General Guidance 5.2.1 Consistency Among the Procedures III.D,E,F, & IV.A,B,F 5.2.2 Cross-Referencing Within and Among Procedures III.A.11 & IV.D
.5.2.3 Operator Aids III.A.12 5.3 Presentation of Information for Readability III.D,E,F & IV.B 5.4 Organization of EOPs 5.4.1 Cover Page N/A 5.4.2 Table of Contents N/A 5.4.3 Scope N/A 5.4.4 Entry Conditions III.A.1 5.4.5 Automatic Actions III.A.4 5.4.6 Immediate Operator Actions III.A.4 5.4.7 Subsequent Operator Actions III.A.4 5.4.8 Supporting Material (Attachments) III.A.12 r
5.5 Format of EOPs 5.5.1 Identifying Information II.A,B,C 5.5.2 Page Layout III.E 5.5.3 WARNING, CAUTION, and NOTE Statements III.A.13,14 5.5.4 Peacekeeping Aids (Training) 5.5.5 Divisions, Headings and Numbering III.A.3 & III.B 5.5.6 Emphasis III.D l 5.5.7 Identification of Sections Within a Procedure or Subprocedure III.A.3 & IV.D l 5.5.8 Figures and Tables III.A.15,16 5.5.9 Use of Flowcharts N/A 5.6 Style of Expression and Presentation 5.6.1 Vocabulary IV.G 5.6.2 Abbreviations, Acronyms and Symbols IV.G 5.6.3 Sentence Structure III.A & IV.A,B 5.6.4 Punctuation IV.F 5.6.5 Capitalization IV.D 5.6.6 Units IV.H i 5.6.7 Numerals IV.H I 5.6.8 Tolerances IV.H 5.6.9 Formulas and Calculations IV.B 5.6.10 Conditional Statements III.A.5,6,7,8 , i
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l PILGRIM NUCLEAR POWER STATION EMERGENCY OPERATING PROCEDURES : EOP-01,"RPV CONTROL" EOP-02, " FAILURE TO SCRAM" EOP-03, " PRIMARY CONTAINMENT CONTROL" EOP-04, " SECONDARY CONTAINMENT CONTROL" EOP-05," RADIOACTIVITY RELEASE CONTROL" EOP-06,"RPV FLOODING" EOP-07," ALTERNATE RPV DEPRESSURIZATION" EOP-08," STEAM COOLING" EOP-09, " PRIMARY CONTAINMENT FLOODING" 1 I
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SHEET # EOP#~ PSTG SECTIONS y 1 1. 3 - Entry Conditions J RPV Water Level RPV Pressure Reactor Power 7 - Alternate Level Control 8 9 - Steam Cooling i 2 2 3 - Entry Conditions ; RPV Pressure l Reactor Power 11 - Level / Power Control 3 3 4 - Entry Conditions Suppression Pool Temperature ! Drywell Temperature Primary Containment Pressure Suppression Pool Water Level Primary Containment Hydrogen Concentration 4 4 5 - Entry Conditions Secondary Containment Temperatures Secondary Containment Radiation Levels Secondary Containment Water Levels 5 6 - Entry Conditions Radioactivity Release 5 6 10 - RPV Flooding i 7 8 - Emergency RPV Depressurization 9 12 - Primary Containment Flooding CORRELATION BETWEEN EOPs and PSTGs
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- KEY FEATURES OF FLOWCHART PROCEDURES
$ 0 PRESENTS ON A SINGLE SHEET OF PAPER ALL INFORMATION h REQUIRED FOR SUPERVISORY PERSONNEL TO DIRECT EMERGENCY RESPONSE ACTIONS t L t. O BEST FORMAT FOR MAINTAINING A BROAD PERSPECTIVE-0N OVERALL PLANT CON 7;rIONS AND ASSOCIATED OPERATOR ACTIONS 1 0 ORDER OF CONDITIONS / ACTIONS IS VERY CLEARLY PRESENTED SEQUENTIAL BRANCH CONCURRENT o FLOWCHART ELEMENTS ARE SELECTED TO REINFORCE THE
" UNIQUE" TYPES OF ACTIONS DECISIONS OVERRIDES CONDITIONAL ACTIONS WAIT UNTIL....
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, Alternate RPV Depressurization Exit this procedure and enter !
is required EOP-07, " Alternate RPV [ Depressurization." p RPV water level cannot be Exit this procedure and enter ! determined EOP 07,"ARernate RPV Depressurization." Av see f injecti n is lined Exit this procedure and enter 8-2 up to the RPV with at least one EOP 07, *Ahernate RPV l mp running Depressurization." I Y l V AIT UNT RPV water V 8-3 level drops to -169 in. Exit this procedure and enter EOP 07,
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4 ACTION PLAN ISSUES
SUMMARY
A) EVACUATION OF SCHOOLS SUB!SSUE ACTION A1) Identify all private schools Identified in the updated ETE and and daycare centers within the EPZ. verified by Boston Edison. , A2) Develop plans and procedures Plan revisions should be complete by for evacuating schools and daycare about 10/31/87. Corresponding centers. procedures will be revised. i A3) Develop plans and procedures Plan revisions should be complete by for early evacuation and dismissal. about 10/31/87. Associated procedures will be revised. School and daycare center evacuation is reflected in the updated ETE. A4) Obtain Letters of Agreement Planners are assisting with the with transportation providers. identification of transportation providers and will assist in securing letters of agreement. i A5) Train drivers as emergency Introductory training modules are workers, being completed and will be ' forwarded to the Commonwealth for review. A driver training module will also be developed. Planners will be available to assist in training program implementation. l I l l 1 ll q
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l ATTACHMENT B OFFICE MEMORANDUM BOSTON *uDISON COMPANY , To: ' Distribution FROM: R. J/ Barrett Record Type A4.08 DATE: January 29, 1988 Dept Doc. PM 88-062 i
Subject:
DESCRIPTION OF POWER ASCENSION TEST PROGRAM J DISTRIBUTION: R. G. Bird K. L. Highfill R. N. Swanson R. A. Ledgett E. J. Howard D. L. Gillispic R. J. Barrett All Section Managers I. PURPOSE: The purpose of this document is to define and establishTest the following in the temporary Post-Refueling / Outage Organization: A. Define the relationship of the Test organization to operations organization. , B. Definition of the organization and description of the ' duties and responsibilities of personnel assigned. C. Describe the scheduling of testing. D. Describe the conduct of testing. E. Describe the independent review of completed test data. II. BACKGROUND: Experience from the testing program after RFO-6 and prudence dictate that the conduct of post-outage testing be carefully planned and , controlled in order to ensure completeness, The following sections of this ; safety and affactiveness. program description document define the scheduling, conduct, J
, and review of testing to be accomplished during startup from RFO-7.
C u a'. E Pv 7 Page 1 of 12 g 4 FEB 5988 l l FOR USE NTRot, ET ATIO
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a Attachment S con'to
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Page 2 . Date: January 29, 1988 III. RELATIONSHIP OF THE TEST ORCANTIATION, TO THE NUCUAR WATCH l ENGINFFR (NWE) AND HIS SECTION CREW The NWE is the senior licensed operator on watch and has The Test complete responsibility for. operation of the plant. The NWE's organization is subordinate to the NWE. concurrence and permission must be obtained for the conduct of all tests. Direction must not be given to watch section it operators for operation of plant equipment except as The pertains to conduct of previously authorized tests. Shift Test coordinator (STC) must work closely with the NWE, may advise him, but must exercise exceptional care to not interfere with the NWE and his operators in discharge of their license and operational duties. IV. STARTUP TEST ORGANIZATION following are descriptions of the duties and The responsibilities of defined positions in the Startup Test organization. The Startup Test Manager may assign additional personnel to perform specific tasks in addition to the positions described herein. An organizational diagram of the Startup Organization showing the relationship of the temporary startup test organization and the permanent station , organization is shown in Figure 1. I
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j l Page 2 of 12
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ATIACN1ENT B can't. ( STalrfUP TERT oRCAMIEATION POSITION RESPONSIBILITIES i c A. STARTUP. TEST _ MANAGER (J. A. SEERY) 1 Major duties and responsib'ilities of this position'are: )
- 1. Development of.the Power Ascension Schedule.
- 2. Development and staffing of the Startup Test Program.
- 3. Development of a master power ascension test procedure.
i
- 4. Ensuring ' all test procedures required for the startup '
and testing are approved. ;
- 5. - Day to day management of the Power Ascension Program keeping the Plant Manager informed of progress and problems. .
