ML20133B804
| ML20133B804 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Davis Besse |
| Issue date: | 07/31/1981 |
| From: | PROFESSIONAL LOSS CONTROL, INC. |
| To: | |
| Shared Package | |
| ML20132B273 | List:
|
| References | |
| NUDOCS 8510070117 | |
| Download: ML20133B804 (31) | |
Text
..
8
_PL C 'w"!~' "" c'"" '"<-
j I
1 ANNUAL FIRE PROTECTION AUDIT,
of the DAVIS-BESSE NUCLEAR POWER STATION UNIT NO. I e
for the l
TOLEDO EDISON COMPANY l
j i
i t
l i
I l
i i
t l
i.
Prepared by:
Professional loss Control, Inc.
Submi tted: July 31, 1981 i
CORRESP00@ENCE PDR i
l P. O. Box 446 'e Oak Indge, Tesnc<sce 3"S30 * (Gl$) 482 354!
1 1
TABLE OF CONTENTS Subject Pages 1.0 I n t ro d u c t i on......................... ' 1 2.0 Method of Evalua ti on....................
1 2.1 Audi t of Pl ant Facilities...............
1 2.2 Review of Fire Protection Related Procedures.....
2 2.3 Verification of Procedures..............
2 2.4 F i re B r i g ade......................
2 2.5 Follow-up to Previous Recommendations from 1980 Annual / Triennial Audit......
3 3.0 Conclusions........................
3 3.1 Administrative Control of Fire Hazards........
3 3.1.1 Control of Ordinary Combustibles and Flammable Liquids....................
3 3.1.2 Control of Ignition Sources..........
3 3.1.3 Integrity of Fi re Ba rriers...........
4 3.1.4 Control of Plant Modifications.........
4 3.2 Fire Protection Equipment Surveillance Procedures...
S 3.3 Fi re Protection Systems and Equipnent.........
6 3.3.1 Fi re Supp res s ion Sa te r Sys tems..........
7 3.3.2 Fire Detection and Alarm Systems........
7 3.3.3 Fi re Supp res sion Sys tems............
7 3.4 F i re B ri g a de.....................
8 3.4.1 Organization..................
8 3.4.2 Training....................
8
~
3.4.3 B ri g ade Equ i pmen t...............
10 3.4.4 P re-F i re Eme rg e n cy Pl ans............
10 4.0 Unresolved Reconnendations from the 1980 Annual / Triennial Audit...........................
10 4.1 Adninistrative Control of Fire Hazards........
10 4.2 Fire Protection Equipment Surveillance Procedures...
11 4.3 Fire Protection Systems.and Equipment.........
11 4.4 F i re B r i g a de.....................
11 5.0 New Re commen da ti ons....................
12 5.1 Fire Brigade.....................
12 6.0 S umm a ry............................
12 Pre and Post Audit Conference Attendance........ Appendix A Technical Specification Surveillance........... Appendix B Procedures Revi ewed.................... Appendix C Test Procedure Documentation Reviewed........... Appendix 0 Fire Door Problems.................... Appendix E j
.,S.
ANNUAL FIRE PROTECTION AUDIT of the y
DAVIS-BESSE NUCLEAR POWER STATION UNIT NO. 1
~
for the TOLEDO EDISON COMPANY
1.0 INTRODUCTION
An annual fire protection inspection and audit was conducted of the Davis-Besse Nuclear Power Station Unit No.1 in accordance with Section 6.5.2.81 of the Technical Specifications. The fi, eld portion of this audit was conducted on July 20 - 22, 1981, by Mr.
M.- E. Howrer, P.
E.,
Vice President, and Mr. C. A. Ksobiech, Fire Protection Engineer, of Professional Loss Control, Inc. (PLC). PLC is a consulting firm specializing in providing fire protection engineering s'ervices to the utilities industry.
Mr. Mowrer acted as team leader, and the audit was conducted under his direct supervision. Mr. Howrer's qualifications exceed those specified by the NRC in Reg. Guide 1.120 as required of the person conducting such an audit.-
This report details the results of the annual fire protection audit as well as the disposition of recomendations included in the triennial audit submitted September 9, 1980. It should be noted tha.t the audit reflects a review of present conditions as compared to recognized fire protection engineering practices, current NRC guidelines, and specific commitments made tc the NRC.
2.0 METHOD OF EVALUATION I.1 Audit of Plant Facilitin A review of the plant safety-related areas was conducted, The emphasis of this inspection was the assessment of the overall effectiveness of the plant's fire prbtection prggram to control 1
fire hazards and maintain fire, protection equipment in these areas as outlined in the SER and in accordance with the appropriate require-ments of Appendix R 10 CFR 50.
2.2 Review of Fire Protection Related Procedures Surveill,ance Test, Administrative, and Emergency Procedures related to the plant fire protection programs were reviewed with respect to the administrative requirements for an effective fire protection pro-gram and implementation of Technical Specification surveillance requirements.
I 2.3 Verification of Procedures A review was perfonned to verify that fire-related procedures are being properly implemented. This review consisted of an inspection
~
of documentation.and spot checking of physical facilities to verify 1
documented surveillance and follow-up corrective actions.
2.4 Fire Brigade An evaluation of the adequacy of the Plant Fire Brigade was performed.
Organization of the Fire Brigade, training for Brigade members, fre-quency and content of fire drills, and type and availability of manual fire fighting equipment were surveyed.
As guidelines for this review, NFPA #27. " Private Fire Brigades," and NRC requirements of Appendix R to 10 CFR 50, APCS BTP 9.5.1, Appendix A. Reg. Guide 1.120, and NRC Supplementary Guidance (dated 6/14/77) on Nuclear Plant Fire Protection Functional Responsibilities, Administrative Controis and Quality Assurance, Attachments 1, 2, an'd 5 were applied (referred to as the D.B. Fassallo letter dated August 29, 1977, at the plant).
In conjunction with the review of the Fire Brigade, a review was con-ducted of the fire emergency planning for plant f acilities. The evaluation of the Fire Brigade and fire emergency planning provided a basis for assessment of the plant's preparedness to handle fire emergencies.
i 2
l-m.
n m-
2.5 Follow-uo To Previous Reconnendations From The 1980 Annual / Triennial Audit
~
Follow-up on the status and applicability of recommendations made during the PLC Triennial Fire Protection Audit submitted on September 9,1980, is included in this, audit. Six recommendations from the previous report were found in various stages of completion and five others have been revised.' These eleven recomendations have been repeated in Section 4 of this audit report.
3.0 CONCLUSION
S 3.1 Administrative Control of Fire Hazards Administrative procedures which control ordinary combustibles, flamable liquids, and ignition sources wert reviewed with respect-to the' Plant Fire Protection Program (see Appendix C for a list of Procedures reviewed).
These procedures were found to be generally adequate.
(Recomendations for improvements appear in Section 4.1.)
3.1.1 Control of Ordinary Combustibles and Flamable Licuids Control or ordinary combustibles and flamable and combustible
~
liquids was found to be generally good throughout the plant and responsibilities are clearly defined-(previous recommendation
. 4.1.2 comple ted). Only fire retardant paint or " treated" lumber is allowed in the plant and no violation of this basic company policy was identified.
(It should be noted that shipping crates are not required to be included in this control, but are removed as soon as is practical.)
The wood ramps reported by the previous audit still remain in place (previous recommendation 4.1.7), however, a work order (WR flo. 81-1013-2) was issued on February 17, 1981, to have them replaced.
3.1.2 Control of !anition Sources Smoking appears to be well controlled throughout the plant.
Control of welding and cutting. operations is satisfactory as outlined in AD 1844.00 with the following ' exceptions.
3
Neither the fire watch nor the supervisor issuing the pemit have necessarily had any fire training for these specific tasks as required in Attachment No. 4 of the NRC Supplementary Guidance on Nuclear Plant Fmetional Responsibilities., Adminis-trative Controls and Quality Assurance. Because of the length of service of the personnel (except for those in construction),
inost have had some fire extinguisher training but there is no requirement for it.
(Revised previous recomendation 4.1.1.)
3.1.3 Inteority of Fire Barriers ST 5016.11 and ST 5016.13 provide the basis for maintenance and surveillance of fire barriers. These procedures do not, however, include fire doors and dampers. Thorough fire wall delineation drawings are available at the plant to identify required fi re walls. A contract (#061-F618878-C1) has been issued to detail all electrical and nechanical penetrations in
~
these walls to facilitate future surveillance.
All doors are lubricated and adjusted as a part of a semi-annual maintenance check, but this inspection is not adequately docu-mented.
Required fire doors are now clearly identified as such.
Several minor fire door deficiencies were noted, however, during tha plant inspection (see Appendix E for details).
Current Tech Spec requirements should be interpreted as requiring the surveillance of fire doors, dampers, mechanical and electrical penetrations even though surveillance is not now provided for all of these penetrations.
3.1.4 Control of Plant Modifications All Facility Change Requests (FCR's) are now formally reviewed by qualified personnel fer their impact on fire hazards of the area, fire suppression already provided, and fire detection already provided (previous recommendation 4.1.3 completed).
4
3.2 Fire Protection Ecuiement Surveillance Procedures The station Fire Protection Surveillance, Test, Administrative, and Emergency Procedures were reviewed to verify their compliance to commitments made in the Technical Specification surveillance requirements.
Appendix' C of this audit report lists the procedures which were reviewed, along with the' approval dates of the "ST" test procedures. According to that review, it was determined that many of the previous recommen-dations (submitted September 12,1980) regarding procedural changes have befr. completed.
However, the following items remain incomplete:
1.
The Technical Specification item (paragraph 4.7.0.1.1g) which requires a flow test to be conducted every three years (Rec. 4.2.2).
This proposed flow test would provide some indication of the interior. condition of the.,undenjround piping.
2.
The Technical Specifications item (paragraph 4.7.9.1.1d) which requires a system flush test on a semi-annual basis.
3.
A triennial test for valve operability of fire hose stations as required in the Technical Specifications (paragraph 4.7.9.3.c2).
Procedure (ST5016.16) is being established to cover these items, but it had not been completed at the time of this audit.
A new procedure (ST 5016.14) is. also being written to verify hydrant operability and maintenance.
Procedure ST 5081.03, " Diesel Fuel 92 - Day Sample Test," indicates that the fuel in the storage tank is to be tested.
A fuel sample for the diesel firt pump day tank would, at least, be equally important to the condition of the fuel to be used during its emergency operation.
It is the fuel in the day tank which will be consumed for the first 8 to 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> of the pump's operation. For this reason, ST 5081.03 should be expanded to require testing the diesel " fuel in the diesel fire pump day tank.
5
Procedures ST 5016.11 and ST 5016.13 address the surveillance and maintenance of penetrations through plant fire barriers.
These procedures, however, do not consider either firt dcors or fi re dampers.
(See section 3.1.3 of this report for additional d,etails. ) All fire barrier penetrations, including fire doors
~~
and dampers, need to be inspected every 18 months in accordance with Technical Specification Paragraph 4.7.10.
Documentation of random surveillance procedures were reviewed.
(A list of the documents reviewed, along with dates and findings recorded, is included in Appendix 0 of this report.) The dates of the documents reviewed were recorded to determine if the required inspection frequency was being met. Any deficiencies noted were checked for accompnaying MWO identification. These MWO's were in turn pursued to verify-that all necessary changes had been completed within a reasonable time frame.
It was discovered while reviewing documentation that documents were not being entirely filled out as required at the beginning of 1981.
Af ter discussing this with the Fire Marshal, it was determined that this error was due to the training of a new employee. Recent documentation indicates that this problem has been appropriat'ely corrected.
Overall, documentation was found to be thorough and complete.
The established system s'eems to be well maintained and effective.
3.3 Fire Protection Systems and Ecuipment The plant audit included a review of the installed water supply and distribution system, fire detection and alarm system, and fire suppression systems.
All existing systems were found to be well maintained and in good working order.
6
n2--s--
s L
a.
3.3.1 Fire Suppression tlater Systems The fire water supply and distribution system was found to be adequate, reliaoile, and well maintained. Two separate fire.
pumcs taking suction from two separate water supplies.are available for plant safety related areas. One pump'is diesel e,ngine driven; the other is motor driven.
Procedures ST 5016.01, ST 5016.02, ST.5d16.08, ST 5016.10, and ST 5016.12 indicate that the water supply, pump operability and valve status is being adequately tested and maintained.
As mentioned previously, a procedure (ST 5016.16) is being written to conduct triennial flow tests and semi-annual flush tests of the water supply system as prescribed in the Tech-nical Specifications.
3.3.2 Fire Detection and Alarm Systems The fire detection and alarm system has recently been expanded and integrated into a computer system. This system is an excellent means of providing quick and accurate location of trouble areas protected by the system. Personnel at the station demonstrated the system during the audit and seem
' to be well trained in its use.
A recommendation in the previous audit called for additional fire detectors to be located in several bays in the 603' elev.
480V Switchgear Room (Room 428).
Installation of the additional detectors has'not been completed, but revision A to the Facility Change Request 79-017 has been written to accomplish the work.
3.3.3 Fire Suppression Systems Water suppression systems (deluge, pre-action or sprinkler) are still the predominant types provided at the Davis-8 esse Plant. All the systems are well maintained and are in gcod repair. No gaseous suppression Systems are provided at this time, although Halon 1301 is being.provided for the new Administration Building.
7
It is important to note that the station has connitted to comply to the NFPA 13 and/or 15 requirements.
The Nuclear Regulatory Comission can, therefore, audit the station by the letter of the code. The code is not specifically intended for nuclear applications; deviations will exist as a result.
Deviations may be sound and justified during design, but the
~~
objectives and philosophy behind them must always be approved and documented for future review. This documentation should be readily available to specifically justify deviations from commi tmen ts.
Documentation has been presented to justify deviations from NFPA 13 requirements as outlined in reconmen-dations 4.3.2 and 4.3.3 in the 1980 audi t report. Deviations to the code (such as distance of sprinkler heads below the ceiling) have been. researched and the design of piping, supports, obstruction of sprinkler spray patterns, and hydraulic analysis, ha's been verified to ensure that the systems will function as intended. Similar documentation is reportedly available for other NFPA code deviations.
3.4 Fire Brigade The fire brigade seems to be generally well prepared and equipped to handle probable fires occurring at the site.
3.4.1 Organization AD 1810.00 defines responsibilities associated with the fire brigade.
Specific personnel assigned to the brigade are listed in Adninistrative Memo No. 39-10 according to their shif ts.
A minimum of four members and one leader are now assigned to the brigade at all times. The brigade is comprised primarily of operations personnel but some security guards have received training so that they are available to supply supplementary man-power as required.
~
3.4.2 Training-Annual training requirements for the outside fire department 8
i w
n
~
.m mu a a n c-s.
e,...
M
are outlined in the Emergency Plan (Sections 8.1.1.c.4.b and 8.1.2.b) but responsibility for conducting this training has apparently not been assigned. This training was last conducted on January 29, 1980, and must be ccmplete'd during this calendar year to meet NRC comitments.
The training' outlines used for fire brigade and leadership training are comprehensive and useful. AD 1828.20 has not been revised since the previous audit, and as a result, personnel could be assigned to the brigade with insufficient training (in practice, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of training are usually required before assign-ment to the brigade). General brigade retraining is usually con-ducted for approximately two hours in each 5-week work cycle, but plant outages have interfered with this process in the past.
Brigade training comi,tments have been met since the 1980 audit and documentation has been improved. These training sessions of ten include practice as well.
No formal retraining is provided for personnel assigned the duty of brigade leader and this must be improved to satisfy
. NRC commitments.
Fire training facilities still do not include any capability to practice fighting fires indoors as required. As a result, practice sessions are limited to outdcor pan and wood fires. Breathing apparatus is not normally used during these sessions but separate training in its use is conducted regularly (new recomendation 5.1.1).
Attachment No. 2, Section 2.0, of the NRC Supplementary Guidance requires that:
" Practice sessions should be held for fire brigade members on the proper method of fighting various types of fires of similar magnitude, complexity, and difficulty as those which could occur in a nuclear power plant.
These sessions should provide brigade members with expe'rience in actual fire extinguishment and the use of emergency breathing apparatus under strenuous con-di tions. These practice sessions should be provided 9
at regular intervals but not to exceed one year for each fire brigade member."
^
Brigade drills have been improved recently. They are, now pre-planned and documented, but should be somewhat expanded to include all items listed in Attachment No. 2, Sections 2.0 and 4.0, of the NRC Supplementary Guidance (new recomendation 5.1.2).
The 5
drills now include a walk-through of realistic fire situations in the plant. " Surprise" fire drills will be conducted later this year as committed.
3.4.3 Brigade Eculoment The fire brigade is provided with sufficient equipment which is adequately distributed (previous recomendations 4.4.3 and 4.4.4 completed) and well maintained.
~
3.4.4 Pre-Fi re Emeroency Plans Comprehensive pre-fire plans have been developed for safety related areas of the plant. The information provided in them satisfies the NRC commitments.
~
4.0 UtlRESOLVED RECOMI4E!10AT10NS FROM THE 1980 ANNUAL /TRIENNI AL AUDIT 4.1 Administrative Control of Fire Hazards 4.1.1 AD 1844.00 should be revised to control ignition sources by requiring the following features in accordance with Attachment No. 4 Sections 2.a and c, NRC Supplementary Guidance on Nuclear Plant Fire Protection Fu,nctional Responsibilities Administrative Controls and Quality Assurance: 1) persons responsible for authorizing permits should have received basic training in fire fighting and fire prevention; 2) personnel assigned duties as a fire watch should have been trained to prevent and fight fires (revised recomendation 4.1.1 from 1980 audit).
4.1.2 A procedure should be developed to require a functional test of all required fire doors and fire dampers at least every 18 months in accordance with Technical Specification" paragraph 4.7.10 (pre-vious recomendation 4.1.4 from the 1980 audi.t).
10
. ~
c 4.2 Fire Protection Ecuioment Surveillance Procedures 4.2.1 A procedurt should be established to conduct a flow test of the distribution system every three years in accordance with Technical Comparison ' ith previous Specifications, paragraph 4.7.9.1.1.g.
w tests will indicate interior condition of underground piping.
A "C" factor calculation is reconwnended to be part of this test (recommendation 4.2.2 from 1980 audit). This item will reportedly be resolved by ST 5016.16 which is now being written.
4.2.2 A procedure should be established, perhaps in conjunction with ST 5081.03, to test the diesel fuel in the diesel driven fire pump day tank in accordance with the requirements listed in the Technical Specifications paragraph 4.7.9.1.2.b (recomendation 4.2.4 from 1980 audi t).
~
4.3 Fire protection Systems and Ecuioment 4.3.1 Install additional p0C detectors in the bays not currently covered at the' 603-ft. elevation, 480V Switchgear Room'428 (recommendation 4.3.1 from the 1980 audit).