- 6. Approval of completed startup test reports.
- 7. Recommendations to the Plant Manager when the plant is ready to ascend to the next test plateau.
- 8. Scheduling of testing and promulgation of the test schedule.
- 9. Interface with the Operations Section Manager to ensure testing and normal plant operations are efficiently scheduled.
- 10. Direction of the Shift Test Coordinators.
- 11. Ensuring complete test data review.
- 12. Approve a waiver of a discre ry test in TP 87-114.
B. MSISTANT STAIPfUP TEST MANAcER ( . . NICHOIAS)
. r is This position reports to the Startup Test Manager and a responsible to him for the following duties:
- 1. Development of test reports, as necessary.
- 2. Technical advice in development of the schedule and necessary changes.
development of a startup test
- 3. Assistance in organization and position responsibilities.
Page 3 of 12 , i
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f- 1 ATDONENT B Wt. I
- 4. Assistance in development of a ma ter power ascension '
procedure. .
- 5. Assistance in the day to day management of the Power Ascension Program as necessary. l
- 6. Ensuring that test data is reviewed for acceptance by >
the proper technical review group, I
- 7. Technical and comp 1ateness review of completed test reports and submission to the Startup Test Manager for approval.
C. REPORT WRITERS This position reports directly to the Assistant Startup Manager. Their duties are as follows:
- 1. On a daily basis, collecting completed tests from the Shift Test Coordinator
- 2. Reviewing test for completeness i
- 3. Preparing test reports for the Assistant Startup Manager's review
- 4. Assisting the Assistant Startup Manager as required (P. SMITH, J. BELLEFEUILLE, P.
D. SHIFT TEST COORDINATORS MANDERINO, J. SABINA) Shift Test coordinators will maintain shift coverage during the testing phase. They will assist the on-shift Nuclear Watch Engineer by coordinating all testing for that shift and obtaining his permission to conduct testing. Their duties will consist of, but not be limited to the following: are conducted in a formal and
- 1. Ensuring tests professional manner.
r
- 2. Ensuring tests are conducted using approved procedures.
- 3. Assisting the NWE in the conduct of pre-test briefings with the shift crew as necessary.
- 4. Keeping abreast of current plant status.
- 5. Ensuring test procedures are properly controlled.
- 6. Maintaining a shift log of the testing status progress on a shift basis.
1 Page 4 of 12
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' . . 4- L* .-- ArrAONENT B con't conjunction .w ith the Nuclear Watch Engineer,
. 7. In recommending changes to the suggested critical path schedule when necest.ary.
- 8. In conjunction with the Watch Engineer, recommending waiver of discretionary tests in TP87-ll4.
- 9. Keeping. the 'startup Test Manager informedofin problems- a timely manner of the status of testing and encountered.
E. REACiGR EiiGINTERS , (J. ABOLTIN) dartrrr.x ENGINEERS. TEST ENGIhrr4 , wxBINE TEST ENGIN*rRS. AND TEST AND TuxiiOVER ENGINEERS are The generic responsibilities of these positions They are similar. These are the Test Directors. responsible for directing and conducting their assigned tests. On shift they report to the Shift Test Coordinators. Their duties will consist of, but not be limited to, the following:
- 1. Directing performance of their approved test after NWE concurrence and permission has been obtained.
- 2. Keeping the Shift Test Coordinator and Nuclear Watch Engineer informed of test status and affects on plant conditions.
- 3. Assisting in pre-test briefings with the shift crew as necessary and directed by the NWE.
V. NUcT.RAR ORGANIZATION SUPPORT FOR OPERATIONS The restart fron. RFO-7 will be achieved with the normal plant Quality support functions, e.g.,. Nuclear Engineering, Assurance / Quality Control, Security, Fire Protection, Planning / Scheduling and Cost Control, Technical Support, and othar nuclear organization groups. ,, l A. Due to the complexity of this outage, there will be a Shift
- Maintenance Representative, as shown and described in BECo Ltr. 87-163.
B. Quality Assurance coverage, although functioning through normal channels will be expanded during the preparation and testing phases of power ascension. Around the clock coverage will be provided by approximately 9 dedicated inspectors. A QA checklist developed specifically for Dedicated coverage of power ascension will be used. chemistry will also be provided. Page 5 of 12
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ATDONENT B con't l VI. CONDUCT OF TESTING j A. This section descr,bes the mechanism for the conduct of P testing during the Power Ascension Program. B.
REFERENCES:
(Latest revision unless otherwise j l indicated) - and
- 1. ANSI N.18.7-1976 Administrative Controls Quality Assurance for the operational Phase of Nuclear Power Plants
- 2. ANSI N.45.2.6-1978 Qualifications of Inspection, Examination, and Testing Personnel for Nuclear Power Plants
- 3. ANSI N . 4 5. 2.11-197 4 Quality Assurance Requirements for the Design of Nuclear Power Plants
- 4. BECo Quality Assurance Manual, Volume II, Section II
- 5. PNPS Technical Specifications, Section 6.8 !
- 6. Procedure 1.3.4 - Procedures
- 7. Procedure 1.4.5 0 PNPS Tagging Procedure
- 8. Procedure 1.5.3 - Maintenance Request C. pgTINITIONS:
- 1. PDC - Plant Design Change
- 2. ORC - Operations Review Committee
- 3. DCC - Document Control Center
- 4. SRO - Senior Reactor Operator
- 5. MR - Maintenance Request y
- 6. RWP - Radiation Work Permit
- 7. TP - Temporary Procedure 1 f
i
- 8. TD/TE - Test Director / Test Engineer. The TD/TE is the individual responsible . for coordinating the performance of a test after permission of the NWE has been obtained.
Page 6 of 12 ] I
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'- _ATTA09ENT 8 can't.
- 9. STC - 'Shif t ' Test Coordinator. The STC is the individual'who is responsible for management of all testing conducted - on his shift. An STC cari The have STC several Test Engineers reporting to him.
reports to the Startup Test Manager.
- 10. PNPS'- Pilgrim Nuclear Power Station D. PREREOtTISITES:
Test Manager will assign Shift Test The Startup and-Coordinators (who meet appropriate educational experience requirements of ANSI N.45.6-1978). He will-certify each STC for the appropriateSheet. level using the These Test Director Qualification certifications will be maintained on file within the
' Startup Test Organization.
E. TEST PREPARATION:
- 1. Upon authorization by the NWE of tests scheduled for that shift, the Shift Test Coordinator can make preparations to start the test.
- 2. Prior to the actual start of the test, the STC will ensure the following:
- a. That a latest revision of the procedure has been obtained from the DCC.
- b. That the procedure to be used is the same revision as the one in the Control Room and is marked " Working copy".
- c. That an approved Maintenance Request is ready for use (if needed). -
- d. That an RWP has been approved (if needed) and is ready to support the test.
- e. That plant operating conditions and system status required for the test have % en set.
needed are l
- f. That manpower and equipment scheduled and available.
prerequisites stated in the
- g. That all prerequisite section of the test procedure are satisfied. 1 Page 7 of 12
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ATDONENT B can't. - ) P. conDDcT pr TzsTIna:
-1. During the conduct of' the test, the Test Engineer will ensure that the personnel conducting the test .
are cognizant. of ' all requirements of the test such as:
- a. All prerequish.tes listed under Section VI of the procedures " Prerequisites".
- b. All personnel initialing the procedure will sign- their full name on a Signature Identification Sheet.
- c. Any calibrated equipment which is part of the Meter and Test Equipment (M & TE) System being used is listed on an equipment log. If the equipment being used is not part of the M & TE be System, a calibration Data Sheet will attached as an addendum to the log.
- d. Test procedures, unless specifically stated, do not permit any exceptions. Any changes to the procedure will be in accordance with PNPS 1.3.4.
- 2. If a problem is encountered during the conduct of a ;
test the following steps will be taken:
- a. Hardware eroblems - If a hardware problem is encountered, the TE will stop the test. A trouble shooting MR will be initiated by the TE. All necessary manpower will be obtained for trouble shooting. The scope of trouble shooting will be to.the extent of; identifying the components affected. Problems identified will be repaired using normal station repair procedures. After the problems are repaired, the test may again be schedule by the Startup Test Group for completion. A%g
- b. software nrobl=== - If the probleth encoudte' red is procedural, the Test Engineer will stop the
- test, obtain the necessary. change to the Test Procedure in accordance with Station Procedure 1.3.4. The Test Engineer may resume the test the changes are made, and with the l after approval of the Shift Test Coordinator and the NWE.
- c. The TE is responsible for timely notification of the Nuclear Watch Engineer and' the STC of any problems encountered and of completion of l the test.
Page 8 of 12
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ATTACINDfT B oan't.
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M222: In sections a. and b. above, if the test procedure permits the TE to do other l sections which are independent of the-section being reworked (hardware) or revised l the TE can continue with the (software), other independent portions of the- test with the concurrence of the STC and NWE.
.G. TEST VERIFICATION:
- 1. After the test is complete, the TE will review the test to ensure the following:"
- a. That the test was conducted as written without
( any discrepancies or exceptions.
- b. That the acceptance criteria of the applicable steps have been satisf actorily. met.
- c. That the QC Group has reviewed all their hold witness points and has signed off and appropriate steps in the procedure. !
- 2. When the above steps are completed, the TE can then sign Section XI of the procedure for Test Completion. The TE will then submit the test to the STC for his review.
- 3. The Shift Test Coordinator will review the entire tesi. procedure to ensure that:
- a. All aspects of the test have been completed in accordance with station procedures.
- b. All revisions of the procedure used during the test are available and attached to the final revision.
- c. After Sections G.3.a and G.3.5 aboYe, 'are completed the STC will sign Section XI, of the
- procedure if required, of the procedure for the STC. The procedure will now be presentedand to the Nuclear Watch Engineer for review approval for completeness.
- d. After the Nuclear Watch Engineer has signed the procedure, the procedure will be sent to the ,
Assistant Startup Manager. ) Page 9 of 12
L :f. (_' '* .- . , , f , .' n ATDONDE B can't.
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' 4.I The Assistant Startup Manager 'will coordinate the Submit to si independent review of the test packages. jthe Startup Test Manager a corpleted y, including the independent reviewtest package results for his eg approval.
- 5. The Startup Test Manager will review the test B/ M f/ W A procedure along within theSection results cfVII the independentAf below. ter 7D #'$
review described approving the test, he will attach the independent //jbeu review to the procedure and transmit to Doc."nent Control for capture. VII. INDEPENDENT REVIEW OF TEST RESULTS An independent review of the test results obtained during the Power Ascension Test Program will be performed by the Systems Group. The support of personnel from the other groups such as NED and Operations will be required as appropriate for their specific expertise. but not be limited to the The review will include following:
. Verification that the test was performed as written.
. All data was taken and recorded as required.
. All acceptance criteria were satisfied.
After completing the independent technical review of a completed test procedure the Systems Group Leader will notify the Startup Test Manager in writingand of the findings. The Startup Test Manager will review collect all independent review sheets. After all applicable procedures and reviewed, the have been satisfactorily performed Startup Test Manager will submit a commitment close-out memo to Comp 1L oce. VIII. ScnruiJLING OF TESTING A. The Startup Test Manager is responsible for schedul'ing of testing and promulgation of the test 's'chedule.MTo will do the discharge these responsibilities he
- following:
- 1. Schedule testing on a shift basis. When tests are required to be done in a specific sequence, that in the constraint will be clearly indicated schedule. In forming the test schedule the Startup Test Organization and the Operations Sectionplant will coordinate to ensure safe and efficient operation.
Page 10 of 12
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- ATTAOMENT B can't.
'2. The Power Ascension Test Schedule will be approved by the Operations ' Section : Manager. This will be integrated inte a Plan of the Day The (POD) PODwhich is will be approved by the Plant Manager.
issued for a specific period of time, with an effective date/ time end an expectation date/ time..
- 3. The testing schedule , may be extended or changed with formal concurrence.of the Startup During Test Manager periods and the Operation Section. Manager.
i when either or both of these managers are not on site, the STC may obtain concurrence by telephone from the absent manager 'and approve test schedule changes for these managers "per talcon" noting the i date and time telephone concurrence wasappropriate obtained. manager will sign the The' absent
' approval during the next day the manager is on site.
IX. COMMUNICATIONS DURING TESTING A. Communications during testing wil A be_ conducted in a good model for use ] precise and professional e conduct of'Tsiting sanner. comm icationsisSI-OP.0006j . 1 cations equipment to b= W in testing will be B. Coma tested for operability whenever possible, in advance of the start of the test. -Use'of the plant paging system for routine operational communications during testing no other j should be minimized and used only- when l appropriate communications paths are available. X. TESTING SUPPORT requested from various NUORG Testing support may be the . Assistant organizations by the Startup Test Manager, Startup Manager and Shift Test Coordinators. Support may be requested in addition to that specified in this document from Engineering, Maintenance and Operations. Any difficulty encountered in obtaining added Manager support shall the through be resolved by the Startup Test
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appropriate chain of command. XI. SEdur.nCING OF THE POWER ASCENSION TEST PROGRAM Conduct of the test program descriptions of major test and H schedules and notices for the test program areafter contained in RFO-7. TP 87-114, Restart Test Attachment 1, Attachment 1 shall be the governing document for scheduling and sequencingshall of testing. Changes to the intent of Attachment 1 be made only after the approval of the ORC is received, using proper procedures for such changes. Page 11 of 12
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/' "*%,, UNITED STATES l' ~' s NUCLEAR REGULATORY COMMISSION
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! OFFICE OF GOVERNMENTAL AND PUBLIC AFFAIRS, REGION I 475 Allendale Road, King of Prussia, Pa.19406 f Tel. 215-337-5330 NOTE: This mailing contains two announcements..
No. I-88-151 September 16, 1988 Contacts: K. Abraham Steve Horwitz NOTE TO EDITORS: The Nuclear Regulatory Commission has received from its Advisory Committee on Reactor Safeguards the attached report on proposed restart of the Pilgrim nuclear power station in Massachusetts.
Attachment:
As stated l
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- UNITED STATES
-{S .g ..VCLEAR REGULATORY COMMISSION Aovisony COMMITTEE ON atACTOR SAFEGUARDS
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Septemteer 14, 1988
. l The Honorable Lando W. Zech, Jr.
Chairaan U.S. Nuclear Regulatory Comission ' Washington, D.C. 205S5
Dear Chairrnon Zech:
SUBJECT:
PROPOSED RESTART OF THE PILGRIM NUCLEAR POWER STATION During September guards, the 341st meeting of the Advisory Comittee on Reactor Safe. 8-10, 1988, Pilgrim Nuclear Power Station.we reviewed the proposec restart of the i plant on August 25, 1938 andMembers e meeting of the Coninittee visited the of our subcamsnittee on Pilgrim Restart was held tatives of Boston Edison Company andin thePlymouth, NRC staf Massachusetts, f on August 26, with represen 1988. During our 341st meeting, we had further discussions with members of the NRC staff and with representatives of the Boston Edison dompany, we receiveo coments from representatives of the government.s lities, as wellofas theindividuals. Commonwealth of Massachusetts and nearby loca-documents referenced. We also had the benefit of the t By. the time of the Pilgrim plant shutdown in April of 1986, the NRC staff had developed serious reservations about the management of the plant. We therefore gave special attention to this issue. More than half of the upper and middle level manage.;:ent personnel of the plant ' have been rep' aced since the plant last operated. The new management group is made up of people of demonstrated competence. A new vice president with responsibility for Boston Edison's nuclear power program was employed about a yeer ago. He has assembled a team whose rnenibers have a variety of experience in naval and comercial nuclear power plant operation. We are favorably impressed by their creden-tials and by the changes in the physical plant and the organizational approach to operation that has occurred since the new group has been in place. Although there is a preponderance of navy as contrasted with comercial nuclear power plant experience, we found no reason for concern in the backgrounds of the present team and in the ap-preach to management and to operation that"has been inaugurated. l l
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, The Honorable Lando W. Zech, Jr. September 14, 1988 Those members of the Comittee who toured the plant were favorably impressed with decontamination of the plant which has occurred, with the way in which the operational staff is organized, and with the way in which operation and maintenance were being performed, We also examined a number of systems that have been installed in an effort to ' improve ~ the plant's and the operational staff's capability to. avoid and to mitigate the consequences of severe accidents.