FCR 79-017, Revision A, has been issued to resolve this recomendation.
4.4 Fire Brigade 4.4.1 AD 1828.20 should be revised to require basic fire training before personnel are asrigned firt brigade duties (revised recomendation 4.4.1 from 1980 audit).
4.4.2 All personnel designated as fire brigade leaders should receive regular documented requalification training (revised recomendation 4.4.2 from 1980 audit).
4.4.3 A fireground training facility should be provided to allow for more realistic " hands-on" training. The training facility should provide enclosed areas with heat and smoke to simulate fires indoors (recomendation 4.4.7 from 1980 audit).
11
4.4.4 Responsibility for annual training of the outside fire depart-ment as outlined in Sections 8.1.c.4.b and 8.1.2.b of the Emergency Plan should be assigned. This training should be completed during 1981 and should incluce a joint dr'ill between the fire brigade and outside fire dt:partment (revised recommen-dation 4.4.8 from 1980 audit).
5.0 NEW RECOMENDATIONS 5.1 Fire Bricade 5.1.1 Fire brigade practice sessions should include the use of self-contained ~or9 thing apparatus under strenuous fire fighting condi tions.
~
5.1.2 The regular fire brigade drill sessions should be expanded to include the information required by Attachment No 2 (Sections 2.0 and 4.0) NRC Supplementary Guidance on Nuclear Plant Fire Protection Functional Responsibilities, Administrative Controls and Quality Assurance.
S.0 SUMARY The plant is provided with extensive, well designed fire detection and fixed water suppression systems. Administrative control of fire hazards, ignition sources, and housekeeping was found to be very good.
The fire brigade seems to be well prepared to handle the fires expected.
It was not possible to observe the b'rigade during an unannounced drill because of the plant operating status, so this subjective evaluation could not be substantiated.
Brigade training is fairly good but is limited by the facilities available.
-Recomendations relating to these topics are contained in Sections 4 and 5 of this audit report.
Eleven of these recomendations have been carried over from the 1980 Annual / Triennial audit, either as revised or repeat recomendations (Section 4).
Only two new recomendations, related to, the Fire Brigade, have been identified by this audit (Section 5).
12
=-
~
~.
O
.e
~
. v..
.1
- e..
APPENDIX A Pre and Post Audit Conference Attendance e
ee 9
e M
O e
f 6 L?:
4 9.
se A.
?-
9 e
e 9
l
(
S f
l l
l t
I DATE:
J u t. v 2 0, /Pe/
Tnc:
9 00 4 m '
Y LOCATICu:
DE 57A 7 /csJ
' *9
(
e, s
j I
w AUDIT ID: '" 7.9 /
a (PRE /N8:*) AUDIT CONF 2RENCE ATTENDANCE NAMI ORGANI::ATION TITLE
- l D :.: e
.J S> :r..u
- u...
r%-., R,-~7..,Ec:.0:
uss c,-m
, u. -
- k. E. kOW9t.E.L Yice 9 R.ES ID G4/T W
T 2-G 5+=. S% I.
H. Oh 7E co QA 4 cro. 7,c,v, i
o o
U D 'n e ~
Y[(o OA 0 9$. 0.h S',.o ~a v
Fun e os i Teco oss As-r. n. ca.
8,//
A),ss, a.
fpg o,oS X,, pp,7 p e 7~,.,,,,
~
co a 1
i
?
)
9 l
T A
8 b
e
DATE:
d u c y 27. /? I/
TIME:
// # #O A N LOCATION:
D 8 I TA T/#d Qjjr ;,ggg AUDIT ID:
70 /
~
s CURC/ POST) AUDIT CONFIRENCE ATTENDANCE NAME ORGANIZATION TITLE 6
f U<L,)
t nee. CA Aru.7eca. Og A i//. L%
Fwe b?s Ru Pas 7 coon.
M. 6.
M OW PJJL 94C. I4 c.
Yics hRas /bs*
kn9-sl Sw T-CCO C)lo S.
AS3t Q Osc bJa PAC., %.
F m. c Pe n. w - esa
%'d w RE'.
se. Ae r h
~
~
L WI&
T S c.s og Ok L o, ~..
nc u.. --.. -- -- - -
n h e_=;
D8 ops 34 SL,, +,
f.
D l-3 A) Ab D.0 0er O W 5 24 r*
u i
I THE TOLECO ECISCN COMPANY EQ3CN P'. AZA 300.V.ACISCN AVENUE TCLF.00. CHiO 43652
~
)
l s
.~
.s :
APPENDIX B Technical Soecification Surveillance I
e O
.=
6 O
s 0
O e
o e
h e
e D
F b
e e
6 e*
y a
h I
e
u-.
Technical Specification Procedure Surveillance No.
Subjec t frequency Number Frequency e
FIRE DETECTION INSTRUMENTATION 4.3.3.8.2 Supervised circuits 6 mo.
ST 5016.04 6 mo.
(operability test) 18 mo.
ST 5016.06 18 ro.
4.3.3.8.1 Channel functional Test Accessible 6 mo.
5016.04 6 mo.
Inaccessible 18 mo.
5016.06 18 mo.
PLANT SYSTEMS - FIRE SUPPRESSION GMTER SYSTEM 4.7.9.1.1 a.
wa ter vc'ume 7 days ST 5016.02 weekly b.
Start Elec. Pump /
31 da,ys ST 5016.02 weekly 15 min. (underground) c.
correct position valves 31 days ST 5016.01 weekly ST 5016.02 weekly ST 5016.09 monthly d.
System Flush 6 months ST 5016.16 e.
valve cycle 12 months ST 5016.10 annual f.
System functional Test 18 months 2.
Pump flow test elec.
ST 5016.08 annual Pump flow test diesel ST 5016.12 annual 4.
Sequential starts diesel ST 5016.01 weekly Sequential starts elec.
ST 5016.02 weekly 9
Flow test 3 years ST 5016.16 4.7.9.1.2 Diesel Pump (operable)
(al) Fuel level 31 days ST 5016.01 weekly (a2) Diesel start 31 days ST 5016.01 weekly (b)
Fuel sample 92 days ST 5081.03 Quarte rly 4.7.9.l.2.c Preventive Maintenance 18 months PM's O
f 8
8-11 Technical Specification Procedure Surveillance No.
Subject frequency Number frequency 4.7.9.1.3 Battery check (al) Elec. level 7 days ST 5016.01 weekly (a2) Battery voltage 7 days ST 5016.01 weekly (b)
Spec. gravity 92 days ST 5084.03.0 Quarterly, (ct) visual 18 mo.
ST 5016.01 weekly (c2) connection 18 mo.
ST 5016.01 weekly PLANT SYSTEMS - SPRAY AND/OR SPRINKLER SYSTEMS
- 4. 7. 9. 2 Spray / Sprinkler Sys Operable (a) valve position 31 days ST 5016.09 (b) valve cycle 12 mo.
ST 5016.07 Quarterly ST 5016.15 annual i
(ci) System functional test adto valves 18 mo.
ST 5016.07 Quarterly 18 mo.
ST 5016.15 annual (c2) Visual of valves 18 mo.
ST 5016.07 Quarterly 18 mo.
ST 5016.15 annual (c3) visual inspection 18 mo.
ST 5016.07 Quacterly nozzle spray pattern ST 5016.15 annual (d) airflow test through each 3 years ST 5016.15 3 years Open head & unobstructed PLANT SYSTEMS - FIRE HOSE STATIONS
'.7.9.3 Fire llose Stations shown operable:
4 (a) visual inspection 31 dyas ST 5016.09 monthly annual (bl) remove hose--rerack 18 mo.
ST 5016.09 annual (b2) inspect gaskets 18 mo.
.ST 5016.09 i
(ct) valve operability 3 years ST 5016.16 (c2) hydro test 3 years ST 5016.09 annual af ter maint. or repair
- Procedure 5016.16 expected to come out in Oc'tober of 1981
.s s
c..
APPENDIX C Procedures Reviewed e
49 O
N 7
=
W e
e
.s e
APPENDIX C Procedures Reviewed During Insoection No.
Procedure Date of Acoroval AD 1801.04 Resolution of Test Deficiencies AD 1805.00 Procedure Preparation AD 1807.00 Quality Control AD 1844.00 Maintenance AD 1810.00 Fire Protection Program AD 1810.01 Control of Combustibles AD 1827.00 Emergency Plan AD 1828.00 Personnel Training Program AD 1828.04 Personnel Training Records AD 1828.20 Fire Brigade Training AD 1838.01 Surveillance & Periodic Testing Scheduling AD 1851.00 General Welding Procedure AD 1851.03 Storage, Handling & Issue of Welding Material
~
EP 1202.35 Fire Emergency Procedure AD 1810.03 Fire Pre-plans (Spot check)
SP 1105.14 Fire Detection System l
SP 1102.16 Station Fire Protection System
{
ST 5016.01 Diesel Pump weekly test Rev.- 3 1/24/81 ST 5016.02 Electric Pump Weellly Test' Rev. 2 12/05/81 ST 5016.04 Fire Detection Channel Functional Test Rev. 2 2/20/81
& Supervisory Circuit Checks ST 5016.06 Fire Detector Functional Tes't 4/22/80 ST 5016.07 Automatic Sprinkler Test P.ev. 1 3/23/81 ST 5016.08 Electric Fire Pump Flow Test Rev. 2 6/16/81 ST 5016.09 Fire Hose, Fire Hose Stations. Fire System Valve Rev. 7 3/03/81 Testing ST 5016.10 Fire Protection System Valve Surveillance Rev. 2 6/09/81 ST 5016.11 Fire Protection System Barrier Surveillance Rev. 1 4/22/78 ST 5016.12 Diesel Pump Annual Test Rev. 3 6/08/81 e
C-i l
l
.y No.
Procedure Date of Aonroval ST 5016.13
~ Visual Inspection of Fire Barriers
_. Rev. 1 5/08/80 6/08/51 ST 5016.15 Diesel Generator & Water Curtain Deluge Annu 1 Test ST 5081.03 Diesel Fuel 92-Day Sample Test ST 5084.03 Quarterly Battery Surveillance oo e
69 4
28 4
em e= e me m eummmane G
G e
am 6
C-ii
,s
~
.e.
n APPENDIX 0 Test Procedure Documentation Reviewed 98 e=
e J
O e
i 9
er 9
L a
~'A s
APPEriOIX 0 Test Procedure Documentation Reviewed,
PT No.
Procedure frequency Dates and Findines ST 5016.01 Diesel F.P.S. Pump weekly 7/16/81 - day tank level Icw, leaking oil from throttle assemb.
WP submitted 7/16/81.
test suspended.
7/02/81 - #1 Battery cap on the "A" Battery broken 6/25/81 6/18/81 1/22/81 - Deficiency. Diesel Fire Pump started at 118# in-stead of 100# 15 psi.
Deficiency. FP 1045 must be left isolated due to large leak in gauge line (no MR found).
ST 5016.02 F.P.S. Elec. Pump weekly 7/13/81 7/05/81 7/02/81 6/22/81 1/28/81 1/17/81 - Pressure switch does not reliably pick up upon restoraton of pressure abcVe the sarpt. causing pump restart ST 5016.04 Accessible Detector 6 months 4/20/81 - Detector doesn't exist Channel Functional Test T-mod submitted Alarm extender doesn't exist T-mod submitted 05 8619 m didn't alann MWO (IC-365-81) 4/17/81 rer'. aced detector DS869Dn wlnew one from stock performed appli:7bi c parts of ST 5016.04 o.k.
ST 5016.05 Fire Detector Functional 18 months 5/16/80 - Performed to ensure oper-Test able after repair.
MWO-IC547-79 4/12/80 - Performed to meet schedule ensure operable after repair. Found TS 8507G l
inoperable, slide link found inoperable in dis-connected position.
Returned to procer position repl3ced TS 8507H.
l
PT No.
Procedure Frecuency Dates and Findings St 5016.07 F.P. Sys Auto Sprink. Sys Cuarterly 4/7/81 - Valve tamper alarm does not annunciate. FCR 80-254 5/22/81 work completed and TP 550.5 run to verify.
1/16/81 - Pressure Switch did not activate. Switches cali-brated and are operable 2/27/81.
ST 5016.08 Elect Fire Pump Flow Annual 10/26/79 - Excessively good test Test results out of range so rev, for acceptance criteria.
ST 5016.09 Fire Hose, Hose Sta.
monthly 3/24/81 - FP-119 found closed. Contacte' Fire System operations & opened 3/26/81 T curb wrench missing Hose House _ III. Submitted WR (no # listed) had one made and placed in hose house 4/27/81 "T" curb wrench missing -
WR submitted to have one made (no WR # listed) 4/27/81 2/20/81 1/20/81 ST 5016.10 F.P. System Valve Test Annual 5/19/78 - (no tool to cycle underground valves) new tool purchased 6/1/78.
7/27/78 - (150 valve guard pipe broken MWR submitted, retested) 11/21/79 - performed to ensure FP59 opr.
11/20/79 - performed to ensure FP63 opr.
6/27/79 - performed to meet schedule (FP61, FP74, & FP108 buried WR/MWO-79-2512 issued)' valves dug out, barricades installed 7/25/79-7/26/79 FP104 broken, MR/MWO 79-2508 repaired 7/23/79 6/11/80 - valve is covered by gravel WR/MWO submitted (No #)
valve uncovered and stroked 6/26/80 e
0-11 M
A
..,[.
PT No.
Procedure Frecuency Dates and Findincs ST 5016.10 F.P. Valve Surveillance 6/11/80 - PIV broken WR/MWO sub., no #,
FP104-filled, stroked &
PIV repaired 6/28/80 3/07/80 - Performed to ensure opr.
a f te r 'maint.
Jockey fire pump out of se rv.
Elec. firepump in service maint. sys. 9
> 135 t 5 psig.
Solution: This is acceptable condition for testing FP20 prerequisite intent satisfied 3/18/80 (no MWOf)
ST 5016.11 F.P.S. Barriers Test 2/14/80 - new 1h" cond. & 2b" conduit do not appear on floor plan 3
Solution:
Penet. SAT, Pro-cedure being MOD to classify.
545-104A-1118-2, 3, 4, 5, 6 Bisco gauged.
Solution: not fire walls and in test in error T-M00 being prepared and WR being sutmitted. 2/16/80 ST 5016.12 Di.asel Annual Fire Pump Annual 10/15/79 AD 1844.00 Maintenance 2/09/81 - MWO IC-266-81 issued for F0Z227 (damper alams in fire watch in effect).
2/10/81 - retested, found o.k.
2/05/81 - MWO IC-262-81 Turbine bearing 2 & 3 fire detector thermal kept going off, went off at 1660F, replaced.
w/ proper 2090F set detector.
t i
0-111 L
.. ~ ~ -..... -
,s APPENDIX.E[
~
Fire Door Problems 4,.
s.
e * *
- tS 8
e O
I
.O
[
oo. og.
.s a
APPENDIX E Fire Door Problems Door 333 to H2 Seal Oil Unit blo' ked open c
(not safety related) 1 s
Door 4298 to the Battery Room does not latch
~
(safety related)
Both doors to Auxiliary Boiler blocked open (not safety related) 9 6
9 e
a 0-i
-~
~
~_m
- -_ ~
o o
CMD nsci DATE I?;)ll A CUMPAtJY ML Moll ANDUM tuous W
Decerber 3; 1980.,
I j
J. H. Shortt j ' gi g / :,..,/,,.
v s.ou 3,.y
~~
R. P. Crouse f n4 _
f fl}. AR gyg susact ggn,4 Nuclear Plant Fire Protection Program,10CTR Part 50, Appendix R The NRC has published 10CTR Part 30, Appendix R, on Fire Protection which would require certain changes for fire protection in nuclear plants operating prior to January 1,1979.
Af ter reviewing the attached material, I am not really sure of the significance of the separation with fire barrier criteria relating to fire protection of Safety Shutdown Capability.
Does the Reactor Coolant Pump Lubricating 011 Collection System, as presently installed, meet the necessary seismic qualifications?
' hat modifications are necessary tu the present
=
fire protection system in addition to those changes required for Safety Shutdown Capability Emergency Lighting, and Reactor Coolant Pump Lubricating 011 Collection System.
Please review the attached material to deterr.ine what actions must b taken and upon what time schedule the actions must be i=plemented RPC:aa cc:
T. D. Murray L CC 3~~" A'.
JHS Q 7 f~
f
_C_3Jf._l_ _
WIT f/ F._L _I ? I 5
__CSM :
!e i _ _l
~~
-~
- n. a a._- m__.
't..U -
~
3, ?.e.1.l. _____.L_.v_2.1 i f l __;_._
GMGi m,f -- 4,4_y9 cac e
lD
<T I FILE: '7/'~v /!
l M*
o i m (v
i I
__l -
i
=
fi -ltf i
i l
W wp,;3;sa.r=n ky aii j.
22ph f n z o-zo a-x V s
- 2 5 ss ? ! 6 e 4 E 5 s
I l
l l
?.
y m
i s
e _;..
l n..
.r ;.
l
,g:.,. A..-
1 i
~
-s
-.. **. i=nwj. -.,,. ~.* p[
- h*I.
.'.,ga.-
A..
.=
6
-g %.
F **,;,,. 3 ** ', *L
' ; "l.
.,..,4'.
m.
.s e
.s p.,
=.
g.%, ',, g. **
5
- :.3.., '
.,3.p.
.. oft,.. >.
.h,....,..
..
- t ' g.' 4'. /'. p-J.
.-* a t.,.
N 4'
+-
==
s
. j g.
. 'aj
<e=sA o
- .f..e%Fahn. ;.
~
s,*mh Af _
,,* * *sfl} r f
p'
' = =
' * - D
- l., h, -
9,,e.
- , %.b*. e, a
ie
,, e ans.
j y,
4
( 8' e.
,.. a.,.,.
e,
,*as
't
.e
.=..;
=.
g-g a
,* eP,. ns.
- s
.,,tj.,..
7...
g
- 3....,
a p.
.. ' ens.r,
'.ak, p _ *.
.,,...q.
..a g,.
33 @
p.v'.
^
.>T
,, = = V'
, 14 a
a
..f g
or,.,,,,,.g
.-p 8
- ', g' y g
, f 5.
4.e.
A p.
,'S.
,M,, e,...
,*, *... i'
'*L
'y w
.'.y,,.?..
.(e,_
_.w
%g,
,4
- t, %. x.
4.
-R 8
s 2'
e y
.Q U._
P ulm
.*I i
.- - [* k. 5
, g,,
f.
,.s.
4,
{ , +
t
.s p
v_
i f
I' 2,
, e 2
.g j#
-a,
.a
+
. g. '..
-p...
' J p,,..