During our visit to the training center, we observed the operation'of the simulator.that has been installed since plant shutdown. It is a modern -and versatile system, and appears to have been effectively integra ted into their training program. Its . capabilities are being used to train the operational staff in the application of the newly installed systems mentioned above. The simulator can be expected to have a marked positive influence on the readiness of the operational staff to deal with both normal and off-normal events. Because of the history of weather-related lun of off-site power at this site, an assured source of emergency power or a demonstrated capability to provide emergency cooling in the absence of electric power is of special importance for the . Pilgrim plant. Since the s hu tdowr. , a number of features have been added that contribute to increased safety of the plant. Among these are an additional emer-9ency diesel generator and a system that permits water from the fire protection system to be made available to the decay heat removal system. These added features should decrease the risk associated with the loss of electric power. However, we are ngt sure, and the staff and licensee are not certain, that the plant systems now satisfy the stat f an blackout rule. We recommend that the staff give particular attention to this item as the rule is being implemented. We understand that use of the hardened vent, for relieving possible torus overpressure during a severe accident, will be reviewed by the NPC staff as a generic issue for all Mark I containments. We intend i to review the matter '.n that context. ' A report, dated August 6,1987, from the Federal Emergency Management , Agency, states that the Emergency Plan which existed at the time of plant shutdown has a number of inadequacies. Information provided to us indicates that significant progress has been made in correcting these deficiencies. We recumend that before startup is approved, a clearly defined program for early correction of these inadequacies be , available and be approved by the NRC staf.f.
w u,w ama m. wa. wana m er ww.- Og( - R 4 j i The Honorable Lando W. Zech, Jr. -3 -{ September 14, 1988 I k k'e believe that, subject to the~ coments above, res tart of the l l Pilgrim sa fety. plant will not lead to undue risk to the public: health and Sincerely, William Kerr Chairman '
References:
- 1. i NRC 2 Confirmatory Action Letter to Su> ton Edison Company dated April l 1,1986 27, 1986 and Supplementary Confirmatory Action Letter dated August 1
- 2. i NRC Memoran6m dated July 275 1988 transmitting Systematic Assess.
- 3. ment of Licensee Performance (SALP) Board Report No. 50-293/87-99
. Memerandum dated June 17, 1987 from W. Russell, Regional Adminis-trator, NRC, to R.
Bird, Boston Edison Company. transmitting i Systematic Assessment of Licensee Perfomance (SALP) Report No. 1 50-273/86-99 4.
~ Memorandum dated May n , 1906 from T. Murley, Regional Administra-tor, NRC, to W. Harrington, Boston Edison Company. - transmitting Systematic Assessment of Licensee Performance (SALP) Report No. j
$0-293/85-99 j 5.
Memorandum dated September 7,1988 from T. Martin;, NPC Region I Office, to R. Bird, Boston Edison Company,
Subject:
Pilgrim j 6. Huclear Power Station Power Ascension Program Memorandum dated September 7,1988 from S. Collins, NRC Region I Office, to R. Bird Boston Edison Company, transmitting NRC Region I Inspection Report No. 50-293/88-21 Integrated Assessment Team , Inspection i 7. Memorande dated May 17, 1988 and July 22, 1988 from R. Gallo, NRC j Region 1 Office, to R. Bird, Boston Edison Company,
Subject:
Inspection No. 50 293/88-11 i 8. Memorandum dated December 31, 1986 from W. Kane, NRC Region I Office to J. Lydon, Boston Edison Company, Subject, Management Meeting 50/293/86-4 -- 9. Boston Edison Company, Pilgrim Nuclear Power Station ACRS Brief-ing books dated August 2, 1988: Volume 1, " Introduction and Restart Plan " Volume 2, " Appendices to Restart Plan," , Volume 3, "Self-Assessment of Readiness for Restart," Volume 4, " Power Ascension Program (PAP)," Volume 5, " Safety Enhancement Program (SEP)"
. <e l The Honorable Lando W. Zech, Jr. -4*
September 14, 1988 10. Testimony dated August 26, 1988 of Representative Lawrence R. Alexander, House Chairman of Massachusetts' Joint Comittee on
- 11. Energy to the Advisory Committee on Reactor Safeguards Statement dated August 26, 1988 of Douglas Hadfield, Director, Civil Defense for the Town of Plymouth, before the ACRS, at Memor-ial Hall, Plymouth, Massachusetts .
12. Press "We theRelease People,dated Inc." August 26, 1988 by Steven Comley representing 13. Statement dated Aucust 26, 1988 by J. Kriesberg, Research Director,. Massachusetts Citizens for Safe Energy to the ACRS Hearing on the i Restart of' the Pilgrim Nuclear Plant 14
'Ouestion for Inclusion in Congressional Record of January 7,1966 Hearing en the Restart of the E11 grim Nuclear Power Plant" (un-dated); Submitted to Ad Hoc Subcommittee on Pilgrim Restart at its Aucust 26, 1988 meeting 15.
Testimony cated September 7,19CS of Representative Lawrence R. Alexander, Hcuse Chairman of Massachusetts' Joitri Comittee on 16. Energy, to the Advisory Comittee on Reactor Safeguards Letter dated September 6,1987 from C. Barry, Secretary of Public Safety, Cnmonwealth of Mar.uchusetts, to W. korr, Chaiman, ACRS 17. regarding readiness [to restart] of Pilgrim Station Memorandum NRC, transmitting:dated August 6,1987 from R. Krim, FEMA, to F. Congel,
" Report of Self-Initiated Review and Interim Finding [ofoff-siteemergencyplanning]forthePilgrimNuclear Power Station," dated August 4,1987 18.
Report dated December 16, 1986 entitled, aReport to the Governor on Emergenc Station"y Preparedness for an Arritient at the Pi1 grim Nuclear Power g and Supplemental Report dated December 1987, submitted by C . V. Barry, Secretary of Public Safety, Commonwealth of Massachusetts
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4 f "*%,, UNITED STATES l' "s NUCLEAR REGULATORY COMMISSION i '! 0FFICE OF GOVERNMENTAL AND PUBLIC AFFAIRS, REGION I
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475 Allendale Road, King of Prussia, Pa.19406 Tel. 215 337 5330 N0. I-88-148 September 16, 1988 Contacts: Karl Abraham ' Steve Horwitz NRC STAFF TO HOLD PUBLIC MEETING ON SEPTEMBER 29, IN PLYMOUTH, MA TO DISCUSS THE PROPOSED RESTART OF THE PILGRIM NUCLEAR PLANT The Nuclear Regulatory staff will hold a meeting from 7:00 to 10:00 p.m. on September 29, 1988, in Memorial Hall in Plymouth, MA to discuss its view of the readiness of the Pilgrim Nuclear Power Station to resume operations. The plant has been shut down since April 12, 1986. Members of NRC management will be prepared to discuss with officials and the public the findings of a recent Integrated Augmented Team Inspection (IATI), and the most recent NRC Systematic Assessment of Licensee Performance (SALP) report. The NRC has committed to make these documents available and discuss their results in conjunction with an NRC Staff decision on the readiness of the Pilgrim facility and management to restart the plant. In addition to these documents, the NRC intends to make available copies of its Restart Assessment Report at the meeting. This document which forms the basis for. the Staff's restart recommendation will also be discussed. The NRC staff at the meeting will be headed by Mr. William T. Russell, Regional Administrator of NRC Region I. Copies of the IATI and SALP reports have been sent to the NRC Local Public Document Room in the Plymouth Public Library on North Street and also have been placed in the local public libraries at Duxbury (147 St. George St.); Carver (Main Street); Kingston (Summer St.); Marshfield (Library Plaza); Plympton (248 Main Street) and Wareham (75 High Street). These documents are being made available as reference documents for those members of the public who wish to look at them before the public meeting. Members of the public may make statements to the NRC staff of up to five
.ainutes each, to allow the maximum number of persons to ask questions or to state their views. Written statements may be mailed to Mr. A. R. Blough, U.S.
Nuclear Regulatory Commission, 475 Allendale Road, King of Prussia, PA 19406. A transcript will be made of the meeting.