1 s.
o.
. W.,%
2 a
,5 g.
.. s
[, 'A
.P k
y.
+
g,--e',
_..); 1, '
' p.
- .f" _4""
(6* *
"e
~
'4 P.* *.
. p..%.
l s,
t' 1
\\' *...,
4
. *., * =
a
, L
...t**
,/>
a_
1
' ' " ',.. [. -.,. '.
e ' 49 4
t M. c%
w
.f' ;, 3, '.....
h *, ; *
'w
- ~
c - p %.. ',' z g,
A s,.
4 4
m c.,.
3..,,
b;
.. e,' - s, /..r ' e
~r' ~ ft
- r -
w t,.gaws. ewe== 4..'
sp h d i.,.
,Wt:->
~.
... ~....
....m. p
ym=
s.
i,~, - w e'
a 7
.g.. g..,,
.....,,~..w,e...
....~....
9.,,
-.p
....,,-.- #,y..
- e..
~
.6ef. :o%%-. f
- - R
-~~'*"t*'-* ~ ~ ~ ' -
4
. " ?,.W' eg ?.5.*-
4
'?..
','.m..,
3,.'
" # *]s st. *.
g g
s*'.
e' - * '. '.
- yk '
2
- c
=.
w..
.-g as
'.d'..e**.'s--
- Q F**'*
y.
"' a e.
. 'y
,% s s*
.t 3,,g.
. ;.+.
e A....s. (.,.. _
,,,4
.g. et.w,,r.,,,
e a
.a.
g.., a,
., g,
..p 1 to..
M.
g...jp*
.i,*
e A
g,- t-
.,,,.e,y. % g..
.. a-.
4....
.r
,k-e
.a
.y f.,W p
,r-g -,sr -
[_
,.3,.,
,,...,.4,be -.. d~ j. c},* _ y.
- ..,,
- e 4,r,
3e,' -
- ,- - 4
+
- l..
S'.
.a
- 4. #.h. d.%y*h fd,W.p' p g W, (
..m,*
,. p,
e ig.
~.,~
7: -.
.'f:.,
.y,,. y -
T
.n g
.~ '
y,
. *'.y.,..
A..
J y %.
.c s
.' a( '
[,.
E' g.
*v q*
2,
. s
\\*
T
,.s sj
,,,. ', k. 4. ;-
,e,,
4.,
pV n%.'
4 -
. 1
- ' e
..q \\^
~
- 4
.eq w.
.s
..e 1
.*-..-p..
=
w.
.w
. b e 4
- e.., s. w _- -
p e,,....
e
^^..
3-
. y
=
y*.
.N 6
,p-
^ W h _,,}
a e
i ha I
I
.,,..s.-.,.k.
..s 3
,.8
, a.:
h.'
[e.
s.
g
.s
,--'Y
[* '~
- b s.
z.
Q.
'.g_
4
'y
- f A 9 -
4p..
9 1
n.
- i
~. ' " _ ',,
e.
s.
,,a.
m-P a 6, d
+
'A
.d y
..as g
m.
.- %[.
~,*
- 4 4
%~
6*
a, g
'a s, E.
r
-e
.o e '...
- , 3 ' _.,
g R,..., <.
(
,/ -
+
94 v
g4 L.
..y
- g w-
~.
^
s - +.:.
_.., r g;
t..
.; g s
.a.
.g
- ' i.,".-,,,, '.,,.
,s e.
.e e
p,-
5(.
'. ".r I -
', ' l
. - ' ' {
A '
e y.,, i s..
L-
- ... q....
5
..t,. ~
.w o
s..
4.,.,...,...
..g 9,. -
as, y,...
f.
pe,.-
-e,..
'A
.. 4
-.s
... =.
s '
e '
[..
~
.w n"un. s '
g gf
.p b..'.
d.
a
..s g'.,
.}
',i a
}
q 0 '., *,',,, -
.O +
e,,
f..
s.c g i" ; ^. : ;
-t
- 5'v-.
. -- 1
- 7.. A
. ts. '. 7 n
?
, s. 7;.;.
.. :,. -...'. -...... u.
e
?. S, ;..
4 --,
e.
4
. - 4j.
..,.,. 3,. p.:,..
$4.,
t, 0., ;..+.x.. - j..-
4 y-
- 4*.
y.
- aM e,
....-~s
.s
..y
,. p.
r... :...;,;. _
t,.,
., ~.. g.. ; z.. q.
- .y,.
~.
b..r
_ l'
.e.
.u
.y
..s
.. c
.r * ~.'
. _..s '
c '. 7.e(
.V
_g_
.g s.
...V-k,.
g # ;g..-,, w.:. -,.
'.. 3
-. g ~~.
.)p k.
2 1
Q gn i
troppu S
t 1 ne Em Rh Uc Sa Ot Lt CA
(
lE
ENCLOSUPE 1: ATTACHMENTS SUPPORTING OUESTION 12
U AU7T70RTry pg LNITED STATES OF AMIRICA cecm) p uma-y 7 jg73 > 9 NUCLEAR RE E T RY CCMMISSICN 7
t-r*
00 30T REMOVE s "="C~
In the Matter of
)
//
)
- 14
':HE TLEDO EDISCN CCMPANY AND
)
f.
"EE C:.EVFJAND ELEC:RIC ILLUMINATING
)
Docket No. 50-346 CCMPANY
)
)
y Davis-Besse Nuclear Power Station,
)
L$
Unit No. 1
)
l3
(
CRDER I.
The Toledo Edison Company (TI:~J) and ne Cleveland Electric Illuminating 3
Company (the licensees), are Mders of Facility Operating License No.
NPF-3 which authori::es the operation of the nuclear power reac*ar known
!T-j as Davis-Besse Nuclear Power Station, Unit No.1 (the facility or Davis-Besse 1), at steady state power levels not in excess of 2772 megawatts ther=al (rated power). De facility is a Babcock & Wilcox (B&W) designed pressurized water reactor (NR) located at the licensees' site in Ottawa County, Ohio.
II.
In the course of its evaluation to date of the accident at the tree Mile Islani Unit No. 2 facility, which utilizes a B&W designed WR, j
the Nuclear Regulatory Camission staff has ascertained that B&W designed l
?@Y 3p5@ tis M 2
Pf m
7590-01
.o reactors appear to be unusually sensitive to certain off-normal transient conditions originating in the secondary system. We features of the B&W I
design that contribute to this sensitivity are:
(1) design of the steam generators to operate with relatively small liquid volumes in the second-ary side; (2) the lack of direct initiation of reactor trip up:n the occur-rence of off-nor=al conditions in the feedwater system; (3) reliance on an integrated control system (ICS) to automatically regulate feedwater flow; (4) actuation before reactor trip of a pilot-operated relief valve on the primary system pressurizer (which, if the valve sticks open, can aggravate the event); and (5) a low steam generator elevation (relative to the reactor vessel) which provides a smaller driving head for natural circulation.*
Because of these features, B&W designed reactors place more reliance on c,
the reliability and performance characteristics of the auxiliary feed-(,
water system, the ICS, and the mergency core cooliry system (ECCS) per-formance to recover from frequent anticipated trensients, such as loss of offsite power and loss of normal feedwater, than do other PWR designs.
mis, in turn, places a large burden en the plant operators in the event of off-normal system behavior durirq such anticipated transients.
)
i l
- It is noted that although features nunbers 3 and 5 do not apply to Davis-Gesse 1 to the same extent as they apply to other currently licensed B&W designed reactors, the other features are fully apoli-cable.
s
a 7590-01 As a result of a preli:tinary review of the tree Mile Island Unit No.
2 accident chronology, the NRC staff initially identified several htr.an errors that occurred during the accident and contributed significantly to its severity. All holders of operating licenses were subsecuently instructed to take a number of imediate actions to avoid repetition of these errors, in accordance with bulletins issued by the Corr.tission's Office of Inspection and Enforcement (IE).
In addition, the IRC staff began an irunediate reevaluation of the design features of S&W reactors to determine whether additional safety correctiens or i.:provements a
were necessary with respect to these reactors. B is evaluation involved nur:.erous meetings with B&W and certain of the affected licensees.
A The evaluation identified design features as discussed above Wich
.~
] <,..)
indicated that B&W designed reactors are unusually sensitive to certain
~./
off-normal transient conditions originating in the secondary system.
As a result, an additional bulletin was issued by II which instrteted holders of operating licenses for B&W reactors to take further actions, l
' ncluding i:tsediate changes to decrease the reactor high pressure trip i
point ard increase the pressuri:er pilot-operated relief valve setting.
Also, as a result of this evaluatien, the 1GC staff identified certain other safety concerns that warranted additional short-term design and procedural cha.ges at operating facilities having B&W designed reactors.
.a-7590-01 These were identified as items (a) through (e) on page 1-7 of the Office of Nuclear Reactor Regulation Status Report to the Comission of April 25, 1979.
After a series of discussions between the NRC staff and the licensees concerning possible design modifications and charges in operating pro-cedures, the licensees agreed in letters dated April 27 and May 4, 1979, to implement promptly the following actions:
(a) Review all aspects of the safety grade auxiliary feedwater system to further upgrade components for added reliability and performance. Present modifications will include the addition of dynamic braking en the auxiliary feedpump turbine speed changer and provision of means for control r
room verification of the auxiliary feedwater flow to the steam generators. 7his means of verification will be provided for one steam generator prior to startup from the present maintenance outage and for the other steam generator as soon as vendor-supplied equipment is available (estimated date is June 1, 1979). In addition, the licensees will review and verify the adequacy of the auxiliary feedwater system capacity.
(b) Revise operating procedures as necessary to eliminate the option of using the Integrated Control System as a backup maans for contro11 irs auxiliary feedwarer flow.
l i
~ -.. -,,,
-, +
1 4
7590-01 '
1 (c) Implement a hard-wired control-grade reactor trip that would i
be actuated on less of main feedwater and/or turbine trip.
(d) Complete analyses for potential small breaks and develop and implanent operating instructions to define operator action.
1 (e) All licensed reactor operators and senior reac*er operators will have completed the Wree Mile Island Unit No. 2 simulator training at B&W.
(f) Sulmit a reevaluation of the ECO analysis of the need for automatic or administrative control of steam generator level setpoints during auxiliary feedwater system operation, previously submitted by TKO letter of Dece=ber.22,1978, in light of the Bree Mile Island Unit No. 2 incident.
(g) Submit a review of the previous TECO evaluation of the September 24, 1977 event involving equi ment problems and depress-urization of the primary system at Davis-Besse 1 in light of the tree Mile Island Unit No. 2 incident.
l In its letters the licensees also stated that the actions listed in (a) through (g) above would, except as noted in item (a), be completed prior to startup fr:m the current maintenance outage.
7590-01
- In addition to these :redifications to be implemented pecer.ptly, the licensees have also proposed to carry out certain additional long-term modifications to further enhan:e the capability and reliability of the reactor to re-spend to various transient events. tese are:
- te licensees will continue to review performance of the auxiliary feed-water system for assurance of reliability and performance.
- te licensees will suMit a failure mode and effects analysis of the ICS to the NRC staff as soon as practicable. te licensees stated that this analysis is now underway with high priority by B&W.
i
~.
- he reactor trip following loss of main feedwater and/or trip of the turbine to be installed promptly cursuant to this Order F
t~
will thereafter be upgraded so that the components are safety grade. te licensees will submit this design to the NRC staff for review.
- Continued attention will be given to transient analysis and procedures for manageoent of small breaks.
- h licensees will continue reactor operator training and drilling of respense procedures to assure a high state of preparedness.
+
O
~
r--
.~.--,,,
7590-01 0 l l
':he comission has concluded that the prompt actions set forth as (a) through (g) above are necessary to provide added reliability to the reactor system to respond safely to feedwater transients and should be confi::ned by a Cocznission order.
The Comission finds that operation of Davis-Besse 1 should not be re-sumed until the actions described in paragraphs (a) through (g) above, with the exception as noted in item (a), have been satisfactorily completed.
For the foregoing reasons, the Comission has found that the p2blic health, safety and interest require that this Order be effective Lmmedi-ately.
e 4
C, III.
Copies of the following documents are available for inspection at the Cor:riission's Public Document. Room at 1717 H Street, N.W., Washington, D.C.
20555, and are being placed in the Comission's local public document room in the Ida Rupp Public Library, 310 Madison Street, Port Clinton, Ohio 43452:
(1) Office of Nuclear Reactor Regulation Status Report on Feedwater Transients in B&W Plants, April 25, 1979.
1 l
7590-01
- C (2) Letters from I.cwell E. Roe (TECO) to Harold Centon (NRR) dated April 27 and May 4, 1979.
f 71.
Accordingly, pursuant to the Atomic Energy Act of 1954, as mended, and the Cocmission's Rules and Regulations in 10 CFri Parts 2 and 50, IT IS EEREBY ORDERED W.AT:
(1) The licensees shall take the following actions with respect to Davis-Besse 1:
(a) Review all aspects of the safety grade auxiliary feedwater system to further upgrada compnents for adde:1 reliability and performance. Present modifications will include the addition of dynamic braking cn the auxiliary feedpump turbine h speed changer and provision of means for control room veri-fictition of the auxiliary feedwater flow to the steam generators.
Ibis means of verification will be provided for one steam generator prior to startup from the present maintenance outage and for the other steam generator as soon as vendor-supplied ecuipnent is available (estimated date is June 1,1979).
In addition, the licensees will review and verify the adegacy of the auxiliary feedwater system capacity.
(b)
Revise operating procedures as necessary to eliminate the option of using the Integrated Centrol System as a backup means l
for controlling the auxiliary feedwater system.
i I
l l
1
7590-01
. (c)
Implement a hard-wired control-grade reactor trip that would be actuated cn less of main feedwater and/or turbine trip.
(d) cosipleta analyses for potential small breaks and develop and implement operating instructions to defino operator action.
(e) All licensed reactor operators and senior reactor operators will have coe:pleted the tree Mile Island Unit No. 2 simulator trainirag at B&W.
(f)
Submit a reevaluation of the ECO analysis of the need for automatic or administrative control of steam generator level setpoints during auxiliary feedwater sys*m operation previously v
sibnitted by TECO letter dated December 22, 1978, in light of the t rea Mile Island No. 2 incident.
(g)
Submit a review of the previous TECO evaluation of the September 24, 1971 event involving equipnent problems and depressurization of the primary system at Davis-Besse 1 in light of the 'Ihree Mile Islard thit No. 2 incident.
(2) The licensees shall maintain Davis-Besse 1 in a shutdown condition until items (a) through (g) in paragraph (1), except as noted in item (a), above are satisfactorily coc:pleted. Satisfactory completion will recuire confir-
[
mation by the Director, Office of Nuclear Reactor Regulation, that the i
o I
l
7590--01 actions specified have been taken, the specified analyses are acceptable, and the specified implementing procedures ar e
appropriate.
i (3) ne licensees shall as pecrcptly as practicable also accompl the long-term modifications set forth in Section II of this Order.
v.
Within twenty (20) days of the date of this Orde r, the licensees or any person whose interest may be affected by this Order may requ est a hearing wit.5 respect to this Order.
Any such request shall not stay the imediate effectiveness of this Order.
TfE NUC '
RE'r.ATORY CCMMISSICN J
[OL %
... V i
( $amuel J. C-ik Secreta:V od the Comission Dated at; Washirston, D.C.,
this $+3 day of May 1979.
I
(
r k
i
r q
l zu sys - SJ l
TCLEoD
%mm EDISDN Docket No. 50-346 LCWELL E. RCE License No. NPF-3 wo e,....,
o Serial No. 506 4tay 22, 1979 Director of Nuclear Reactor Reg.ilation Attention: Mr. Robert N. Reid, Chief Operating Reactors Branch No. 4 Division of Operating Reactors United States Nuclear Regulatory Con: mission Washington, D. C. 20555
Dear Mr. Raid:
Under data f May 7, 1979, B&W transmitted to Dr. Roger Mattson the document
"$valuatit-of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plants, Volumes I & II", which provided information related to the plant behavior during certain transients with special emphasis on "arious small break scenarios. This information was in response to previous
.cc:ssicsents of B&W to supply such information and was intended to be in respense to com=it=ents in my April 27, 1979 letter (Serial No. 497). The information in the B&W document was for all B&W 177 Fuel Assembly plants in-cluding a lowered loop configuration, and the pingle raised loop configuration Chich is the Davis-Besse Nuclear Power Station, Unit 1.
Our detailed review of these documents showed that although the Davis-Besse configurations and details were generally included, the documents were pre-pared more specifically for the lowered loop configuration plants. This could lead to possible misinterpretation in translating these results specifically for Davis-Besse operation. En order to provide a more meaningful document for sur use and hopefully for yours, while providing additional information pupplementing Volumes I & II, we have worked with B&W to develop Voltme III which includes information specific for Davis-Besse Unit 1.
This Volume III will be the source document from which changes to existing station procedures or new procedures are being developed as a result of these further analyses.
ga Yours very truly, O
fph{
b
./,-
/^
h b
LER: cts cc: J. Zwet:1g Operating Reactors Branch No. 4 DivisiopCG EDSCN CCMgNYReactors - USNRC - Washin2 ton D. C. 20555 of Ooeratin THE TC ECISCN PLAZA 300 MAC:SCN AgENUE TCLE::0. CHC 436'
[ *1 4
y
s i,
D EVALUATION OF TRANSIENT BEHAVIOR AND SMALL REACTOR COOLANT SYSTEM BREAKS IN THE 177 FUEL ASSEMBLY PLANT 4
t.
, ' ~
MAY.16, 1979 1
i VOLUME 3 RAlsED Loor PLAT:T (DAVIS BessE 1) n Babcock &Wilcox Revisice 1 MayfJ,9,1979 geos2-90 W h l
SAPf ^
EVALUATION OF TRANSIENT BEHAVIOR AND SMALL REACTOR COOLANT SYSTEM BREAKS IN THE 177 FUEL ASSEMBLY PLATE MAY 16, 1979 VOLUME 3 This volume supplements the information in Volumes 1 and 2 to more completely reflect the unique characteristics of the Davis Besse plant.
Those portions of Volumes 1 and 2 which are truly generic and therefore already cover the Davis Besse plant are not repeated in Volume 3.
Those portions are so noted in the Volume 3 Table of Contents. Where changes have been made specifically for Davis Besse, those sections or apoendices have been either replaced or supolemented and are also so noted in the Table of Contents.
Revision 1 Revised May 19,1979
s
(
VOLUME 3 TABLE OF CONTENTS
1.0 INTRODUCTION
Supplementary Infomation, 2.0
SUMMARY
AND CONCLUSIONS Yolume 1 is generic.
3.0 TMI-2 INCIDENT BENCMARKS FOR CADDS AND CRAFT Volume 1 is generic.
4.0 LOSS OF FEEDWATER SAFETY EVALUATION
- New section added.