.'h THE COMMONWEALTH OF MASSACF SETTS
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- # 7 DEPARTMENT OF THE ATTORNEY GENERAL
$ JOHN W McCoAMACK STATE OFFICE BUILoiNo oNE ASHBUAToN PLACE. BOSTON 02100 169 ,
JAMES M SHANNoN
^"C""***^' May 23, 1988 NI #
Donnie H. Grimsley Director, Division of Rules and Records I Office of Administration and Resources Management EfdEDOM 0F INFO?.MATI United States Nuclear Regulatory Commission ACT REQUEST 7735 Old Georgetown Road kOM d[-dk Bethesda, MD 20814
Dear Mr. Grimsley:
Ox Wr-a6ff This is a request under the Freedom of Information Act (FOIA), as amended 5 U.S.C. 552. I hereby request all records concerning the Restart Plan (including all referenced appendices) submitted on July 30, 1987 by Boston Edison Company to the NRC for the restart of the Pilgrim Nuclear Power Station. This request includes, but is not limited to, all documents concerning the NRC's review of the July 30, 1987 Restart Plan and all of its subsequent revisions. Also included are all records exchanged between Region I and NRC headquarters personnel concerning the Restart Plan, records of telephone conversations between NRC and Boston Edison Company personnel concerning the matters addressed by the Restart Plan, and all records related to review teams assigned to review the Restart Plan and the physical plant (such as the Pilgrim Restart Panel and the Integreated Assessment Team). I reserve the Attorney General's cight to appeal the withholding or deletion of any material. As authorized in the Nuclear Regulatory Commission's regulations at 10. C.F.R. S 9.41, we believe that furnishing these documents without charge to the Massachusetts Department of the Attorney General would be a governmental courtesy appropriate to this request, particularly given the importance of the documents sought and our limited budget. As you know, the FOIA also authorizes you to reduce or waive fees when release of requested information would benefit the public interest. .As Attorney General for the Commonwealth, M ~ _ - - - - - - - --- - -- - _ - a
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we believe the information requested would clearly be in'the , public interest, and therefore the fees for searching and i reproduction should be waived. . The documents requested are for the Attorney General's review of the conduct of the Boston Edison Company during the outage of the Pilgrim Nuclear Power Plant. The Attorney General conducts such a review regarding the reasonableness of the Company's actions as part of his i representation of Massachusetts consumers before the Massachusetts Department of Public Utilities. Provision of the requested documents with waiver of fee will both. facilitate this review process and reduce the cost of providing this governmental service for Massachusetts citizens. Based on the above and on our limited budget, I therefore request that you waive any fees. As provided in 10 C.F.R.-S 9.41, I reserve the Attorney General *a right to appeal any denial of fee waiver or reduction. If you have any questions, please call me at 617-727-2265. Fursuant to the FOIA, I will expect a reply within 10 working days.
.l Sincerely,
).+ P m - f%
Douglas G. Carrey-Beaver Assistant Attorney General Utilities Division Public Protection Bureau (617) 727-2265 DC-B/cb
b T COMMONWEALTH OF MASSAC,HUSETTS G N .1 DEPARTMENT OF THE ATTORNEY GENERAL
$ JOHN W McCoAVACK ST ATE OFFICE BUILotNG ONE ASHBUATCN PL ACE BOSTON
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- JAMES M SHANNoN A f70mNEY GEN (RAL N '"
May 23,'1988 547
- Donnie H. Grimsley EREEDOM OF INFORMATION Director, Division of Rules and Records Office of Administration and Resources Management [JOT REQUEST United States Nuclear Regulatory Commission ((c)/f 7735 Old Georgetown Road Bethesda, MD 20814 'd[~dh $
Dear Mr. Grimsley:
This is a request under the Freedom of Information Act (FOIA), as amended 5 U.S.C. 552. I hereby request all records concerning the NRC's review of Boston Edison Company's proposed Safety Enhancement Program for the Pilgrim Nuclear Power Station as described in Mr. R. G. Bird's July 8, 1987 letter to Mr. Steven A. Varga. This request includes, but is not limited to, all documents exchanged between the.NRC and Boston Edison Company on this subj ec t , meeting memoranda, transcripts, and/or presentation 1 materials, and records of telephone conversations between NRC and Boston Edison Company. I reserve the Attorney General's right to appeal the withholding or deletion of any material. As authorized in the Nuclear Regulatory Commission's regulations at 10. C.F.R. S 9.41, we believe that furnishing these documents without charge to the Massachusetts Department of the Attorney General would be a governmental courtesy appropriate to this request, particularly given the importance of the documents sought and our limited budget. As you know, the FOIA also authorizes you to reduce or waive fees when release of requested information would benefit the public interest. As Attorney General for the Commonwealth, we believe the information requested would clearly be in the public interest, and therefore the fees for searching and reproduction should be waived. The documents requested are for the Attorney General's review of the conduct of the Boston Edison Company during the outage of the Pilgrim Nuclear Power l S9 fur MP~ f-
L.' l<. I . Plant. The Attorney General ~ conducts such a' review regarding the reasonableness of'.the Company's actions as part of his-representation of Massachusetts consumers betore the l Massachusetts Department'of Public Utilities. Provision.of the: requested documents with waiver of f ee will' both f acilitiate i this review process.and reduce the cost of providing this I governmental service'for Massachusetts citizens. Based on.'the above and on our limited budget, I therefore request that you waive any fees. As provided in 10 C.F.R. S 9.41,.I r e s e'I v e the Attorney General'a right-to appeal any denial of fee. waiver or reduction. If you have any questions, please call me at 617-727-2265. Pursuant to the FOIA, I will expect a reply within 10 working days. Sincerely, 1 r }., P 4 S% Douglas G. Carrey-Beaver Assistant Attorney General Utilities Division Public Protection Bureau (617) 727-2265 DC-B/cb _=_______=____________
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[5 II, i ,,;,c f q. .Qi " 5 , f, ; . , ..[' , N ' s , SO57DN EDf50N Pilgnm NucCar Ptmer Station .
- Rocky Hill Road Plymouth, Massachusetts 02360 i ,
l ' Ralph G. Bird Senior vice Presdent- Nuclear February 29,1988 BECo Ltr. 88 033
- U.S. Nuclear Regulatory Commission Attn: Document Control Desk t Washington, D.C. 20555 Docket No. 50-293 License No. DPR-35
Subject:
Resconse to NRC Ouestions and Concerns on the Pilarim Power Ascension Procram Plan and Confirmation of Aareement Durino 2/11/88 Telechone Conference. .
Dear Sir:
Enclosed is Boston Edison Company's response to Mr. S.J. Collins letter of January 28, 1988 regarding the Pilgrim Power Ascension Program. Also enclosed is Boston Edison's confirmation of agreements made during our telephone conversation with Mr. Collins, other Senior NRC Hanagers and staff personnel on February 11, 1988 at approximately 1330 hours. As discussed in the above conversation, Boston Edison considers that a discussion between NRC and Boston Edison personnel interested in this matter, would be beneficial in assuring full understanding of the basis for the Power Ascension Program.
.c. IM l PJH/b1
Enclosure:
Enclosure 1 and 2 l
Attachment:
Attachment A and B f acc: Mr. William Russell Regional Administrator, Region I J U.S. Nuclear Regulatory Commission 475 Allendale Rd. l King of Prussia, PA 19406 Mr. Samuel J. Collins i Deputy Director Division of Reactor Projects U.S. Nuclear Regulatory Commission l 475 Allendale Rd. King of Prussia, PA 19406 SR Resident Inspector - Pilgrim Station
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Boston Edison Company Docket No. 50-293 l Pilgrim Nuclear Power Station License No. OPR-35 f Enclosure I - Resoonse to NRC Ouestions and Concerns on Pilarim Power Ascension Procram Plan
Background:
Questions #2 and #3 appear to be related to the technical issues raised in Confirmatory Action Letter (CAL) 86-10. The pre-startup corrective actions to address the CAL 86-10 issues have been completed. These corrective actions were reviewed by NRC Region I Inspectors on September 21-25, October 5-9 and October 12-23, 1987 at Pilgrim Station. The results of these reviews are documented in NRC Inspection Report 50-293/87-46 pages 11 & 12. dated December 7, 1987 and in Boston Edison's exit meeting minutes for Inspection 87-46. , A list of the Boston Edison submittals to the NRC and the NRC Inspection l Reports that address CAL 86-10 is provided (Attachment A). NRC Question #1. 1 How will management assessment of performance at each plateau in the power ascension program be documented? i g Response to Question #1 Hanagement assessments of performance will be documented on a standard form entitled, " Assessment Summary" to be completed by the Oversight and Assessment Team and signed by the team members. This form is being issued as part of a training module for training of the Oversight and Assessment Team. , NRC Question #2. How will it be demonstrated that major plant transients (such as scrams from high power and flow) will not induce HSIV closures, since the cause for previous MSIV closures has not been conclusively determined? Response to Question #2 The spurious closures of the MSIVs which have been the subjectpf intensive evaluation did not occur during high power or high flow conditions. In both events MSIV closures occurred with the mode switch in the startup/ hot standby
'hosition. Plant testing with the mode switch in the run position would not
. address .the HSIV closure issue observed in April 1986.