5.0 THE SMALL BREAX PHENOMENA - DESCRIPTION OF PLANT BEHAVIO New section added.
h 6.0 SMALL BREAX ANALYSIS New section added.
Accendices 1
Natural Circulation in Operating B&W Plants Volume 2 is generic.
2 Steam Generator Tube Thermal Stress Evaluation Volume 2 is generic.
3 Restart of RC Pumps in a 50% Voided System Voluwe 2 is generic.
4 Operating Guidelines for Small Breaks. Part II New section added.
5 Michelsen.Recort Assessment Volume 2 is generic.
4
s
\\
The infonnation of Transient Behavior contained in Sec Fuel Assembly Plant," dand Small of the However, the foll ated May 7,1979 Reactor Coo "Evalua tion
- report, stem Brea~ks i
~
important disti owing additional
, is h ng features information is prov 7
1.1 nguishi esse.
of ded the Davis Besse to discuss The plant.
raised loop arrange shown on Figure 1-1, whilment for the Davis B Figur e 1-2.
two plant While the RCS vole the lowered loop a e plant is ess umes rrangem Anventory in the lotypes, to.. Davis B approximately the are ent is shown esse on Se Davis Bess above the arrangement ops has the same for the elevation majority of
<mcovery for e arrangement assures of the the a given reactor not provided.
size break a longer time to core.
in the
- 1. 2 hThe Davis Besse plan event emergency coola core d* on, nt is rpr essure injection pu where the In the makeup pumps compared makeup pumps mps as and also to the two high event an ESFAS signal as high pr other 177 FA plants serve
@ umps are is receiv d, the hiessure injection pum actuated.
of 1 tTpsig.
For Davis Besse e
ps.
gh pressure injecti
, For
, HPI is the ECCS analyse s for Davis Bessecomparison, the fl actuated at an ROS on are shown in Figure 1-3 ow
_ characteristics prisiure and HFI oumoare The the B&W ass pressure injectishown in Figure 1-tactual characte s
As cs for lower tnan 1400 on pump provides hig seen, the Davis Bess can be a Davis Besse psig.
her flow Ao increase the tot rates for pressur e nigh can be lined up te pdelivered *ead of al es rovide suction to th! HPI the pumps, the LPI pumps.
pumes Tnts line-u:
6 (t;ypically called piggy-back) would increase the shutoff head from 1630 C
psig to 1830 psig.
The makeup pumps which can also be powered from on-site diesel power could be used if required to supply makeup flow at pressures higher than the shutoff head of the HPI pumps.
The Davis-Besse makeup ptbp flow characteristic curve is shown on Figure 1-5.
Also shown are the makeup pump /HPI pump head characteristics for the lowered loop plants.
1.3 Aux 111arv Feedvater The auxiliary feedwater system is a safety grade system consisting of huo turbine driven pumps each of which is rated at 1050 gym at 1050 psig (including 250 spm recirculation flow). The pumps are lined up so that one pump supplies only one generator. This system is used exclusively for accident and transient mitigat,1on and not for normal unit startup and shutdown.
On the Davis-Besse plant, with the raised loop arrangement, the level in the steam generator is maintained at 120 inches of 550*F water (96 inches indicated)on the startup range instrumentation if there is SFAS incident level 2 initiation; otherwise, the steam generator level is maintained at 35 inches of 550 77ater"6ci~chT,st[alirup range iTschiEEiiation! ~
1 l
s 4l
~~~
FRESSURIllR %{;
h~,
4 i
r i
I*
l I!
e i-t
'I l
I:
II i
iI 75'-lhg 6-S Il l
i i
1, i.
(
t s2'-s,g I
Q*
,V
~
c.~ -M:.9:~ o l
l 999 5
3 l
!I 1
s'.e-1 1 l, /
4 i
e i
3'd*
-l>
l I
I g
'- - +.
~d
/
9
.) M+;
1#
M
- / &
v k
l I
n'.e-t i
l 4
t, wws
, REACTOR CDOLANT SYSTEM ARRANGDiENT-ELEVATION (RAISED LOOP)
~.--
. Figure L-1 I
t
--e w
.,n.--m, e
- 3.
t
{
STELE I
i EEMERATOR
- L STEAR
(
y j,\\
n i
GENERATOR i
/
i r,
ei i
.1 i
i e
8 f.' 0 ' _
i i
-i i
i f
i REACTOR I',
J-8 g'
79 -t B -0" y
N tEL:TcR
(
}
I e
1
,i CUDLANT i
5' 5
o 3
I l
i 42'- 88 PusP e
i
[ /
.. H =i i
w'
- al u - yv.
s 2
i s
y,-
N
=
-u.
7 2, -c-p.,g.
i 3,
i 1
29 *- 3 y '
i l
g i
s t
i
/
I
~
j s'-ca s
9' l
l o
ss 3 s s s s ys s
REAETDR DDDLANT SYSTEM ARRANGEMENT-ELEVATION (LOWERED LODP)
Figure 1-2 m
e e.6 em m em e
-e
e e
6 e
e e
__g M
2 4
w J
Y Cl1.
8 C.
o a 0 a
Lea e o m 5 *"*
L.
G
-e A
e>
.m T
Mw q
h ad E I
dg T
6 W
,g kN N >
L-
$g LJ
.a a
l g;
Le C.
6"
= _g l
2 3
p M
m
"~
m p
en __
'N X
p e
^
8.
=
=
a-Z 2
er
=
2
-.]
m.e t
d..
i g.
2:
r si L3
.g 6
c --
g7 s
7_
g n:
- - = = - ---- -
... = _
s l
/
=.. -
l
~
4 i
l i
r twee 4 Mith Pfeid.re 7h s een3N""Pu::;o n;_
s i
f
,! Iotal__Flou'JIncludine _Racirculaden) i i
{
cola I;et'r~ersute
. ~ -
6 i
1 1600 w
i x
v I
i g
x X
1
\\
j x
(
i x.
j 4
i
\\
6 1400 i\\
4 N
N ao
\\
5
\\
.i i.
o t
e i
\\
v x
\\
e x
i x
4 1200
~
\\
t
\\
r i
6
\\
i 00
\\ 1
\\t
}
s
\\
.k i
\\
I i
s 1000 w
i 1
i s
i i
i s
w
\\
i 1
a r
\\
I T
i 1
L 800 N
.a s
.x i
r 1
a s
y 1
.n w,.
\\,
,z,
.o; 600 i
.g i
s i
1 s
r o
1
,3 p
.s i
e f
, 1 i
400
, i
- \\
l r
\\
i
{
's
\\--
200 t'
\\
'\\'
i 1
i
\\
\\
\\
0 200 400 500 800 ko00 Pump #1 Tiow (G7M) a
,,,c.,,,,,,, - -. - -,, -,.. - -, -,. - - -
-~,,,c,-
s e
i 1
I
^
t 3
e e
6 g'
i 3
J l
I e
r V
t 6
mg I
6 g
c
(
i a A j
i _.
w e
e w
QA -
n e
A 6
i
/
e e
i 6
i
- /
'N t
W on i
i e
1
[
n a
Y i
a /
4-
~
e t
I w*
N 8
/ !
N
/
%1.
f
/
f
+
r*
4
/
/
m;.
4 g
i j 4
1-t K.
i s
6
/
v s s
M
/
6 6 -
O A
/
mJ F i
g
/
1 9'n.
M f
f
- /
I. ", n
..~
C N
6 6
/
f 6
t
'.)
w q
i w
J
/
p 6
n v
N r
d" i
L t
/
M 6
i e
pg/
.w ame
/..
e e
=4 i
i
/
(
r 6 f x
uN
/7 %
i p-Vs i
e d
e
/ +
_r.
t
\\
\\
f J.6 b
/
e '?
.ls
\\
e
%,f '
e f
%w
' 1 is P
\\f t
/
\\
/
_e
/.
i e
\\
r
/
e.
_/,
t
\\
/ /
l e
e dl _
/
Us
/-
'ss I
'r
/
t e
/
=
2 J
J
_f
?
w I'
f e
/
r
/ '
Ai
./
st
,w
- m f
/
g gi l
f si i
y
'I e
f (ib e
i W
/
f v'
e
=
1 n
e e
s I
/
_%~
i o
/-
r et i
J u
a u
jy I
G 23
=
t 7
9-g 6
/
6 s
zw9 g
___f g
, e
/
i U M.e t
6 as t a i e i
a s:
kEk
' f r
O y-yg i
b x,
+
L
- 2:
2 m =.
4 m
C.C LM C
6 L. a s
C a
( w =m=
+
W
_,C
'i C.
U. @
O L
3 @a 2
G=
e l
m -sc 6 3.:
=-t ec b-l T. =* C O
M Y
eqy w
e I
4 J
1 g
a e
=-
V==~-
em e
w g
=
N
'3 3
s s-e ae
- W
_t e
no 9
t t
O 2.0 Suman and Conclusions The inf ormation in Section 2 of Volume 1 of " Evaluation of Transient C
Behavior and Small Reacter Coolant System Breaks in the 177 Fuel Assembly Plant," dated May 7,1979, is generie in nature and is therefer's applicable r
to Davis lessa.
~
i a
I d
e e
o j
a k
e S
...a.
w--
.,cy,,
,,..c.,
4,mr-
,,7,.-.
3.0 T.-2 Incident Benchmarks for CADDS and CRAT*
The inf ormation in Section 3 of volume 1 of " Evaluation of hansient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Asse=bly Plant," dated May 7,1979, is ganaric in nature and is therefore' applicable
~
to Davis Besse.
l e
9 4
i
\\
4.0 LOSS OF FEE 0 WATER SAFETY EVALUATION Note: The information below in Section. 4.0 extends the information contained in Volume 1.
To aid the reader, those arelts which have been revised or added specifically for Davis Besse are noted in the right hand margin.
I 4.1 Resoonse of the B&W Svstem The B&W NSS is designed to accomodate certain secondary system upsets, such as turbine trip and loss of one feedwater pump, without a reactor trip. Therefore, directly acting protective functions to trip the reactor en loss of'feedwater and/or turbine trip were not provided.
Instead, the reactor is protected from overpressurization during loss of feedwater events by a reactor trip which functions on high primary system pressure. As originally designed, this reactor trip was set at 2355 psig (relative to amanimalwating br,ssure of 2155 psig). On a loss of main feedwater, reduction of secondary side cooling causes a primary side pressure and temperature increase which results in a rise to the reactor trip setpoint in approximately 8 seconds, whereupon the reactor protection system pr:rnptly shuts down the reactor terminating the initial pressure ris e.
In addition, the design makes use of a sma119ilot operated relief valve 4PORV) at the top of the prtssurizer. This valve, in the original design configuration, was set to operate at 2255 psig, 100 psig below tne reactor trip setpoint. Thus, this valve actuated on each overtressure transient which results in a high pressure trip anc soecifically it actuated for each less of feecwa e-rea c:ct even. in the S&W clar.t.
.I
6 Following the assessment of the TMI-2 incident of March 28, 1979, changes to reactor trip and PORY relief setpoints were maJe to eliminate t'he likelihood of PORV actuation in loss of feedwater events and other anticipated transients which have been observed in B&W plants.
These changes and their analytical bases are described in Section 4.2 below for the lowered loop plants.
1 An additional analysis was made of the Davis Besse raised loop configuration. The loss of feedwater transient was analyzed for DB-1 and the results confirm the conclusions reached for the realistic lower loop model case.
The FSAR case shows that for a RC high pressure trip setooint t
of 2300 psig, the peak pressurizer reached is less than 2500 psic for both lowered loop and raised plants. The analytical bases 1
for the raised loop analysis are also described in Section 4.2.
s istpoint revisions described in Section 4.2 essentially eliminate the possibility of PORV actuation in loss of feedwater events if the auxiliary feedwater system functions nonna11y. However, if the ifdection of auxiliary feedwater to the steam generators is delayed, post-reactor trip decay heat will once again cause the primary coolant system temperature and pressure to increase toward the setpoint of the pressurizer relief valve. Section 4.3 below contains a parametric study of the effect of dalay in auxiliary feecwater i
initiation following loss of feedwater events.
An imediate (anticipatory) reactor trip on less of feedwate-or turbine t*1 can provide additional time before auxiliary feeowate-is ecuired fellowing a less cf main feedwater. Section 4.4 be'.o.
presents analyses of the effect cf triccing tne reactor imediactly on loss cf main feedwater.
L
.--.__.._,--m,__~
1 Finally, Section 4.5 presents event tree analyses of reactor coolant system expected behavior following a loss of feedwater event and assuming various failures of control and safety equipment.
4.2 PORV and Hich Pressure Trio Setooint Study A number of alternatives were considered for providing assurance that the PORY will not be actuated during anticipated transients which have occurred or have a significant probability of occurring on B&W nuclear steam systers. The alternatives include:
1.
Restricting reactor power to a value which would assure no actuated of the PORV. The reactor protection system, design pressure) and PORY setpoints remained at their current values.
2.
Lowering the high-pressure reactor trip setpoint to a value which would assure no actuation of the PORY. The design pressure of the reactor and the setpoint for PORY actuation remained
, at their current values.
3.
Lowering the high-pressure reactor trip setpoint and adjusting the operating pressure (and temperature) or the reactor to assure no PORY actuation and to provide adequate margin to acconnodate variations in operating pressure. The setpoint for PORY actuation remained at its current value. This alternative would reduce net electrical output.
4 Adjusting the high-pressure trip and the PORV setcoints to assure no PORV actuation for the class of anticipated events of concern.
The oesign pressure of the reactor remained at its current value.
An analysis of the imcact of tnese various alternatives anc
- neir contribution :o assuring nat the P08V will net actuate fer tne class of anticipated transients of concern has been ecmoleted.
Tne results snow that tne comoination of cecreasing the hign pressure 3
reactor trip setpoint and increasing the PORV setcoint provides the required assurance.
A sensitivity study was performed to identify combinations of
~
initial RCS operating pressure, high-pressure trip setpo1nt, and
~
PORV setpoint which will result in reduced probability of PORY actuation following anticipated transients. The anticipated transients of concern are:
1.
enssmf external electrical load.
2.
Kurbine trip.
3.
Aoss of main fee & tater.
4.
Loss of condenser vacumn.
- 5. " Inadvertent closure of rain steam isolation valves (P.SIV).
The loss of external electrical load, loss of condenser vacuum, and inadvertent closure of the MSIV's all give secondary pressure increases equivalent to, or less severe than, a turbine trip.
Therefore, eurbine 1: rip and loss of main feeewater bound these
" anticipated transients.
Anticipated transients which have not occurred at B&W plants (low probability events) are:
(1) low-worth red group withdrawals, and (2) moderator dilution accidents. The moderator dilution accident can result in a high-pressure trip, but FSAR analyses show that peak pressures are well-bounded by loss of main feecwater transients. Some rod group withdrawals can be shown to conservatively result in ceak cressures exceeding the cressurizer sa*ety valve setpoint.
however, tnese events have a very low pecbability cf occurrence wnien assures that the recommenced setpoint enanges are effective for truly anticioated transients, and these events can be excluded from further consiceration.
Y
i q
The objective of reducing the frequency of PORY actuation can be reached by increasing the PORY setpoint, increasing the RCS i
nominal operating pressure, decreasing the high-pressure trip setpoint, or through a combination ef these three adjustm'ents.
1 Since these three adjustments are not all equally desirable, a parametric study was perfonned to allow for a setpoint selection which would be the most' desirable.
For each of the two events analyzed (turbine trip and loss of feedwater), peak RCS pressure following the event was obtained as a function of high-pressum trip setpoints for three setpoints, namely 2255 psig, 2305 psig, and 2355 psig. This parametric study was perfonned for three different initial operating pressure values, 2155, 2105, and 2055 psig. For each initial pressure assumed, a corresponding average temperature was selected which would maintain the initial cont..'. ion DNB margin at the value for which the plant was licensed. All analyses were perfonned assuming no PORY actuation.
i The parametric study described above was perfonned twice, once utilizing a realistic model, then utilizing the FSAR model. The differences t
between the two models utilized are listed in Table 4.2-1.
The parameter values used in the malistic model correspond to the TMI-2 benchmark l
t case described in Section 3.2.
The results of the analysis are presented in Figures 4.2-1 through 4.2-4.
The figures consistently show that, kJ given high-pressure trip setpoint, the peak system j
gressure following the event incmases with decreasing initial RCS pressure values. This effect is due to the longer time "necessary to reach the trip setpoint from a lower initial pressure, thus i
resulting in a & nager pressure overrhoot. 'This result suggests that the operating pressure should not be changed (this is also oesirable f
---m.
, - ~. -..
-r._
-m.-..
-.4-,
from a safoty standpoint bscausa n'o accid:nt annlyacs hava bscn parform:d for RCS pressure lower than nominal plus errors). For the lowered loop plant,
' he combination of a lower high-pressure trip se point (2300 psig) and a higher t
PORV setpoint (2450 psig) vill assure the desired results of no PORV actuation i
following turbine trip and LOFW, as evidenced by the peak pressure curves for expected (realistic) conditions. A typical pressure trace for a 1oss of main
~
feedwater event is shown in Figure 4.2-5.
The second important point derived f rom this study for the lowered loop plants is that for initial RCS pressure of 2155 psig, RC high pressure trip setpoint of 2300 psig, and PORV setpoint of 2450 psig, neither the PORV nor the code safety valves (2500 psig) are challenged under realistic assumptions. This result also holds true for the code safety valves under FSAR conservative assumptions. Thus, the revised setpoints for reactor trip on high RC pressure and the PORV are effective in reducing the frequency of possible safety valva lif t following anti-c1 pated transients.
An analysis has been performed to determine the sensitivity of the peak pressurizer pressure during a loss of feedvater (LOFW) transient to the, high reactor coolant (RC) pressure trip setpoint for a raised loop plant (DB-1).
The analysis is an exact duplicate of that described above except for differences due to raised loop design and plant specific RPS errors that were used in the FSAR case.
Because Davis-Besse Unit 1 is a raised loop plant, the primary coolant i
volme of the hot and cold legs will differ from TMI-2. As a direct result of I
volume differences, the flow transient times (i.e., time for coolant to flow from reactor outlet to inlet of steam generator, I
6
etc.) vill differ.
Th3 prassuro drop between core outlec and pressurizer vill also be different, due to the raised loop configuration. FSAR and realistic cases were considered using the sa=e definitions as established above for the lowered loop analyses. The TSAR case censiders instrumentation errors vhile the realistic case does not. Tor DB-1, the high RC pressure trip setpoint error is 16 psi, while for other 177 fuel asse=bly plants, the error is assumed to be 30 psi.
The results of the DB-1 analysis are shown in Figures 4.2-6 and 4.2-7 for the realistic and TSAR cases respectively. The figures show the relationship between the peak pressurizar pressure and the high RC pressure trip setpoint for a LOW transient.
For comparison, the results from Figures 4.2-1 and 4.2-2 are also shown on Figures 4.2-6 and 4.2-7.
In the analysis for Figure 4.2-7, the pressurizer code safety valves were set at 2500 psig for all cases to show under what conditions the PORVs would open. Figure 4.2-6 shows that using a high RC trip setpoint of 2300 psig, the peak pressurizer pressure vill be approximately 2350 psig.
This provides sufficient margin to the actuation of the pilot operated relief valve (PORV) which opens at 2400 psig. Figure 4.2-7 (FSAR-conservative) shows that the pressurizer safety valves (PSV) vill actuate if the PSVs are set below 2480 psig when the RC high pressure trip setpoint of 2300 psig is used.
If credit is taken for the PORV opening, then the peak pressurizar pressure can be reduced
,,,po 2433,psig for the high RC pressure trip,setpoint.cf 230,0 psig.
I 7
4.3 Parametric Study for A W Delry D:vic-Besco 1 is dtsignsd with a rodundcnt ecfoty grade Auxiliary Feedvater System; therefore, loss of all feedwater is a highly improbable However, in the extremely unlikely case that all feedwater is lost event.
for a period of time, the basic system responses provided in Section 4.3 of Volume 1 of " Evaluation of Transient Behavior and Small Rmactor Coolant System Breaks in the 177 Fuel Assembly Plant", dated May 7, 1979, would be applicable to Davis-Besse.
4.4 Anticinatorv Reactor Trip The information contained in Section 4.4 of Volume 1 of " Evaluation of Transient Behavior and Small Reactor Ccolant Systa= Breaks in the 177 Fuel Assembly Plant", dated May 7, 1979, which demonstrates t.he typical system responses to anticipatory trips, is similar for Davis-Basse.
4.5 Potential Loss of Teodwater Accident Secuences In order to examine the effects of potential failures, including a total auxiliary feedwater delay, following a loss of main feedvater, event trees have been prepared for the cases in which main reactor coolant pumps are available (Figure 4.5-1) and in which reactor coolant pumps are off (Figure 4.5 ~
h aarmal loss of main feedvater event, a reactor scram occurs shortly h fter the loss of feedvater and the safety grade auxiliary feedwater is initiatec
- vithin 40 seconds. The auxiliary feedvater control system controls heat removal and tha reactor system stabilizes in the hot shutdown condition within a few minutes.
This normal path is shown in the left hand side of Figure 4.5-1.
p.tifact of an auxiliary feedwater safety; grade level control systen single
< failure would only affect one steam generator and result in overcooling.
This
' overcooling could lead to contraction of the RCS and a loss of pressuri:er
{
level indication.
If this occurs, the RCS could become depressuri:ed and the high pressure injection system would be initiated 8
cutomatically on datoction of,ths low RCS prassuro condition. Th2 high pressure injection system would add water to the primary system and refill the pressuriser whereupon the operator could take control and restore i
pressurizer pressure control and terminate the transient. This path is shown on the left hand side of Figure 4.5-1.
The path for a starting-system failure of one train of the safety grade auxiliary feedvater syste= is also shown in Figure 4.5-1.
This path also leads to acceptable results.
The right hand side of Figure 4.5-1 is illustrative for the case where the two redundant safety grade auxiliary feedvater trains are arbitrarily considered failed. Normally AFW would always automatically be available to at least one steam generator even under single failure conditions.
In the total auxiliary 4eedvater delay case, af ter about three minutes, the PORY vin open at its setpoint of 2400 psig. If auxiliary feedvater is restored promptly af terwards and the PORV reseats properly, the transient vill be M
(
terminated in the normal way. If the PORV does not close properly, primary systemepressure vill fail rapidly and high pressure injection system vill be actuated automatically.
This again viu lead to a small break LOCA safely within the system design capability as bounded by analyses presented in the next section.
As Figure 4.5-1 iudicates, the pressurizer win fin in about eight minutes of further total auxil'iary feedvater delay, b ry. system inventcry n ss up to about 30 minutes of total auxiliary feedvater delay without high f
spressure injection can be accommodated before core damage is likely to occur.
Analyses presented in the next section of this report cover this case. If eauxiliary feedvater is actuated within 30 minutes, the transient can be ter-l iminated without core damage.
Figure 4.5-2 shows a similar analysis for loss of feedvater event acco=:anied by a loss of power and, therefore, a loss of forced RCS flow.
A nor=al systen response again shown at the left hand side cf the figure provides for auto-matic initiation of auxiliary feedvater within 40 seconds anc control ef 9
i
O a'uxiliary feedvater level to 35 inches of 550 F vater on the startup range. As discussed in a later section of this report, this has been shown by test and attual plant experience to yield excellent core cooling by natural circulation. The reactor system resumes stable natural circulation shortly after the event. As before, a single train f ailure of AFW would lead to acceptable results.
The right hand side of Figure 4.5-2 shows scenarios for significant delay in the initiation of both safety grade auxiliary feedwater trains.
As with the previous figure, these can terminate in a small break LOCA if the pressurizer relief valve is actuated and fails to close during the course of the subse-quent event. As before, this small break is within the capability of the system to handle safely as demonstrated by the analyses presented later in this report. As before, a period of auxiliary feedwater interruption without high pressure injection of up to 30 minutes can be sustained without core damage. Initiation of auxiliary feedvater at or before this time leads to transient termination by design analyses.
4.6 Conclusions The analyses and discussions in Section 4 on this report lead to the following conclusions:
1.
With adjustments which have been made to the setpoints of the pressurizer pilot operated relief valve and the setpoint of the high reactor coolant system pressure trip, PORV actuation is not expected for anticipated transients.
This very significantly reduces the probability that improper operation of the PORV will lead to a loss of coolant event.
2.
Although Davis-Besse has a redundant safety grade AFW system, the effects of the delay of both trains has been examined. Analysis shows that initiation of auxiliary feedvater at 120 seconds (three times the design value for auxiliary feedvater initiation time) does not lead to repressuri-1 zation to the PORV or pressurizer code relief valve setpoints.
Io
\\
)
i au; ciliary feedwater level to 120 inches of 550 F water (96 inches indicated) on the startup range. As discussed in a later section of this report, this has been shown by test and attual plant experience to yield excellent core cooling by natural circulation. The reactor system resumes stable natural
~
circulation shortly after the event. As before, a single train f ailure of A W would lead to acceptable results.
The right hand side of Figure 4.5-2 shows scenarios for significant delay in the initiation of both safety grade auxiliary feedvatar trains. As with the previous figure, these can terminate in a small break LOCA if the pressurizer relief valve is actuated and f ails to close during the course of the subse-quant event. As before, this small break is within the capability of the system to handle safely as demonstrated by the analyses presented later in this report. As before, a period of auxiliary feedvater interruption without high pressure injection of up to 30 minutes can be sustained without core damage.
Initiation of auxiliary feedwater at or before this time leads to transiedt termination by design analyses.
4.6 Conclusions The analyses and discussions in Section 4 on this report lead to the fullowing conclusions:
1.
With adjustments which have been made to the setpoints of the pressurizer pilot operated relief valve and the setpoint of the high reactor coolant system pressure trip, POK7 actuation is not expected for anticipated transients.
This very significantly reduces the probability that i= proper operation of the PORV will lead to a loss of coolant event.
2.
Although Davis-Besse has a redundant safety grade AW syste=, the effects of the delay of both trains has been examined. Analysis shows that initiation of auxiliar-/ feedvater at 120 seconds (three times the design value for auxiliary feedwater initiation time) does not lead to repressuri-ation to the PORV or pressurizer code relief valve setpoints.
[0
I Thus, a significant margin fer manual hackup to th2 sofoty greda auto-matic auxiliary feedvater initiation systens exists before actuation of the POR7 is expected.
If both safety grade trains of auxiliary feed-water are delayed even longer, filling of the pressurizer due to ther=al swell of the reactor coolant water, requires approximately 10 minutes.
3.
Prevision of anticipatory reactor trip on loss of main feedvater provides additional 2argin for auxiliary feedvater actuation before RCS repressuri-zation and pressurizar filling.
4 Finally, a systematic examination of potential failures in mitigating systems following a loss of main feedvater event, shows that in all cases in which auxiliary feedvater is supplied within 30 minutes of the loss of main feedvater event, the transient can be safely terminated without core damage, o
k Il
.ah.,
%!fe're::es 3e:vea: 74411s::.: Medal
- I 75.2J. h.ede
? art:n er 75 AJ. ::ede; laa.lis-1:==de; Medara::: ta=pera:::e :saff,
-0.S =10~-
-1.05 10~6 th/k/*?
Depp* ar coeff, !.k/k/*7
-1.2 a10'3
-1.36 = 10"!
4
?;essurizar spray Off Ope:a as
?:sssurizar he.atars off Operates Eigh to pressure rea:: :
0.65
- 0. t.
- 1p sa:pc1== i=s::':
delay ti=a, s Eigh Er pressure reae:::
30.0 0.0
- 1p se:p:1=: 1ss::'s at:ct, psi 4
O f
I e
i
'\\
I l
l t
~
,,m... -,....,.
-r
- - 4 5 D *,,
,1 i M.
Fn.,
pu..W..c, p
-m Pai
...G
+
as. C.e,.r.
...... /* *,.,I w,.,..
- f....
.. ~ ~.. '. ~ _ '......
6 mI.u..
n.
- a..
- ~
/
.....sO,A m.
....._.....o;;-:........ _. _...
,y............................... -
.. ~................. _
g.....
9
~....
'".?."...
- ;. !: ' *. a ;.
t.**
.. 8.....;..** ? ? ': " "" ** :;.........
- . -....: * ;.; *
- J '.*.. ".?." J ?! ; " * ** *..
~$
- ed.
..... 'I
=
... :..; r.. l...............
q O
g L
4.
J..
.......L I
f- '
'8
- ?
t '
.'...'b
..:s '";4!!!*:
f.
3
.W.
y,.
1..
.. g.. n.f..
g g 4....
-3
............a....**
...M........,
.......... 1......,, g..;.. :........g.
,,...a.,... _.......
.6...
y
.g 3
I.ed
............3................................
.a g
,g,_,a 3 %.
6
- a....
.a i
- 9. E ed
/
- 43 r
.t...
g.... a.. t..
4.
- g. C I..
.J..
s
/-
-e 8
.=. O
- g. U
....J
....... lr.
..l'..'.;
I'***
C.
%.......0..
m......
e
..t
.,....1..
.. m 1.
g eg 3..
.t
.s.
b.......f g
..1....*...._'*.---...*.'."...'".7 3........
s.t
,.g...
.g 12 W
....: ".;; ~."
- l 4
l
==
" ~
M.**.;.......,;
2 g
.e g
s C-t
'I 4
A cc
_+..: 4.Mo mo e
-?
..!.:H IG H. RCS *. !'.CIWF..
v 0
TA P.6T.*. PQ: M, P.'tG a
fw
- u. 2
.....M.... h..........
.;mr,
, g,n., l..
,;n s -;..
..C
.C. q.. g g,
..4...
e P.
.g.
-..e......
... m...s...
- 3"v y.
- e. e e e
.I
.m
......t
.. w..
m..
.a...r r..:. 2.m :.
v sr..
t t
I t
l
m.
.M m.
m gg 4 AP MWh
.l.
m
..... ~.
..... _......6 M..
. =.....
+....
._e.
..-_.w-
_m M
$==
w
...../*-
.g
~
'.:E t
- ' '* * '*IE
..~.. -... "
.'?..'-. .......
....... 3. -.. /. ! V..
....4........
.. =.. -
..f*.
q
.g.
ie..,.
..e............
.a.
N & 8 4 %,.
,.j..
m.ge q
n=
a?
- p r M== w. s.eM. M =,= f\\.,g g*
g
- 9. A" =F *Md?%,
- C
.g at._
pgIG
.s...._... _. a Jt $$.... _.... :.....
t 3
e
@..t.
.L...... _..
.........I g..
.... _... _.....m........_..
....8....
t.:
.W
...... L... _.. ; /. ; ;; ;..:.... ; " '...
F.
..h.
...............t.....f.......f......
.aM.
.....s.
...s. W..
..........a.......
..a he
.. g
- "*- -Tur-.-*
.::n. ::... !.. _;..s....
v.
.i 4.
e.
.......?.
.t
..t........3.
. 2.
......, &&1..
.4
........e.6
.s.......
......'s.-
.g.......E*
...s
~
...,...8..
t....
_......... ~ =-;-.,.
".":;.:.:;.....-.=~;;;.;......---....
- -.. M.~;.*
I
.- a g.
- (
6,
.'...;o
.;&.;*:..1..
1"._.".....,
.. a.
- := ;.,.. ;. M.. * ;. ;..*.* ;'..
2
....... 4. g.
.9_.
- ;;.. - ;; : :.. m..:........
....e.....,
..--- =
............... _:..:... t.
.n..~.
W
........ = = = =..
f................
.8....
I..........,.
.g.e......9~
.g 23:c g..
b.t p.-t..,
y
. ' i.*. ;... ; 6.....:...
- g..
g
.4
..a 4
.t 8.
.....g"..1 g-
.;g:o m__.
.agc l
_..._..t.
=
[
i p:.HMH MS Pn55uRE l
.TU.*. 57PM T, F'3 G e
1 t
=.:
....s..
m.
..... r.
e
...,a...:
.....a....
.y
.c
.a.
- r..
i r..
n.. c........ -,....
0......
, n.:..:
s...
......m.0...e
- 4...
\\........
N.
i..
.......... a...
- v...., s,.r n
Lowered Loop and 2500 psig code l
Safety Valves l
l l
.t
do, "aN @ A*
r PR.e.c 5 A=.. =.
- s..
- i..
. P4M*A. 7G.
3
........~.....
I
,f.m.,..,%
y
- #. a.
__...--.m..
..'.f'...
t.
. w, g.i. - " d7ww
=.n=.=..=;...*=.=;..=.=,.
, p#
....y.-
.. - - ~. ~.....
.f='=.=.. b,sene""..'."".."*'
.m.
==.='b...
..t**.-...r.
. : " * : " :*.::..- '.. " ": m-
...-*T.**-
"'...............~
~
..-.t..
- : 1 a. *.. "...
gr;
.g T
E
...... - O 8')
g
.4 y.
4/)
..:t j - ?
?"-
,. g
.l........
..........l...,....6.
y....
...........1.-
. t..
.i.
....1..........
.....l.
'. "* 1 M.
.....s
.L.:. ;.
.a. : 2... h..
..t......
.. P ;." ;
.a."...J *
.....e
!.*.*.".".1.l..
2.....
.....I..........
...... /
..mm g..
a$..
- {..
I o
t
.W e..
e 6* -
e.
=
- ~ ;.Y.
e-
":"?'" "
..."": 7 T..--
'. I C..
e
. _........t..
3..
..m..
....,,6.
...... '....**...". t ' -... J '.;;...;........."..9*"".
6..
..7,u...,...,..;
.. ea! :..
r.....,.....
r..
- r.
yt
.r.
=.
...4
.8
--M g,.
t.r
.-3.......... :-
g
..L..
r e
...i. : -;....
e*.
,E??
. & &O g
t.
t"
.e g
g 4
4
- ...e i P.
.d.' ".%, A.t 8
.t.
Ht.
1 p=.
t.
u.?..
.am.
.m==...
~ -....
...n..
%y r a r..
.,....,.....e.
g....
...e....
p.
. P....s 7
..g
.M.
p
\\
.. =.
f.
s,........sJ........s g.W.g.
- w-
.l.
... -.. m.
.d..1%
4 g.
4 b....
..7....-... %8. g.
J. p..
.p..
".. U,.,3
.,0 P. 4. 3.,. M.
.. s... r%
..s
.s..
Lowered Loop i
e e
- me i e.
.s=
3
(
3
,,.......t
.......m..--
r i
1 o
am..
..g Q
1
. =
er
%$h. [:N.,. N.
- '*=.==---
.mm '.'""'t
.,...3..
.,... e,--
.....,......s..
.........m.m..-.
.........m......
g..
i.
F*
- 8..........
e
..e.s... :. V/.."..
e.....
.n g
..F
.. "O" am.
. * =..
.mu...-
m.
..73...
....m...-.
~
4U.. -,.
.. ~
a 4;...........
.1
.I.
+
~-
e
..e...
e a
....m._........
.i..
.g
-*?:"=***.";."":";*".";*"***"""*'"""...m...
m.-
.e.
-?***
- ..:""***.****.*?"T** ** -.-
- "."?
......".."*"y...
4
- 8... W...
.......';.P.;..........i.;.
.a....
.a.................
............I..
-.e...
f....
.. 4
..4
......l
........I.
..1..........................
1
..mm..-..
....um.
.........f.
I
..' gg r
..A
. I...
s I.
- 1. ". ".; P.
- * - * ~. ".
....... *...+-- -.. -"
T==
- . ;;2"
?"* : " "*. ;.*;..........:"...M".*... - =..=f l
-- "** - -**** t
...=r
......g,
. **.................. =.......
e.".".'
..... t.........
.........9.......,........
...........t
...==-
...=. =-
P"..
e t
.n.
- 3...
"r.
E.,.
.3..........
...................t......~.........................
I
,f i
7 =..
?
.e....
,......_a, y
......3 m... _....
,... M
.o.
q r.-
t -
r r.;_1 y y g.g.
v.
2
.S.e.c. w.
g
~.
t
- e..
3...
.s i..
.m%,
... u. ;;;. p.-
. 3..
J... ;; ~ &%GM. AfS T.nTd. *
.a.
. - y.s. i.~m. p5y
~
.s
.. ~....... _......
p m
. ;. a :..
2
- u..
s,/
..,..c.
(
..i.....
..- 5,... ")........,...
m: x.,.. :.m e -
- i
.-.m c......
... a
.2
.,. v.
ms..
s ;-u.
.. :..s.
Lowered Loop and 2500 psig code Saf etv. Valves c
.-o
. +, - -
..,m-
0 9
e e
e 95-6 I
a t
a 4
=
g W
+
2 e
Q w
s
\\
i yr u E 8
w s
E g
)
2 I
g :F A
g
% {W s' w.-
g 96o
> > <2 w
F s
.- w N.c t R
\\
,$Em N &
5N o 3 rs*
W2
[e 4
M [e *I".
{. E y
\\
-b.
w s
W F
r w
v
\\
y' 0**
s C.' e M
'P 38 o
u to e
s'g, "u $0"-
f2
.c
==A.