The potential contributors to the inadvertent MSIV closures were identified as: (1) Hode switch (2) Hater. level instrumentation (3) Loose grounding connections of PCIS relays. - Page 1 of 4
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y $ : H .. : q , ' . . . .: , : . .o u ._ ; a . l Boston Edison Company- . Docket.No. 50-293-Pilgria Nuclear Po er Station License No. DPR ; i - Bos. ton' Edison believes that the cause of inadvertent MSIV closures has been l corrected by:
- 1. - Replacement of the mode switch
" 2. Replacement of water level instrumentation
- 3. Repairing the neutral connections on the PCIS relays.
I To demonstrate that future reactor start up and shutdown transients will not induce MSIV closures, a test (TP 86-81) will be conducted as part of the Power Ascension Program. Conduct of this test will attempt h duplicate, under i controlled conditions, the plant power and flow conditic,es of 4/4/86 and j 4/12/86 when the MSIV's unexpectedly closed. The test will utilize General l Electric Transient Analysis Recording equipment to monitor the trip signals l which may have led to the original MSIV closures in April 1986. If the MSIVs do not close under the duplicate conditions, then we will have demonstrated 7' j that the actions taken t0 resolve the problem were successful and corrected the problem. ,gM ;
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dC Question #3 and #4: How will it be demonstrated that plant response to MSIV closure transients is still in accordance with design?.'(Your response should address any procedures and personnel responses necessary to safely accomodate the transient.) How will it be demonstrated that the dynamic response of the plant to a Loss ,h g r\ of Offsite response Power should address event, at power, whether procedures is in accordance and personnel training are # M with design h appropriate to safely accommodate this transient.) Response to Questions 3 and 4 The questions relate to plant response after MSIV closures and loss of offsite power. During the events of April 4 and April 12 the plant transient response to the inadvertent NSIV closures was in accordance with design. Loss of s offsite power was not associated with the April 4 or April 12 events. The loss of power event that occurred on November 12, 1987 was with the reactor shutdown and therefore did not exhibit any unexpected plant dynamic response. Review of the original startup test program showed that: /*
- 41) The MSIV response was in accordance with the design.
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(2) The plant response to loss of offsite power was in accordance with the
. design. A{,t .m i I h +1v '
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A review of past plant; history has identified transient events which demonstrated that the MSIV's responded in accordance with the design. The same review also idenTifiedloss of offsite power events where the plant responded according to design. He therefore believe that the original startup testing, the plant transient history, and loss of offsite power events have adequately demonstrated that the plant responds according to design. Page 2 of 4
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Boston' Edison. Company . Docket No. 50-293. Pilgrim Nuclear Pooer Station License No. DPR-35 Additionally, when unplanned events occur during plant operation, the actual 1 -plant response is evaluated and examined for indications of -abr>ormal behavior per procedure 1.3.37 " Post Trip Review". This procedure was reviewed by the NRC as part of NRC's audit of BEco's response to Generic Letter 83-28. I You further inquired about procedures and personnel training necessary to I safely accommodate plant transient responses from these' events. These transients are . treated in operating procedures and are . covered by existing { routine training and by special simulator refresher and E0P training.
^
NRC Question #5 _ l How will the test program demonstrate and/or evaluate the extent to which Rosemount 1153 transmitters are sensitive to pressure oscillations during ; major transients, and whether " ringing" of the output will penstary l actuations of safety -n systems? y .. i Response to Question #5 t No special actions for Rosemount Transmitters are planned b. sed on the-following: , (1) Rosemount, based on it's' industry wide investigation, has advised Boston
/ Edison that occurence
- a. is not wide spread y h has never happened simultaneously to two transmitters in a plant
- c. is random (2) a single transmitter failure would not a:tuate a safety system when l redundant channels are in service g ,
(3) adequate controls exist to detect transmitter performance abnormalities. ! l The Rosemount transmitters installed at Pilgrim this outage that could be { susceptible to sensitivity problems are part of the Analog Trip System. In i this system, the circuit logic requires a coincident transmitter signal from ; redundant trip channels to actuate a safety system. Therefore a single , transmitter failure would not challenge a safety system when refundant .m channels are in service, nor would it preclude actuation if ne(ded.en M6 j Ehe surveillance program for these transmitters is a once per day" channel Lk check which compares the redundant channels for each parameter. TheP V surveillance provides a direct indication of transmitter performance and; is N adequate to detect a small on-scale _ shift. i N uestion #6 What will be the administrative controls for conduct of testing and review of test results, including control of the performance of testing, methods for j- recording, evaluating, reviewing and approving test data and results, and the l methods .for identifying and correcting deficiencies noted in systems and procedures? Page 3 of 4 h
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- i. . Boston Edison Company Docket No. 50-293 Pilgrim Nuclear. Power Station License No. DPR-35 Res'ponse to Question #6 h The details of the administrative controls for conduct of testing and review l of test results during the Power Ascension Program are docriented in an i internal BECo memorandum of January 29, 1988 from the Plani. Manager to managers in the Nuclear Organization.
p While'this memorandum is an internal document which would not ordinarily be i forwarded to the NRC, we are providing this (see Attachment B) in accordance
- with your specific request during the 2/11/88 telephone conference.
Page 4 of 4
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- Boston Edison Company Docket No. 50-293 Pilgrim Nuclear Power Station License No. OPR-35
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Enc 1osure 2 - Confirmation of Aareements Durina 2/11/88 Telechone Conference On 2/11/E8 a telephone conference between managers from NRC Region I. NRR. NRC Resident Inspectors and Boston Edison was conducted. The purpose of the conference was to discuss the technical basis for Boston Edison's response to the NRC's letter of January 28, 1988 which transmitted questions and concerns on the Pilgrim Power Ascension Program. AGREEMENTS
- 1. During the Power Ascension Program Boston Edison will release Pilgrim for New England Power Exchange dispatch as soon as the reactor reaches full power. q
- 2. When Pilgrim reaches 100% reactor power, Boston Edison will conduct its final assessment of the Power Ascension Program at Assessment Point 4.
The results of this final assessment will be recorded, documented and submitted to the NRC to document closure of~ the Power Ascension Program. 1 L J' Page 1 of 1 l
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6 . _ Boston Edison' Company. ' Docket No. 50 ~293 Pilgrim Nuclear Power Station License No. DPR-35 0 l d .. . L Attachment A \ Boston Edison Submittals and NRC Insoection Reoorts Addressina Technical Issues Raised in Confirmatory Action Letter (CAL) 86-10 o Y. j Dgg DESCRIPTIO$ LETTER NO./ REPORT NO. 5-15-86 BECo's first response to NRC LTR. #86-062 Confirmatory Action Letter ,
#86-10. f 5-16-86 Augmented Inspection Team RPT #86-17 (AIT)' Report BECo's second response to LTR #86-079 i 6-16-86 to NRC Confirmatory Action I Letter #86-10.
. 8-29-86 BECo's supplemental response to LTR #86-128 NRC Confirmatory Action Letter
#86-10.
9-17-86 8Eco's submittal of additional LTR #86-141 information as committed in Letter #86-128. 3-1-87 BECo's response to NRC letter LTR #87-039 concerning Management Meeting - 50-293/86-41. 7-30-87 BECo's transmittal of the LTR #87-130 Restart Plan (Rev. 0). 10-15-87 BECo's transmittal of the LTR. #87-163 Power Ascension Program. 60-26-87 BECo's transmittal of LTR.#d-172 , Volume 2. Rev. 1 of the
- Restart Plan.
l 12-1-87 NRC Inspection Report RPT. #87-42 I (closure of 86-17-04). 12-7-87 NRC Inspection Report RPT. #87-46 (closure of 86-17-02) (update of 86-17-01). l Page 1 of 1 I
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' ATTACHMENT B OFFICE MEMORANDUM BOSTON ~4DISON COMPANY TO: Distribution FRON: R. J Barrett Record Type A4.08 DATE: January 29, 1988 Dept Doc. PM 88-062
Subject:
DESCRIPTION OF POWER ASCENSION TEST PROGRAM DISTRIBUTION: R. G. Bird K. L. Highfill R. N. Swanson R. A. Ledgett E. J. Howard D. L. Gillispic R. J. Barrett All Section Managers I. PURPOSE: The purpose of this document is to define and establishTest the following in the temporary Post-Refueling / Outage Organization: A. Define the relationship of the Test Organization to Operations organization-, B. Definition of the organization and description of the- ' duties and responsibilities of personnel assigned. C. Describe the schaduling of testing. D. . Describe the conduct of testing. E. Describe the independent review of completed test data. II. BACKGROUND: Experience from the testing program after RFO-6 and prudence dictate that the conduct of post-outage testing be carefully planned and , controlled in The order to ensure completeness, following sectdons of _this safety and effectiveness. program description document define the scheduling,' conduct, and review of testing to be accomplished during startup from RFO-7. CUMENT [ RELEASED Page 1 of 12 p( l 1 FEB 5 988 i FOR USE NTRot, ST ATio
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Attcchment B con't.