- g. l.
g l
M
\\
w I
s
'J.Q I
Rw I
1
>4 ad O
)
s o
Z s.
g h-t a"2 R*
i W
s OFo wrw 3
,(
s h
kjh p
a us 2
w w
2 g
g g-a i
6n2 E
s
.J o
s ' c' v s
b 4
LD <
Cd \\1 j
m 3c
-n W
1 w.4 e
/
e
~a N., -
N
)
e d
y j
o 4
g N
N b
uw-N N
N E
m e
i
1 j
p
,. a us.
_,, =
s ri
-M..
N w en.
+
am
,s. p W
m-.__
m f
ws r,4 r rrf x c n t5 m-- _-.-- --
o'e
-- ~ w=nrs s F)h.__ _
m um -
.=n -.
MCh)h.
~[
r M.-
3 hh::.:
g$
/
m..
- ".'.;. ~.-
h -
- 5 h
.yqw
=-
4 t
t_ _ _
_4,::
_,,_ :~-
CtM':': l- -
_/.
~M v.,.-
fs
~
\\
g gg "w
-~.-
-C 2
.:.p-
'/ :-
g S
- ~g--
I I_I
~[,
- : :- ::=
_4 y
~
., - ~
C
,,7 -s A~-
e= q N L-
_( G.
_ pf,.. _ p
--4*-;
Et?
Q D 7 E-, m wi[p'd$p_ y{, _
E "Nd
.w
.~
ame
,"*[,}g{,. 2 g_.,;b '--
- ~~
7-
- =. "
~
Y Vdie M'D 77%**Mh7lf
] ( [. fT*:."-
~- -
g.
- = = = = - - - - = = -
9 --f72 E - v a W M a T 6)~g); _._ :,,;
C
=
j s- - -
WWNh %.- r 152?*'* (**& [ ? L,' C _ __ _ _-.
~ - ~' : " ;.:;. ;, : -
4 M ICT :" ij""*. EO( ggf /ffg
-; ? ;...--....
r"--;-..:..;-.-----~---~;==-.,.;.,,-;;.,,_._
' ' ~" CC k.t n l Ts b*J $ A ny n ; 21a : g r,. g _. _..:__._..
~
- t 6iWK1 -3 p s.- t ty p e i -3 =.:==-=- - = =
_ _ _ _ _ _ _' * ~ ' ' '
~
- -. - - - - -..........:. :;.;=,,, -
~~
~ ~ ~ ~ - ~ -
' C I
- 2: ; "'-- -'~ '
^ enom L -
- -- ~_ ! - -* ~ ;2 3 2 ;,=,; -- ; ; ~ ~ - -
e em
)
l l
\\
i
e
- e.._
e
' t=~:: = r =-.~
- _..-...=_.
--- - --- -_:.: = =_;_ =
s r-:
c
= -. _
-- g
==. : :...
3
--._3
.... _ =..
om.;.-
-. r--. _. -=-_._
""--'--"~:""=;.__....,. _:-
..r.-.
J
~
G ENEA /C.
(.
= _ _
- = = r--r-.
Q_-
...a.-
-*,,,)_.----
+- -- / w-*"~.N :- __- - - - - - -
v
~
) -- --==S
, _.= p--
~ - -
g
.,t
._mr
- ; ' 7.E.i21:i:. ---
E..
__.-s..._..,
. -.. ' _ ~....
........ _ __. _ _.. r y =F~ - - f'-
_P.
_ _. _-~~;
.$--_._...... ~
p.
, :./ *:M.,. _ _
z-
~
==- - Nw= =:-
-. - - - = =.
- o m m =.__ _. - w. f,.:=f::iiO/'. ---
c.-
.==
a.
+
_ =. =
e
.,_--,2,.
- - =. VMLSup..=-t r vr._~~-
.j;a m
. J.
~
6._.
g p, :=:-..-.-_
-y=
.a
_.e
.....r.ogr w no -
_.. i
_. -.as r.1 w
~
==1
. u--
.s.s.
.. :n%,,
_-~.-
-~
.p-A..
==-
=14;.-.45 g ~..
_.y.___,s._
===..-.:=::-=--.~...
~
s._
.. %.. _._ d.l.. _... _ _-~~
r=.--:____.
_.. _. :. W ---
r-v__-
s a
._:==.
g__; - - = -
.. --ti=
~
e.-
~Q
~.=
- -==_.. :.w/ :.: - :.:.n==u. :. :-- -...:
- -~. -... ::=.=
.:=..
=- n
- A..:-
.- G -- : -... :.... : ::. ". -
- 6 :::=..
=b i
I
=.......-.:.
. r..: =. =.:. --- -... _..
.47 n--
.i
\\f_ -
.d
- ,__ g.
- _... ~ :,,... ;n
- n..=:;..-----...',....
f-=
.__.7
..m-
,w1._.
c=. r,% - =.-- :_
. '/ -
- . a =.".::.=--
=
=--
'~ t ::=.-
==r--"
- ut -
s m
-....-. ::.-: -- -=::2. ru=.= *. - ;;;.
2
.... _ =
i
=.:
- =.=- g.,g
.... :.-p
_.-.-y
- 2; :.. ;
. : =: +
- _=_;;;,,.. f.,.;..
- .= =...,.,. :...... --. ;=. ;.. :
_g,__.
g 5
t
== _-
- . _=.:.. =.. : -.:=. _.. _.=/ ~.. : -
e
. = - -.
M.p,p Q.
t g n -. :::=. ::= _ = =:= : -. :.=:. -
.,.... :: =..-
l 3 -
+
l
- i. : -
E
..--l...-
- 21 4" C :-- f..S". e --- _ :
-....g_....-.
= c_,
- r t t s,w nL
, _.. _ ;;. w.. ) M g
- g
.-w. =_: :. -.=.-.::... u _ :.==-- -
.._$ e er L..
4 n. ])
3 p_.. :::.-..::..
i g
. a::, :.
a r..
.._ e _.;;. ; -. :
_...a. a 2=
,t."-'-._-. _... _.
---**-*---,;--:J'*=*r.._._a----':--**--""--
g
. - _.... :* - 9 ' :
pt gj= 2-h --D-.=
- - 9" ;:i.", - =-- ~=.E- ':::.G i - H : V!P n:' M : -
-3
- - ";"- :T, - -
t.
Ce'
- ^ ~
W
~..;d -
.. ~.....* _.. _.
g,_,;
._..-...'_'~--.'._..=~-WO. ' '- - - -
Y*
. =..
p_..
---=__..____,_~~.-.$W
.y.--.._..'
i n
1
_:.-___.2-c l
J E.:
... : :=:::... -
- -.-- =: =.:.._.-.; w,,,_gQfggssu ggy p_S. gygn (s_psg.....=.:._ =.:. :.-.=:..==
-a p
-r.
= :.:...
.=_.=:--....- -.-.-.
=.
~ --
- -.....v
~*
=:.=...
- : =. - ; =: = u -. : :. - - - : =.:_. :..=: u. - =: :- v.. _-_.....--...:._= =u:_:.._;u..=.----._._-.=.
= = =...--
=
- -- - - m - :.:.._. u. - _ :
l
-r
- _m
-._.=.Fjgy g gf. 4. 2 - - -
=
- -.-.=:.=__.. - -... _ - -
.2..
- -- _: _. :. =..:. _=-. pqy. pqSsumx_,p_g553u p.: g _ _.. __.
==
.=e - =
i.
_.... -.... =... -.. _ _. _ -.. ACS.PR E* 5uRE TR t7.SCP::ii
- _=
=.
A LW _M C. Q.d Q Q,,,,,'j gaf Q 3l.
. fc4 FS A A C - W. T:.C N
.W.
t I
'I C PC Mi.h% L.CCN.0 l'.F CE
~.
-.. _. - - -.....'.2-psig code sa.f ety valves --
l
)
.... m.c
- . J *.J = ': =-* !
.._. :...._ -iiU :.ER:-iE= u = = ; H a ::.:: =_ :..-i.e m
e l
l l
p Iligli RC Pres' ure RT s
(2300 psig)
~
1 F Rnth trains AW Delayed 1f A N "on" A W "on"'
h PORV "open" 6 3 min)
Two Trains One Train (2400 psig)
(40 Sec)
(40 Sec) y V
Pressurizer Full N 8 min) 1 r SC Level Overcooling SC I.evel
(
r l'
Control A W "on" 30 min Delay (M in)
(SG leyel 35 in) 3 r
, r 1 f A m "on" 0:erator Manual Cont rol of AW i f 1 r 1 '
T I
1 IIPI Actuation.
PORV closes PORV "open" PO4V PORV Pzr Empties Falla Closes SC I.evel (120 in)
Open Pro erly IIPI Actuation IIPI Actuation (SG Level 120 in)
(SG Level 120 in) 1r 3 r Terminate IIPI i
IIPI actuated Manual Control A N
}
(SG Level 120 in) g Small p j l.0CA Terminate flPI N 50$-)
Control A N v
I Transient Terminated j'
Figure 4.5-1 Event Tree Scenario for a I.0W Transient with RC Ptimps Availalile (Davis-Besse 1)
^
liigh RC F,vceura l' (2300 psig) l Both Trains AFW Delayed y
1 r AlW "one" AFW "on" Two Trains One Train i
PORV "open" (+3 min)
(40 see)
(40 aee),
,2400'psig
~
Pressurizer Full (p 10 min) 3 py SC Level e
Overcooling Control 30 min
. (35 in)
Delay SC Level AFV control (35 in)
E# Y 3 r p
AFW "on" (SC Leve' 35 in.),
p PORY Works Operator Manual I.
Controi of AFW PORY 1 r P0 (V PORY fails closes Lower Probability Op n Open properly of HPI Actuation
.,,hg-Pzr Fnpties HI.
SC Level (120 in)
HPI Actuation Actuation Actuation (SG Level 120 in)'
(SO Level 120 in) l (56 level 12OIN
.l
~ if Small RCS refill and Terminate HPI Break 4---
Steam Natural Control AFW LOCA j
Circulatlon 4
RCS nepressgrization I
i g
No RC Pump Start or Bump RC pump y
Natu ral
(
Circulation Terminate llPI (~50'F)
Control AIN PSV *open" l
(2435 psig)
Trailstent Terminated i
Figure 4.5-2 Event Tree Scenario for LOFW Transients Nat.u ral Forced e
i with I,oss of RC Pumps (Davis-Besse 1)
Circulation Flow t
5., 0 The S=all Break Pnancmena - Descrietion of Plant Behavio q
This - s e c tion -va s-revgi tt en-f o r -Davis-B e s se-Uni c 1.
A 1:ss-ef-cocia: 4::ide:: is a e di:10: in vr.i:h 11:uid 1:ve::: y is 1:s:
f::= the rea:::: :ccia : sys:e=.
Cue :: the loss cf = ass f := the rea: :: :::;an-systa=, :he :s: sig=ifi:as: sher:-te= sv=p:= :f a los s-of-ce:La=: a::iden: is a: ::::::en e redu::icn 1: -he rea:::: = cola:: sys:e= ;; esse e.
~he rea::::
p ::e::i:: systa= is designed :: :-ip the rea ::: c icv presse e.
nas should occur before the reae e coelas: systa= reaches sa: uratic: ::=di:10:s. ne existe::e of sa:ura:ed condi:10:s withi: the reac::: systa= is :he principal le:ger-: := 1 dica:1c of a LOCA and requires special ec sidera:ie: in :he d evelopne:: ef opera:1:s precedures.
yellev1=g a reae:c: ::1p, 1: is ce:essary to re=cve decay haa: f := :he rea:::: :::e :o p;even-da= age. However, so long as the reac::: c=re is kept covered with eceli=g va:er, core damage vill be avoided.
~he ICOS systa=s are f
desig=ed :: respend aute=a:ically to lov reactor coola : pressure ec:di:1 cts and taka the initial a::icts to prote:: the reae::: core.
- hey are sized to p;cvida sufficie:: vate :o keep the reae::: core cevered even vi:h a single fai.lce 1:
the ICOS systa=s.
Subseque=t operator acticus are required ul:1=ately := plate
- he plan: i= a 1:=g-ter= cooling : ode, ne overall obje::ive of the aut==ati:
a=arge::y cc:a cocling sys:e=s and the follevup cperate: 1::10:s is :: kee: the rea:::: :ere eeel.
A detailed discussic: cf s=all break phe==:e:alogy is ;;esas:ed is -his se::1 :. nis dis:uss1== represet:s ?ar cf :he opera:ing ;; edure puidelines f:: the devel:p=e:: cf detailed opera:1:3 precedres.
Par: !! prese::s :he ::re de: ailed step-by-s:e; guideli es and is 1::1uded in Appe:dir..
~he res=::se :f the ;;1=a y systa= :: a s=all break
"_'.1 dif f er ; ea:1-depe :in; :: se break si:e. i:s '.::a:1:: in : e sys:::. :per::1 : :f :he rea: ::
- Lat: ; ::s. :he c.ber :f I; 3 :: airs fu= :1_1Eg, a:f :he availabili: :f se::_da-s:.de :::li:g.
Figures ; a:d ~ sa:v ;;::s f ?.05 ;; essure and
/
^
p;sssuri:a level histories f:: various :::bina:10:s of parace:ers, indi:sti:g the vide ra:ge of behavi : -hi:h is possib
- e.
3.1 S =a' '. Er ea k s vi: 5 Au.xi' ia--- F e e dva t e:
3 ere are four basi: : lasses of break respense fe; s=all breaks vi:h au=1112:7 feedva:er; these are:
1.
LOCA large attugh depressurize :he raat:c =ccian: systen.
2.
LOCA which stabilizes a: app;cx1=ately see::dary side pressure.
3.
LOCA which may repressurize is a sa:ura:ed condi:1c {lo~? M u=ps {,
4 Small I.0CA which s:abilizes at a primary sys:a: pressure gras:a tha:
sa endary systen pressure.
ne sys:e= ::ansian:s fer these breaks are depicted in yig=e 1.
3.1.1 LOCA I.aree I:=uth te Deeressurize Reseter Coelas: Ses:e:
Curves 1 and 2 of yig=e 1 show the respense of ROS pressure :o breaks tha: are large enough in combisa:ic: vi:h the ICOS te depressurize the systes := a stable low pressure. IOOS injet:fo: easily exceeds core boil-off and assures core cooling. Curvas 1 and 2 of yig= e 2 shov the pressurize: le"el transia=:. Rapidly falling pressure causes the he: legs to satura:e quickly. Cold leg te:pera:= e raaches sa:u-atic: somewha: la:a as RC pumps ecas: devs or the RCS depressuri:es belov the se:: dary side sa: uratic: pressure.
Sir.ca these breaks are capable of depress =izing the RCS 1.-ithou aid of the staa: resera:c s, they are esse::ially unaff e::ed by :he ava11abili:y of amary feedut ter.
Operatie ef the RC p=:s aise plays li::la cle 1: the': curse cf events. Otha -han verifying : hat a l'.
ISTAS a::1.s have bes: ::=;1e ed. *he epara::: needs :
ske== a:-icts :e bri:g
)
I
- he svs:e :: a safe stab *e ::=d1:10. Rapid dep;essurizatie ef the s:aa:
1 ge.e:a:::s veu*.i :.17 a:: :: a::alerate ROI depressuri:a:1::. :: is. hevever. :::
l l
=a:essary. 5 ar:=g :he ?.0 p=:s :::e they are 1:s: is ::: desirab.e f:: :.:.s j
i i
l
- lass f break.
- ::g-:ar: :::'.1:; vil'. re:cire the :pera::: :: shif: he L?~ ;=: su:: :-
- :te rea:::: ::::a:.:=e:: buticia; s==:.
1 E
t 1
i
4 5.1.2 LOCA Which Secbilizes et Approximately Steendarv Sido Prassura Carve 3 of Figure 1 shows the pressure transient for a break which is too small in combination with the operating EPI to depressurize the RCS. The stean generators are therefore necessary to rs:nove a portion of core decay heat.
If the reactor coolant pumps are not operating, and the pressure has stabilized near thesecondarysidepressure,RCSpressuremayeventuallybeginfab.ingasthe decay heat level decreases. If the RC pumps are operating, pressure may or may not decrease.
System pressure could ultimately increase to some stable level as the EPI refills and repressurizes the RCS. The assumption for this case is that secondary cooling is maintained. Curve 3 of Figure 2 shows pressurizer level behavior. ' Curve 6 of Figure 2 shows refilling by the EPI. The hot leg te=perature quickly equalizes to the saturated temperature of the secondary side and controls primary system pressure at saturation. The cold leg temperature remains slightly subcooled.
If the HPI refills and repressurizes the RCS, the hot legs can become subcooled. The insnediate operator action is to verify ESFAS functions, the steam
(.
' generator level has automatically gone to 120 inches of 550 F vater, and, if RC pumps have not failed, leave only one RC pump running in each loop.
Follow-up action if RC ptsaps have been lost is check for natural circulation.
This is dene by gradually depressurizing the steam generators. If this test is failed, intermittent bumping of an RC ptsap should be performed as soon as one is available. Continued depressurization of the steam generators with either forced or natural circulation leads to cooling and depressurization of the RCS. The operator's goal is to depressurize the RCS to a pressure that enables the ECCS to exceed core boil-off, possibly refill the RCS, and to ultimately establish long-ter= cooling.
3 3
\\
5.1.3 LOCA Which May Reprecsuriza in a Saturated
(
Condition (No RC Pumes)
Cut,re 4 of Tigure i shows the behavior of a small break in which the RC pu=ps have been lost and the break is' too small in combination with the EP'.
to depressurize the primary system. Although staan generator feedvater is available, the loss of primary system coolant and the resultant RCS voiding eventually lead to interruption of natural circulation.
his is followed by i
gradual repressurization of the primary system.
Once enough inventory has been lost from the primary system to allow direct steam condensation in the region of the staan generators contacting secondary side coolant, the primary system I
is forced to depressurize to the saturation pressure of the secondary side f
Since the cooling capabilities of the secondary side are needed to continue to remove decay heat, RCS pressure vill not fall below that on the i
i secondary side. EPI flow is sufficient to replace the inventory lost to boiling in' the core, and condensation in the steam generators removes de heat energy.
he RCS is ir a stable thermal condition and it vill remain there uctil the operator takes further action.
De pressurizar level response is characterized by Curve 3 of Figure 2 during the depressurization
, and Curve 4 of figure 2 Juring the temporary repressurization phase. During this tran-
- sient, hot leg temperature vill rapidly approsch saturation with the initial system depressurization, and it vill remain saturated during the whole transient Cold leg temperature vill approach saturation as circulation is lost
, but may remain slightly subcooled during the repressurization phase of the transient Later RCS depressurization could cause the cold leg temperatures to re ach saturation.
Subsequent refilling of the primary system by the EPI might cause ta=porary interruption of steam condensation i= the steam generato r as the primary side level rises above the secondary side leve:..