. Date: January 29, 1988 Perse 7 III. REIATIONSNTP OF triE TEST ORCANTEATION TO THE NUcf.FAR WATCH ENCINFFR fNME) AND HIS SEenON CREW The NWE is the senior licensed operator on watch and has The Test complete responsibility for. operation of the plant. The NWE's Organization is subordinate to the NWE.
concurrence and permission must be obtained for the conduct of all tests. Direction must not be given to watch section operators for operation of plant equipment except as it pertains to conduct of previously authorized tests. The Shift Test coordinator (STC) must work closely with the NWE, may advise him, but must exercise exceptional care to not interfere with the NWE and his operators in discharge of their license and operational duties. IV. STARTUP TEST ORGANIZATION description's of the duties and The following are of defined positions in the Startup Test responsibilitiesThe Startup Test Manager may assign additional organization. personnel to perform specific tasks in addition to the positions described herein. An organizational diagram of the showing the relationship of the Startup organization , temporary startup test organization and the permanent station organization is shown in Figure 1. l 5 : j.\ , e Page 2 of 12
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ATrJONENT B can't. , l 4 j STARTDP TEST ORCANTF.ATION e POSITION RESPONSIBILITIES i A. STAR'mP TEST MANAGER (J. A. SEERY) Major duties and responsib'ilities of this position are:
- 1. Development of the Power Ascension Schedule.
- 2. Development and staffing of the Startup Test Program.
- 3. Development of a master power ascension test procedure.
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- 4. Ensuring all test procedures required for the startup and testing are approved.
- 5. Day to day management of the Power Ascension Program keeping the Plant Manager informed of progress and problems. .
- 6. Approval of completed startup test reports.
- 7. Recommendations to the Plant Manager when the plant is ready to ascend to the next test plateau.
- 8. Scheduling of testing and promulgation of the test schedule.
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- 9. Interface with the operations Section Manager to ensure testing and normal plant operations are efficiently scheduled.
- 10. Direction of the Shift Test Coordinators.
- 11. Ensuring complete test data review.
- 12. Approve a waiver of a discretionary test in TP 87-114.
B. ASSISTANT STARTUP TEST MANArmn (K. R. NICBOLAS) r This position reports to the Startup Tesit Manager and is a responsible to him for the following duties:
- 1. Development of test reports, as necessary.
- 2. Technical advice in development of the schedule and necessary changes.
development' of a startup test
- 3. Assistance in organization and position responsibilities.
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i *- Q . 1 . ATIA09 TENT B can't. i o 4. Assistance in development of a mEster power ascension procedure. P 5.- Assistance in the day to day management of the Power Ascension Program as necessary.
- ' 6. Ensuring that test da.ta is reviewed for acceptance'by r the proper technical review group.
- 7. Technical and completeness review of completed test reports and submission to the Startup Test Manager for I
[ approval. C. REPORT WRITERS This position reports directly to the Assistant Startup Manager. Their duties are as follows:
- 1. On a daily basis, collecting completed tests from the Shift Test Coordinator
- 2. Reviewing test for completer $ess Preparing -test reports for the Assistant Startup 3.
Manager's review
- 4. Assisting the Assistant Startup Manager as required SNTFT TEST COORDINATORS (P. SMITH, J. BELLEFEUILLE, P.
D. NANDERINO, J. SABINA) Shift Test coordinators will maintain shift coverage during the testing phase. They will assist the on-shift Nuclear Watch Engineer by coordinating all testing for that shift and obtaining his permission to conduct testing. Their duties will consist of, but not be limited to the following: Ensuring tests are conducted in a formal and 1. professional manner. J'
- 2. Ensuring tests are conducted using approved procedures.
- 3. Assisting the NWE in the conduct of pre-test briefings with the shift crew as necessary.
- 4. Keeping abreast of current plant status.
- 5. Ensuring test procedures are properly controlled.
- 6. Maintaining a shift log of the testing status progress on a shift basis.
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- 7. In conjunction with the Nuclear Watch ' Engineer, recommending changes.to the suggested critical path schedule when necess.ary.
- 8. In conjunction with the Watch Engineer, . recommending waiver of d.scretionary tests in TP87-ll4.
- 9. Keeping the Startup Test Manager informed in problems a timely j manner ~ of the status of testing and of encountered.
E. READIOR ENGINFERS. (J. ABOLTIN) isrwurr n ENGihmins. TEST ENGIh dks. TURBINE TEST ENG1hrGS. AND ii.5T AND TuxNOVEB ENGINEEBS generic responsibilities of these positions are The They are similar. These are the Test Directors. responsible for directing and conductingthe theirShift assigned Test tests. On shift they report to Coordinators. j Their duties will consist of, but not be limited to, the
- following:
- 1. Directing performance of their approved test after NWE concurrence and permission has been obtained.
- 2. Keeping the Shift Test Coordinator and Nuclear Watch Engineer informed of test status 'and affects' on plant conditions.
- 3. Assisting in pre-test briefings with the shift crew as ;
necessary and directed by the NWE. , , V. NUCMAR ORGANIZATION SUPPORT FOR OPERATIONS 'I
- The restart from RTO-7 will be. achieved with the normal plant Quality l
support functions, e.g., Nuclear Engineering, Assurance / Quality Control, Security, Fire Protection, Planning / Scheduling and Cost Control, Technical Support, and other nuclear. organization groups. Fu .. , A. Due to the complexity of this outage, there will be a Shift ,
- Maintenance Representative, as shown and described in BEco ,
Ltr. 87-163. B. Quality Assurance coverage, although functioning through , normal channels will be expanded during the preparation and ' testing phases of power ascension. Around the clock coverage will be provided by approximately 9 dedicated l inspectors. A QA checklist developed specifically for power ascension will be used. Dedicated coverage of 1 chemistry will also be provided. , Page 5 of 12
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2 ATT>ONENT B can't VI. CONDUCT OF TESTING A. This section descr, bas the mechanism for the conduct.cf-testing during the Power Ascension Program. B. REFERENCES (Latest revision- unless .otherwise indicated) , ANSI N.18.7-1976 Administrative 3ntrcls and
- 1.
Quality Assurance for the Operational Phase of Nuclear Power Plants
- 2. ANSI N.45.2.6-1978 Qualifications of Inspection, Examination, and Testing Personnel for Nuclear Power Plants
- 3. ANSI N. 4 5. 2.11-1974 Quality Assurance Requirements for the Design of Nuclear Power Plants I
- 4. BEco Quality Assurance Manual, Volume II, Section-II .
PNPS Technical Specifications, Section 6.8 )
- 5. 1
- 6. Procedure 1.3.4 - Procedures
- 7. Procedure 1.4.5 0 PNPS Tagging Procedure
- 8. Procedure 1.5.3 - Maintenance Request C. DEFINITIONS:
- 1. PDC - Plant Design Change
' 2. ORC - Operations Review Committee
- 3. DCC - Document Control Center
- 4. SRO - Senior Reactor Operator
- 5. MR - Maintenance Request r
- 6. RWP - Radiation Work Permit
- e
- 7. TP - Temporary Procedure
- 8. TD/TE - Test Director / Test Engineer. The TD/TE is the individual responsible for coordinating the performance of a test after permission of the NWE has been obtained. .
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- 9. STC - Shift Test' Coordinator. The STC is ' the individual who is responsible for management.of all testing conducted on his shift. An STC can have several Test Engineers reporting to him. The STC
' reports to the Startup Test Manager.
- 10. PNPS - Pilgrim Nuclear Power Station D. PREREQUISITES: -
The 'Startup Test Manager will assign Shift ' Test Coordinators (who meet appropriate educational and experience requirements of ANSI N.45.6-1978). He will certify each STC for the appropriate level using the These Test Director Qualification Sheet. certifications will be maintained on file within the
- Startup Test organization. i
'E. TEST PREPARATION:
- 1. Upon authorization by the NWE of tests scheduled for that shift, the Shift Test Coordinator can make preparations to start.the test.
- 2. Prior to the actual start of the test, the STC will ensure the following:
- a. That a latest revision of the . procedure has been obtained from the DCC.
- b. That the procedure to be ased is the same revision as the one in the control Room and is '
marked " Working Copy".
- c. That an approved Maintenance Request is ready ;
for use (if needed). +
- d. That an RWP has been approved (if needed) and is ready to support the test.