If the depressuriza-tion capability of the break sud the EP is insufficient to offset decay heat.
the pr'=ary systen vill once more repressurize.
f S is decreases EP1 flow and increases loss through the break until enough RCS coolant is lost te once i
, more allow diract senas condsssation in th2 sesam gsnsrator.
This cyc1 bahavior vill stop once the EPI and break can balance decay heat or the operator takes some action.
The operator's immediate action is to verify completion of all I fu=ctions. Pollowi=g that, he should ensure that the steam genitator level is at 96 inches on the startup range and check for natural circula: Loc.
If it is positive, he should depressurize the stea= generators, cool and depressurize the primary system, and attempt to refill it and establish long-ter= cooling.
If the system fails to go into natural circulation, he should open the PORV long enough to bring and hold the RCS near the secondary side pressure.
Occe natural circulation is established or an RC pump can be bu= ped, he vill be able to continue depressurizing the RCS with the sea.a= generators and as-tablish long-term cooling.
5.1. 4 Small LOCA Which Stabilizes at P > P, Curve 5 of Pigure 1 shws the behavior of the RCS pressure to e break in which RC pumps are lost and cooling is accomplished by natural circulatics.
Eigh-pressuri injection is being supplied and exceeds the leak flow before the pressurizar has emptied. The primary system remains subcooled and natural circulation to the staam generator removes core decay heat.
The pressurizer never empties and continues to control primary system pressure. The operator needs to ensure that ISTAS actions have occurred.
There is no need to throttle EPI flow since the EPI system is incapable of injecting coola=t into the primary system when the primary system pressure exceeds 1708,psig]
5.1. 5 Small Breaks in Pressurizer The system pressure transient for a small break in the pressurizer vill behave in a manner sd-*1ar to that previously discussed.
The initial de-pressuriza ion, however, vill be mere rapid as the ini ial inventory loss is entirely steam.
The initial rise in pressurizar. level shown in 71gure 3
{
vill occur due to the pressure reduction in the pressurizar and an insurge of 1
cooAant esto tus prosaurizar from ths RCS. Once tha racetor trips, systa= con-
' tr ction results in a dacreasing laval in the prassurizar. Fleshing vill ultimately occur in the hot leg piping and cause an insurge into the pressurizer.
'Ihis ultimately fills the pressurizar.
For the remainder of the ::ansient, the i
pressurizer vill remain full. Toward the late stages of the transient, the pressurizar may contain a two-phase r.1xture and the indicated level vill show that the pressurizer is only partially full. Except for closing the PORV block valve, operator actions and system response are the.Ame for these breaks as for similar breaks in the loops.
5.2 Transients with Initial Resoonses Similar to e Small Break 4
Several transients give initial alarms similar to small breaks.
These transients will be distinguished by additional alarms and indications or sub-sequent system responses.
Overcooling transie=ts such as steam line breaks, increased feedvater flow, and ste$= generator overfill can cause RCS pressure decreases with low-g pressure reactor trip and ESTAS actuation. But steam line breaks actuate low steam pressure alarms for the affected steam generator, and steam generator overfills result in high steam generator level indications. The overcooling transients vill repressurize the primary system because of EPI actuation, and will return to a subcooled condition during repressurization. The immediate actions for both overcooling and small break transients are the same.
A loss-of-feedvater transient will result in a high reactor system pressure alarm but does not give an ISTAS actuation alarm.
i A loss of integrated control system power transient s. arts with a high RC pressure trip. After the reactor trip, this becomes an overcooling transient and vill give low reactor syste= pressure and possible ISTAS actuation.
Stean generator levels re=ain high and the systa= becomes subcooled during repressuri-za.io..
. _..........numauc pro toction during tho
' carly part of small break transients, tharchy pron. ding adoquate time for cua11 breaks to b2 identified and appropriate action taken to protect the syste=.
5.3 Transients That Might Initiate a LOCA C
~
t There are no anticipated transients that might initiate a LOCA since l
the PORV has been reset to a higher pressure and vill not actuate during 8
i anticipated transients such as less of main feedvater, turbine tr1p, or loss
~
of offsite power.
However, if the PORV should lif t and fail to ressat, there are a nu=ber of indications which differentiate this transient from the anticipated tran-sients identified above. These include:
ESTAS actuation Quench tank pressure / temperature alarms Saturated priu ry system Rising"pressurizar level r
These additional signals vill identify to the operator that in addition to the anticipated transient, a LOCA has occurred. In the unlikely event that small breaks other than a malfunctioning PORV occur after a transient, they can be identf fied by i=1:1a117 decreasing RCS pressure and convergence to saturation conditions in the reactor coolant.
Small break repressurization, if it occurs, vill follow satr. ration conditions.
By remaining aware of whether the reactor coolant re==4n= subcooled or becomes saturated after transients, the operator is able to recognizs when a small break has occurred 7
a
~
- ~ - -. -
...._,,.y
\\
-.Y 1
1
\\
M 0.-m.
g a
g
_2
- g i
f f
t i
t l
6 i
b 4
7 N'
t, N
/..
(
i g
(
s g
j I*%
{
!N W
a i
.i
=
W.
?
- E w
..=
=
w e
=-
l K
l 1 O
.C m-I g
>.J i
l g
e4
=
b#
l w Sk S
M _N =
l 6
l l
M M
I
-A*
?
O g
t ~
.O h.
- )
a C
m
=
=
5 0
f.) ' a. ^...p. t a
M b
n L.-
O MO
.OO.M
.W O
.O O
M
.ggM$
g
=.
9
+
.*.6
..e.6 6
6 eW asm
.h
_= 6 e.*
m.
.e*
. m e.
esu.
.m..
.e m
W G
W=6 mM M
ee
.6
.Ib g g........
m O
e..
.e e
. W.
.mu..e wM
.e y
e.
M
......mm m.
.a*
- . 6 m.
N.
.6 O
I s.
t d
.N...._f
.,. c
-
- I.
i e
n N.
t N.
I N*
.C.
==
M
.d
,,,,,. m.
.N
'... _.,_.- -.. = - -
- l I
i l
t 1
h t
==
2 e
e
EC t
6
==
m ""
=
bW g
M e
L E_-
g.
N.
- N 1
- 2"
=
t 6.
3@
w* d
~
- 4 EA m
L=
M C
b
=
I s
e.
I DD o
I O
R A
.a.
.e
- t-
=
/ r,.-~,
e,
,.. y....
./.
e w
.. to =.
,r 75 f
ennWh W
Mme 4
~
m.
FIGURE 3 G
-s, Ne.
PRE 550RIZ....
.2 5%.LL o.L.E.R LEVEL..VS T.,Ii4E FOS.
aK I.,., arc 25vR4 u.rs.
=
t.;;
a
.M.
%.t w
W W
a e.a.
q W.
- d. h 4.
g.
J l
6_
p.
~~
_. = '
/ y.r.
s I
.o t
i i
1
'~
6.0 SMALL BREAX ANALYSIS Section 6 of Voluee 1 of " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant,"
dated May 7,1979, presents computer analysis results of small, breaks in the RCS which are intended to augment those in the SAR submittals.
Three types of analyses are provided:
1.
Quantification of the maximum delay permissible before auxiliary feedwater must be established.
2.
Evaluation of srall breaks in the pressurizer which support the philosophy by which B&W establishes the small break worst-case location.
s 3.
Evaluation of breaks which support the philosophy by which the small break spectrum break sizes are chosen (specifically breaks which
'~
will undergo a repressurization during *,he transient).
The work performed for "ection 6 was done primarily on lowered-loop plant arrangements. However, these analyses also confirm the validity of References 1 through 3 F. Volixiie I.for the riiTe7-loop; Davisw3 esse cenfigurat:
~
Thus, the conclusions drawn from Section 6 apply equally well to the raised-loop arrangement. The sections below discuss the applicability of Section 6 of Volume 1 to Davis Besse.
6.1 Introduction Section 6.1 of Volume 1 of "EvalukkTo~n~ of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assemoly Plant," dated May 7,1979, is acclicable ko Davis Besse.
6.2 Events Evaluation (a) Quantification ef the Maximum Delav permissible Before Auxilia y 1
Feeewater Must be Established: Davis Besse-1 has a recundan:
safety grace auxiliary feecwater system. With this system, icss
/
\\
of all feedwater is considered to be an extremely unlikely event.
The three analyses described in Section 6.2.1 of
~'
Volume 1 which were performed on the lowered loop pla'nt mooel conservatively character.ize the expected performance of the Davis Besse-1 system in the extremely unlikely event that all feedwater is lost for a period of time. The essential modeling assumptions and analysis method described in Section 6.2.1.2 of Volume 1 art appropriate for Davis Besse with the exception of the auxiliary feedwater system interruption which is as discussed above.
~
Two cases are given for breaks wnich are large enough to continue depressuri:ation and to actuate the HPI system after all secondary cooling is terminated. The essential system behavior predicted by the two analyses given for this category of breaks in paragraph 6.2.1.3.3 and paragraph 6.2.1.3.4 of Volume 1, is typical of Davis Besse as well as for the lowered loop plants. ~ In these two antlyses, system pressure promptly falls below the high pressure injection system actuation setpoint and remains below the shutoff head of the high pressure injection pumps at Davis Besse. Thus, the high pressure injection system at Davis Besse would be effective throughout these transients and would deliver more water per unit time chat is considered in the detailed analysis given in tne rtfErenced sections. Hence, for these twe breaks anc tne class of breaks' tney reoresent, safe system performance is excected fer the Davis Besse plant without core damage and we can cencluce that tne criteria of 10 CFF. 50.46 are satisfied 2
for the Davis-Besso plant without the use of cuxiliary fosdweter for all breaks larger than 0.02 ft, assuming 2 ECCS trains are available.
I The analysis of a 0.01 ft2 break given in paragraph 6.2.1.3.5 1
of Volume 1 also characterizes the expected system behavior for the Davis-Besse plant for a break of that size.
In this analysis, the prompt reactor coolant system depressurization is not sufficient to actuate the high i
pressure injection system. Auxiliary feedvater is asstaned not to be i
available during the first 20 minutes. Thus, during this period, reactor I
core decay heat is being removed by boil-off of primary coolant inventory l
through the pressurizer relief system. Since the high pressure injection 3
system is not functioning, its head-flow characteristics are ut.important to the predicted behavior. The raised loop configuration of the Davis-Besse Plant results in a more favorable condition with respect to available reacter coolant system inventory to cover the core at the end of 20 minutes than for the specific case analyzed in paragraph 6.2.1.3.5 of Volume 1.
_ This is due to the fact that with the raised loop configuration, almost au the inventory outside the reactor vessel is held at a level higher than the reactor vessel nozzles.
Thus, given an equivalent boil-off of reactor coolant system volume, the raised loop Davis-Besse configuration vill have more available water remaining in the reactor coolant system to cover the core than the lowered loop configuration for which the detailed analysis was performed.
This is discussed more completely in Section 6.2.1.3.6 of Volume 1.
Initiation of auxiliary feedvater to the steam generators within 20 =inutes results in primary system depressurization i
3 r
-n a
to th2 high prcssure injection system initir. tion point.
Primary system preocura continuss to fell so that ths Davis-Bassa high 4
pressure injection system pumps will be fully effective.
Based on these analyses, we conclude that delay of both redundant i
trains of safety grade auxiliary feedvater flow of-up to 20 minutes in the Davis-Besse plant will not result in uncovering.of the core and the core cladding temperature vill remain within a few degrees of saturated fluid temperature. These analyses support the conclusion that compliance with 10CTR50.46 is assumed for the Davis-Besse plant.
Section 6.2.2 of volume 1 is not applicable to Davis-Besse.
(b) Small Breaks in the Pressurizer: It has been shown (see References 1.-2, and 3 of Section 6, voltane 1) that breaks located in the pressurizer and the hot legs are less severe than breaks located in the cold leg.
To-reverify that fact, two cases involving leaks in the pressurizar were analyzed. These cases were provided in Section 6.2.3 of Volume 1.
The basic system behavior for either raised loop or lowered loop plant is tha same and can be characterized as follows:
1 (1) A rapid system depressurization due to steam release out the relief valve. This will result in reactor scram and ESTAS actuation.
(2) The indicated pressurizar level vill initially increase due to the break. After reactor scram, the pressurizer level decreases due 4
to system contraction. Following saturation of the hot legs, there vill be an insurge into the pressurizer resulting in the pressuri-zar going solid.
(3) After the pressurizer is filled by a two-phase mixture, a low quality mixture will be discharged through the valve. This will result in a large increase in the leak flow rite.
(4) primary side pressure will ultimately be maintained at approximately 1200 psia.
(5) Long-term cooling will be established by one HPI train.
The Davis Besse pilot operated pressurizer relief valve is 80 larger in relief capacity than the pilot operated relief valve modeled in the specific cases given in Section 6.3.2.1 of Volume 1.
However, as discussed more 4
, completely in paragraph 6.2.3.2.3 of Volune 1, the general phenomena observed will be the same as those observed in the breaks specifically analyzed and the consequences of the resulting break would be acceptable. In fact, as discussed in paragraph 6.2.3.2.3, breaks of this size and larger are bounded by the existing LOCA analyses. Thus, these analyses show that for the Davis Besse plant, small breaks in the top of the pressurizer equivalent to and larger than the relief capacity of the pilot cperated relief valve do not lead to core uncovering or core damage and meet the acceptance criteria of 10 CFR 50.46.
(c) Evaluation of Small Breaks which will Underce a Reoressurization burine the Course cf the Transient:
Since the course of these breaks ray be expected to be affected by the head and flow enaracteristics of tne plant's hign pressure injectio.3 3) ster, 1
plant-specific analyses have been performed for Davis Besse.
This analysis is described in Section 6.2.5 of Volume 1.
The results indicate that the core will remain covered througneut S*
ths transient and thet the critoria of 10CFR50.46 cra satisfied, thus establishing the capability of the Davis-Besse plant to safely accoc:mo-date reactor coolant system breaks in this range.
An additional case is reported in paragraph
.2.4.3.2 of Volume 1 showing the effect of asymmetric feeding to the staan generators. This case demonstrates that one generator is sufficient to guarantee core cooling in the lowered loop design. The difference in loop arrangement and EPI pump characteristics in the Davis-Besse design do not affect the characteristics of the results of this case, and this casa demonstrate the conclusion that one loop guarantees sufficient cooling for the Davis-Besse design as well.
6.3 Conclusions l
The analyses which have been performed and are documented in Section 6 of Volume 1 demonstrate that the Davis-Besse ECCS systems vill control small breaks and satisfy the criteria of 10CFR50.46. Specific conclusions applicable to the Davis-Besse plant include the following:
1.
In the highly unlikely event that both redundant trains of the safety grade auxiliary feedwater system are delayed, restoration of auxiliary feedwater at 20 minutes (and pc,ssibly later) is sufficient to assure that core daage does not result from small breaks (assuming two HPI trains operational).
2.
Analyses of relief valve failures at the top of the pressurizer show that a single ECCS train is sufficient to assure that the core remains covered during these transients.
9 G
+
3'.
The consequences of breaks in the hot legs or in the pressurizer has been demonstrated to be bounded by small break analyses performed for breaks in the cold leg pump discharge piping.
4.
For very small breaks in the Davis-Besse plant which require the steam generator as a heat removal system, it has been shown that system repressurization may occur. However, the establishmer_t of steam condensation by the steam generator as a beat removal mechanism controls the repressurization and assures effective action by the high pressure injection system to maintain the core covered with water and prevent core damage.
5.
It was demonytrated that asynsnetric auxiliary feedwater injection (a sin 61e steam generator) provides adequate heat removal to assure that core uncovery does not occur.
M h
k i
7
Appendix 1 N3tural Circulation'In B&*=' Operating Plants The information in Appendix 1 - Rev. 1 of Volume II of " Evaluation of Trsasient Behavior and Smmil Reactor Coolant System Breaks In the 177 Fuel Assembly Plant" dated S/7/79 is applicable to this section.
e 1
e b
i e
.v.-
-- ~
-n
s APP:ndix'2 - Steam Gr.ntrater Tubs Thermal Strcss Evaluation
~~
The i= formation in Appendix 2 Volume II of " Evaluation of Transient C
j Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly
{
Plant" dated 5/7/79 and Supplement 1 submitted by letter from 7. H. Taylor to Dr. R. J. Mattson dated 5/10/79 is applicable to this section. -
4 i
4
App ndiz 3 -
Sensitivity Study on th2 Effcet of Starting a Reactor
~
Coolant Pump in a Highly Voided System The information in Appendix 3 volu:ne II of "Evalua:1on of Transien:
Behavior and Small Reactor Coolant System Breaks in the 177 yuel Assembly Plant" dated 5/7/79 and a letter from J. H. Taylor to T. M. Novak dated 5/10/79 is applicable to Davis Bessa.
f e-9 e-ms-
s APPENDIX 4 OPERATING GUIDELINES FOR SMALL BREAXS The attached infonnation has been prepared to serve as a basis for developing detailed emergency operating procedures for situations involving small breaks in the reactor coolant system.
It defines the symptems, innediate aetions, I
precautions and followup actions. Fo'r various combinations of equipment availability, the required actions are outlined to take the plant to the long-term cooling mode.
The procedural guidelines form Part II of a two-part document which has been provided to the operating utilities. Part I is included as Section 5.0 of this report and describes small breaks in phenomenalogical terms.
e
/
4
,s
PART II - OPERATING GUIDELINES FOR SMALL BREAKS 1.0 SYMPTOMS AND INDICATIONS (IWEDIATE INDICATIONS) 1.1 EXCESSIVE REACTOR COOLANT SYSTEM (RCS) MAKEUP
- 1.2 DECREASING RCS PRESSURE 1.3 REACTOR TRIP 1.4 DECREASING PRESSURIZER LEVEL
- 1.5 ESFAS ACTUATION
- 1.6 LOW MAKEUP TANK LEVEL *
- May not occur on all small breaks.
2.0 IMMEDIATE ACTIONS 2.1 YERIFY CONTROL ROOM INDICATIONS SUPPORT THE ALARMS RECEIVED, ERIFY AUTOMATIC ACTIONS, AND CARRY OUT STANDARD POST-TRIP ACTIONS.
2.2 YERIFY THAT INJECTION FLOW FROM EACH HPI PUMP IS SALANCED W.EN HPI IS INITIATED.
2.3 VERIFY THAT APPROPRIATE ONCE-THROUGH STEAM GENERATOR (OTSG)
LEVEL IS MAINTAINED BY FEEDWATER CONTROL (BY ICS CONTROL OF MAIN FEEDWATER OR SFRCS CONTROL OF AUXILIARY FEEDWATER).
2.4 MONITOR SYSTEM PRESSURE AND TEMPERATURE.
IF SATURATED CONDITIONS OCCUR, INITIATE HPI.