- a. That plant operating conditions and system status required for the test have % en set.
- f. That manpower and equipment needed are scheduled and available.
prerequisites stated in the
- g. That ail prerequisite section of the test procedure are satisfied.
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F. CONDDCT pF TESTING:
- 1. -During the conduct of the test, the Test Engineer
' 111 ensure that the personnel conducting the test
- are cognizant of all requirements of the test-such ( - as:
- a. All prerequisites listed under Section VI of-4 the procedures " Prerequisites".
- b. All personnel initialing the procedure will 1
,l~ sign their full name on a Signature l Identification Sheet.
- c. Any calibrated equipment which is part of~the Meter and Test Equipment.(M & TE) System baing used is listed . on an equipment log. If the equipment being used is not part of thewill M & TE be o System, a calibration Data Sheet attached as an addendum to the log.
h d. Test procedures, unless specifically stated, do not permit any exceptions. Any changes to the procedure will be in accordance with PNPS 1.3.4.
- 2. If a problem is encountered during.the conduct of a test the following steps will be taken:
If a hardware ; ;;1em is
- a. Hardware eroblems -
encountered, the TE will stop the test. A trouble shooting MR will be initiated by the TE. All necessary manpower will be obtained for trouble shdoting. The scope of trouble shooting will be to the extent of identifying the components ' affected. Problems identified will be repaired using normal station repair procedures. After the problems are repaired, the test may again be schedule by the Startup Test Group for completion. 39 .; . ..g f,, l b, software erabl=== - If the probisNm encountered is procedural, the Test Engineer will stop the a test, obtain the necessary change to . the , Test Procedure in accordance with Station Procedure 1.3.4. The Test Engineer may resume the test changes are made, and with the after the approval of the Shift Test Coordinator and the NWE.
- c. The TE is responsible for timely notification of the Nuclear Watch Engineer and the STC of any problems encountered and of completion of the test.
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' '- ATIACINENT B can't.
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.In sections a. and b. above, if the test procedure permits the TE to do other ' sections which are independent of the section being reworked (hardware) or revised - (software), the TE can continue with the ' -other independent portions of the test with the concurrence of the STC and NWE.
G. TEST VERIFICATION:
- 1. After the test is complete,-the TE will review the test to ensure the following:
- a. That the test was conducted as written without any discrepancies or exceptions.
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- b. That the acceptance criteria of the applicable steps'have been satisfactorily met. !
- c. That the QC Group has reviewed all their hold and witness points and has signed off appropriate steps in the procedure.
- 2. When the above steps are completed, the TE can then sign Section XI of the procedure for Test Completion. The TE will then submit the test to the STC for his review.
- 3. The Shift Test Coordinator will review the entire !
test procedure to ensure that:
- a. All aspects of the test have been completed in accordance with station procedures.
- b. All revisions of the procedure used during the test are available and attached - to the final revision.-
- c. After 5ections G.3.a and -G.3.5 above,, are ;
completed the STC will sign Section XI, of the !
- procedure if required, of the procedure for the STC. The procedure will now be presented to the Nuclear Watch Engineer for review and approval for completeness.
- d. After the Nuclear Watch' Engineer has signed the procedure, the procedure will be sent to the Assistant Startup Manage 2'.
Page 9 of 12 L_____-________--__.-__-_____
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The' Assistant Startup Manager will coordinate the
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4 ~. independent review of the test packages. Submit to the Startup Test Manager a corplated test . package including . the independent review results for his approval.
- 5. The Startup Test Manager will review the; test procedure along with the results of the independent review described in Section VII below. After approving the test, he will attach the independent
' review to the procedure and transmit to Document Control for capture.
VII. INDEPENDENT REVIEW OF TEST RESULTS An independent review. of the test results ' obtained during the Power Ascension Test Program will. be performed by the Systems Group. The support of personnel from the . other groups such as NED and Operations will be required as appropriate for their specific expertise.
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but not be limited 'to the
- The . review will include following:
-\J . Verification that the test was performed.as written.
' - . All data was taken and recorded as required.
- All acceptance criteria were satisfied.
After completing the independent technical review of a i completed test procedure the Systems Group Leader Will notify the Startup Test Manager in writing of the findings. l The Startup Test Mant.ger' will review and collect all ; independent review sheets. After all applicabic. procedures I the I have .heen satisfactorily performed and reviewed, Startup . Test Manager will submit a commitment close-out memo to Compliance. VIII. SCHEDULING OF TESTING . A. The Startup Test Manager is responsible for. scheduling of testing and promulgation of the test / schedule.: the To discharge these responsibilities he will. do t\',.
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i , I following:
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- l. Schedule testing on a shift basis. When tests are
- required to be done in a specific sequence, in that the constraint will be clearly indicated
\g ,V schedule. In forming the test schedule the Startup
.i Test organization and the operations Section will coordinate to ensure safe and efficient plant
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' operation.
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- 2. The Power Ascension' Test Schedule will be approved by the Operations Section Manager. This will be f'
4 integrated inte .a Plan of the Day The (POD) PODwhich will ' beis p~ approved by the Plant Manager. with an issued - for a specific period of time, effective date/ time and an expectation date/ time. The testing schedule may be extended or changed 3. wi'th formal concurrence of the StartupDuring Test Manager periods
" and the Operation Section Manager.
when either or both of these managers - are not ' on site, the STC may obtain concurrence by telephone from the absent manager and approve test ~ schedule changes for these managers "per telcon" noting the a date 'and time telephone concurrence wasappropriate obtained, ! manager will sign the The absent {
- approval during the next day the manager is on
[' site. IX. COMMUNICATIONS _DURING TESTING A. Communications during testing willA be conducted in a good model for use ) i
. precise and professional manner.in.the conduct of testing commu B. Communications equipment to be used in testing will be in advance of-tested.for operability whenever possible, Use'of the plant paging system the start of the test.
for routine operational communications during testing and used only when no other should be minimized appropriate communications paths are available. X. TESTING SUPPORT be requested from various -NUORG Testing support' may the Assistant organizations by ' the Startup Test Manager, Support may Startup Manager and Shift Test . Coordinators. be requested in addition to that specified in this document and Operations. Any from Engineering, Maintenance difficulty encountered in obtaining added Manager support shall the through be resolved- by the Startup Test appropriate chain of command.
- XI. SEcuraclWG OF Trus ruwrm ASCENSION TEST PROGRAM Conduct of the test program descriptions of major test and schedules and notices for the test program are contained in Test after RFO-7.
TP 87-114, Restart Attachment 1, Attachment 1 shall be the governing document and sequencing of testing. Changes to for thescheduling intent of i Attachment 1 shallusing be made only after the approval of the proper procedures for such changes. ORC is received, Page 11 of 12
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Pilgrim Nuclear Power Station , , , , . Rocky Hill Road Plymouth, Massachusetts o2360 [
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g f.][ MN' I Ralph G. Bird . Senior Vice President - Nuclear p ri) y4, )ggg BECo Ltr. #88 079 i U.S. Nuclear Regulatory Commission F Attn: Document Control Desk Washington, D.C. 20555 . Docket No. 50-293 License No. DPR-35
Dear Sir:
This letter is submitted in response-to the NRC Staff questions transmitted by NRC letter dated March 18, 1988 regarding the Pi' grim Nuclear Power Station Restart Plan . Attachment 1 contains a restatement of the NRC question followed immediately ! by the Boston Edison response. Please contact me directly if there are any
. questions regarding this letter.
R. G. ird PJH/bl Attachment cc: Mr. William Russell Regional Administrator,- Region i U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Sr. Resident Inspector - Pilgrim Station Standard BECo distribution
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a.. . . n . ATTACHMENT 1 l*. BECO Resconse to NRC Review of Pilarim Restart Plan , 0 NRC Ouestion 1 k Volume 1, Chapter 2 of the Restart Plan needs to be updated to reflect the'recent organizational and personnel changes. BECO Resconse The recent organizational and personnel changes within the nuclear organization of Boston Edison Company will be described in a chapter on organizational and personnel changes.in the Final Report of Management Self-Assessment of Readiness to Restart. This report, which documents the results of the self-assessment, is a companion document to the c. Restart Plan and the Power Ascension Program. According to the current schedule, the self-assessment report will be issued in mid-Hay,1988. , N_RC Ouestion 2 , Under the new organization, how will the functions attributed to the Planning and Restart Group as defined in Volume 1, Chapters 2 and 3 be accomplished? For example, the Work Planning and Estimating Branch provided each section}}