3.0 PRECAUTIONS 3.1 IF THE HPI SYSTEM HAS ACTUATED BECAUSE OF LOW Ph!SSURE CONDITION, IT MUST REMAIN IN OPERATION UNTIL ONE OF THE FOLLOWING CRITERIA IS SAT!SFIED:
e
1.
THE LPI SYSTEM IS IN OPERATION AND FLOUING AT OF 1000 GPM IN EACH LINE AND THE SITUATION HAS 20 MINUTES.
OR 2.
ALL HOT AND COLD LEG TEMPERATURES ARE AT LEASE 50 0
BELOW THE SATURATION TEMPERATURE FOR THE EXISTING RCS FRESSURE, THE LEG TEMPERATURES ARE NOT MORE THAN 50 HOTTER 0
SIDE SATURATION TEMPERATURE, AND THE ACTION IS NECESSARY TO PREVENT THE INDICATED PRESSURIZER LEVEL FROM G 0
HIGH. IF 50 SUBC00 LING CANNOT BE MRINTAINED, THE HPI SHALL t
BE REACTIVATED. THE DEGREE OF SUBC00 LING BEYON 0
LENGTH OF TIE HPI IS IN OPERATION SMALL BE LIMITED BY PRESSURE / TEMPERATURE CONSIDERATIONS FOR TH 3.2 PRESSURIZER LEVEL MAY BE INCREASING DUE TO RCS CONDITIONS OR A BREAK ON TOP OF THE PRESSURIZER.
3.3 IF HIGH ACTIVITY IS DETECTED IN A STEAM GENERATOR
' LEAKING GENERATOR.
BOTH STEAM GENERATORS MAY NOT BE ISOLATED OTSG COOLING IS REQUIRED FOR DECAY HEAT REN VAL.
3.4 OTHER INDICATIONS WHILM CAN CONFIRM THE EXISTENC 3.4.1 RC DRAIN TANK (QUENCH TANK) PRESSURE (MAY R 3.4.2 INCREASING REACTOR BUILDING SUMP LEVEL.
3.4.3 INCREASING REACTOR BUILDING TEMPERATURE.
t 3.4.4 INCREASING REACTOR BUILDING PRESSURE.
3.4.5 INC'4 EASING RADIATION MONITOR READINGS INSIDE C
~
3.4.6 REACTOR COOLANT SYSTEM TEMPERATURE BECOMING SAT RELATIVE TO THE RCS PRESSURI.
3.4.7 HOT LEG TEMcERATURE EQUALS OR EXCEEDS PRESSUR 3
3.5 HPI COOLING REQUIREMENTS COULD DEPLETE TE 30 RATED WATER STORAGE TANK, AND INITIATION OF LPI FLOW FROM THE REACTOR BUILDING SUt?
TO THE HPI PUf75 WOULD BE REQUIRED.
3.6 ALTERNATE INSTRUENT CHANNELS SHOULD BE CHD:ED AS AVAILASLE TO CONFIRM KEY PARAMETER READINGS (I.E., SYSTEM TEMPERATURES, i
PRES $URES AND PRESSURIZER LEVEL).
3.7 MAINTAIN A TENPERATURE VERSUS TIME PLOT AND A CORRESPONDING 4
. TEMPERATURE - PRESSURE PLOT ON A SATURAT10ft DIAGRAM. THESE PLOTS WILL MAKE IT ROSSIBLE TO TRACK THE P1ET'S CONDITION THROUGH PLANT C00LDOWN. PRIMARY TEMPERATURE AND PRESSURE WILL DECREASE ALONG THE SATURATION CURVE UNTIL SUBC00 LED CONDITIONS ARE ESTABLISHED. THIS WILL,BE INDICATED BT PRIMARY SYSTEM PRESSURE NO LONGER FOLLOWING THE SATURATION CURVE, AS PRIMARY SYSTEM TEMPERATURE DECREASES. WHEN THIS OCCURS, PRIMARY SYSTEM PRES $URE SHOULD BE CONTROLLED SY ADJUSTING HPI FLOW, T0 MAINTAIN
~
- 50'F SUBC00 LING.
i 4.0 FOLLOWUP ACTIONS 4.1 IDENTIFICATION AND EARLY CONTROL' 4.1.1 IF HPI HAS INITIATED BECAUSE OF LCN FRESSURE, CONTROL HPI
.IN ACCORDANCE WITH STEP 3.1.
4.1.2 IF BOTH HPI TRAINS HAVE NOT ACTUATD ON ESFAS SIGNAL, I
START SECOND HPI TRAIN IF POSSIBLE.
4.1.3 IF RC PRESSURE DECREASES CONTINU00SLT, VERIFY THAT CORE FLOOD TANXS (CFT'S) AND LOW PRESSURE INJECTION (LPI)
HAVE ACTUATED AS NEEDED, AND BALANE LPI'.
i e
e
- g '~
4 i
~
4.1.4 ATTEMPT TO LOCATE AND ISOLATE LEAK IF POSSIBLE. LETDOWN WAS ISOLATED IN STEP 2.1.
OTHER ISOLATABLE LEAKS ARE PORY (CLOSE BLOCK VALVE) AND BETWEEN VALVES _IN SPRAY LINE (CLOSE SPRAY AND BLOCK VALVE).
4.1.5 DETERMINEAVAILABILITYOFREACTORCOOLANTPUMPS(RCP'S)
AND MAIN AND AUXILIARY FEEDWATER SYSTEMS.
4.2 ACTIONS WITH FEEDWATER AVAILABLE TO ONE OR BOTH GENERATORS 4.2.1 MAINTAINONERCPRUNNINGPERLOOP(STOPOTHERRCP'S).
IF NO RCP'S AVAILABLE, GO TO STEP 4.2.4 BELOW.
4.2.2 ALLOW RCS PRESSURE TO STABILIZE 4.2.3 ESTABLISH AND MAINTAIN OTSG COOLING BY ADJUSTING STEAM PRESSURE VIA TURBINE BYPASS AND/OR ATMOSPHERIC DUMPS.
COOLDOWN AT 1000F PER HOUR TO ACHIEVE AN RC PRESSURE OF 250 PSIG. REFER TO PRECAUTION 3.7 FOR DEVELOPMEhT OF TEMP.IRATURE AND PRESSURE PLOTS. ISOLATE CORE FLOOD TANK WHEN E00F SUSC00 LING IS ATTAINED AND RC PRESSURE IS LES 700 PSIG. 60 INTO LPI COOLING ?t1 APPENDIX A.
4.2.4 IF RCP'S ARE NOT OPERATING:
4.2.4.1 ESTABLISHANDCONTROLOT5[dEkLTO9t, INCHES INDICATEDONTHESTARTNRANGkIN5.WINTJION.
4.2.4.2 IFRCPRESSUREISDECRIASING,WAITUN1.'.IT STABILIIES OR BEGINS INCREASING.
IF IT BEw.'NS i
INCREASING, GO TO STEP 4.2.4.4.
4.2.4.3 PROCEED WITH A CONTROLLED C00LDOWN AT 1000F/HR.
BY CONTROLLING STEAM GENERATOR SECONDARY SIDE PRESSURIS. MONITOR RC PRESSURES AND TEMPERATURES l
, DURI'.G COOLDOWN AND PRbCEED AS INDICATED BELOW.
l
4.2.4.3.1 IF RC PRESSURE CONTINUES TO DECREASE, FOLLOWING SECONDARY SYSTDi PRESSURE DECREASES AND WITH PRIMARY SYSTEM TEMPERATURES INDICATING YATURATED CONDITIONS, CONTINUE C00LDOWN UNTIL AN RC PRESSURE OF 150 PSI IS REACHED, AND PROCEED TO STEP A.4 0F APPENDIX A.
4.2.4.3.2 IF RC PRESSURE STOPS DECREASING IN RESPONSE TO SECONDARY SIDE PRESSURE DECREASE AND REACTOR SYSTEM BECOMES SUBC00 LED, CHECX TO SEE THAT THE FOLLOWING CONDITIONS ARE BOTH SATISFIED:
A) ALL NOT,,AND COLD LEG TEMPERATURES
(
ARE BELO T SATURATION TEMPERA-TURE FOR THE EXISTING RCS PRESSURE.
.AND
. STEAM GENERATOR SECONDARY SIDE-3ATURATION TEMPERATURE.
IF THESE CONDfTIONS ARE SATISFIED A
- REMAiNSATISFIED.CONTINUEC00LDO ACHIEVE AN RCS TEMPERATURE (COLD LEG 280'F, AND PROCEID TO STEP A.I 0F g
APPENDIX A.
NOTE: IF THE CONDITIONS ABOVE ARE MIT AT BELOW 700 PSIG, THE CORE FLOOD TND'S SHOULD BE ISCLATED.
4 i
NOTE:
IF THE PRIMARY SYSTEM IS 500F SUBC00 LED IN BOTH HOT AND COLD LEGS AND PRIMARY SYSTEM PRESSURE IS
~
ABOVE 250 PSIG, STARTING A REACTOR COOLANT PUMP IS PERMISSIBLE.IF SYSTEM DOES NOT RETURN TO AT LEAST 500F SUB COOLING IN TWO MINUTES TRIP PUMPS.
IF FORCED CIRCULATION IS ACHIEVED, PROCEED TO STEP 4.2.
4.2.4.3.3 IF RC PRESSURE STOPS DECREASING AND TH CONDITIONS OF 4.2.4.3.2 ARE NOT MET OR CEASE TO BE MET OR IF RC PRESSURE BEGINS
. TO INCREASE. THEN PROCEED TO STEP 4.2.4.4 BELOW.
(
4.2.4.4 RES70RE RCP FLOW (ONE PER LOOP) WHEN POSSI THE INSTRUCTIONS BELOW.IF RC PUMPS CANNOT BE 1
OPERATED AND PRESSURE IS INCREASING, GO TO STEP 4.
4.2.4.4.1 IF PRESSURE IS INCREASING, STARTING A PLHP IS PERMISSIBLE AT RC PRESSURE GREATE THAN 1600 PSIG.
4.2.4.4.2 IF REACTOR COOLANT SYSTEM PPISSUF
{
STEAM GENERATOR SECONDARY PRESSURE i
PSIG OR MORE " BUMP" ONE REACTOR CO PUMP' FOR A PERIOD OF APPROXIMA SECONDS (PREPERA3LYINbPERAELESTEAM
^
GENERATOR LOOP.) ALLOW EACTOR CCOLAN SYSTEM PRESSUkE TO ST EILIZE.
CONTINUE C00LDOWN.
IF REACTOR C00.'. ANT SYSTEM PEIS AGAIN EX3EDS SECONDARY PESSUE BY 60 7
b
-.y-
-3
=. -
p,-%--
--~
PSI, WAIT AT LEAST 15 MINUTES AND REPEAT THE PUMP " BUMP". BBMP ALTERNATE PLWS SO THAT NO PUMP IS SUPJPED MORE THAN ONCE IN
~
AN HOUR.
- THIS JRY.BE iEPEATED, WITH AN INTERVAL OF 15 MINUTES, UP TO 5 TI!ES.
AFTER THE FIFT)! *8 DIP", ALLOW THE REACTOR COOLANT PUMP TO CDETINUE IN OPERATION.
4.2.4.4.3 IF PRESSURE HAS STABILIZED FOR GREATER THAN ONE HOUR, SEC0itCARY PRESSURE IS LESS THAN 100 PSIG AND PRIMARY PRESSURE IS GREATER THAN 250 PSIG, BUMP A PUMPl
+
WAIT 30 MINUTES, AND START AN ALTERNATE FtMP.
4.2.4.5 IF FORCED FLOW IS ESTABLISRER, GO TO STEP 4.2.3.
(-
4.2.4.6 If A REACTOR COOLANT PUMP C3RNOT BE OPERATED AND REACTORC0blAfiTSYSTEMPRESSUREREACHES2300PSIG, f
CPEN PRESSURIZER PORY TO REEUCE REACTOR COOLANT SYSTEM PRESSURE.
RECLCSE PERY WHEN RCS PRESSURE FALLS TO 100 P.SI ABOVE THE SECONDARY PRESSURE.
~
REPEAT IF NECESSARY. IF PEV IS NOT OPERABLE, PRESSURIZER SAFETY VALVES EILL RELIEVE OVERPRESSURE.
4.2.4.7 MAINTAIN RC PRESSURE AS INDICATED IN 4.2.4.6 IF PRESSURE INCREASES. MAINTAIX THIS COOLING MODE UNTIL AN RC PUMP IS STARTED CR STIAM GENERATOR COOLING IS ESTABLISHED AS ICICATED BY ES ASLISHING CONDITIONS DESCRIBED IN 4.2.4.3.1 OR 4.2.4.3.2.
WHEN THIS OCCURS, PROCEED AS DIRECTED IN THOSE STEPS.
Y GO TO STEP 4.2.2 IF FORCED FLOW IS ESTABLISHED.
B
APPENDIX A I
LPI COOLING
=
A.1 DETERMINE IF PRIMARY COOLANT IS AT LEAST 50 F SUSC00 LED..IF 0
4 TO STEP A.3.
A.1.1 START LPI PUMPS.
IF BOTH PUMPS ME OPERASLE GO TO STEP A.2 1
FOR ONE LPI PUMP OPERABLE MAINTAIN OTSG COOLING AS FOLLO THE OPERABLE LPI PUMP WILL BE USED TO MAINTAIN SYSTEM A.1.2 CBTAIN PRIMARY SYITEM CONDITIONS OF 280 F AND 250 PSIG.
0 A.I.3 ALIGN T E DISCHARGE OF THE OPERABLE LPI PUMP TO THE SUCTIO 0F THE HPI PUMPS AND TAKE SUCTION FROM THE BW5T.
IF THE BWST IS AT THE LOW LEVEL ALARM, ALIGN LPI SUCTION FROM THE RB SUMP AND. SHUT SUCTION FROM BWST.
A.1.4 START THE OPERABLE LPI PUMP SPECIFIED ABOVE. THE HPI-LPI
('
- ~
SYSTEMS WILL NOW BE IN " PIGGY BACK" AND HPI FLOW IS MAI SYSTEM PPISSURE.
s A.1.5 60 TO SINGLE RC PUMP OPERATION.
A.I.6 WHEN THE SECOND LPI PUMP IS AVAILABLE AL'IGN IT IN THE DEC HEAT E DE AND COMMENCE DECAY HEAT REMOVAL.
(DECAY HEAT SYSTU4 FLOW GREATER THAN 1000 GPM). SECURE REMAINING RC PUM DECAY HEAT PIMOVAL IS ESTABLISH'D.
E VERIFYTHATADEQUATENPSHEkISTSFORTHEDECAYH CAUTION:
PUMP IN THE DH REMOVAL MODE.
IF INADEQUATE, TRANSFER TO LPI MODE.
A.1.7 REDUCE REACTOR COOLANT PESSURE TO 150 PSIG SY THR; FLOW.
CONTROL RC TEMPERATURE USING THE DECAY HEAT SYS BYPASS TO MAINTAIN SYSTEM PRESSURE AT LEAST 50 PS j
PRESSURE, TO ASSUPI THAT NPSH REQUIREMENTS FOR THE DECAY HEAT PUMP ARE MAINTAINED.
1
-.--e,-
--e
,.,-.-,-2--.
..--w..
_,n
_.c.
--r---
--.ce,-m-,--
-,---,.,,--y
,,um-y,w-
,, -,, - - ~., - -.
--~w-
A.1.8 SECURE THE HPI PUMP AND SHIFT THE LPI PUNP SUPPLYING IT TO THE LPI INJECTION MODE.
A.1.9 REDUCE, REACTOR COOLANT TEMPERATURE TO 100'F BY CONTROLLING THE DECAY HEAT SYSJEM COOLER BYPASS.
NOTE: IF ONE OF THE LPI/ DECAY HEAT PUMPS IS LOST, RETURN TO OTSG COOLING USING NATURAL CIRCULATION OR ONE REACTOR COOLANT PUMP (A1).
A.2 C00LDOWN ON TWO LPI PUMoS A.2.1 MAINTAIN RCS PRESS,URE AT 250 PSIG AND REDUCE RCS TEMPERATURE TO 8
280 F.
A.2.2 ALIGN ONE LPI PUMP IN THE DECAY HEAT REMOVAL MODE.
A.2.3 SECURE ONE RC PUMP. A SINGLE RC PUMP IS NOW OPERATING.
A.2.4 ~ START THE DECAY HEAT' PUMP,IN THE DECAY HEAT P.EMOVAL MODE, AND WHEN DECAY HEAT SYSTEM FLOW IS GREATER THAN 1000 GPM, SECURE THE RUNNING RC PUMP.
A.2.5 REDUCE RC PRESSURE TO 150 PSIG BY THROTTLING HPI FLOW. CONTROL I
oAbTEMPERATURETOMAINTAINATLEAST50PSIMARGINTOSATURATION s
FRESSURE.
~
A.2.6 START THE SECOND LPI PUMP IN THE LPI INJECTION MODE. SECURE HPI PUMP.
A.2.7 SHIFT LPI SUCTION FROM THE BWST TO THE REACTOR BUILDING SUMP WHEN SUFFICIENT NPSH IS AVAILABLE.
NOTE: THIS IS DESIMBLE TO AVOID UNNECESSARY QUANTITIES OF WATER IN CONTAINMENT.
A.2.8 REDUCE REACTOR COOLANT TEMPERATURE TO 100 F BY CONTROLLING THE 0
t DECAY HEAT SYSTEM COOLER BYPASS.
NOTE: IF ONE OF THE LPI/ DECAY HEAT PUMPS IS LOST, RETURN TO OTSG COOLING USING NATURAL CIRCULATION OR ONE RC PUMP.
t l
/0 1
.* A.3 COOL DOWN RC SYSTEM AT SATURATION e
.A.3.1 MAINTAIN RC PRESSURE AT 250 PSIG.
A.3.2 ALIGN ONE LPI PUMP TO SUCTION OF THE HPI PUMPS AND THE SUCTIO TO THE REACTOR BUILDING SUMP. (SHUT BWST SUCTION VALVE F PUMP.)
A.3.3 WHEN THE BWST LEVEL REACHES THE LO-LO LEVEL LIMITS. ' START TH LPI PUMP AND SHUT THE HPI PUMP SUCTION FROM THE BWST.
A.3.4 WHEN PRIMARY SYSTEM TEMPERATURE BECOMES SUBC00 LED BY AT LE 0
50 F, GO TO A.1.1.
A4 COOLDOWN WITHOUT REACTOR COOLANT PUMPS A.4.1 RCS INITIAL CONDITIONS ARE: PRES $URE 150 PSI, TEMPERATURES AT SATURATION.
A.4.2 ALIGN LOW PRESSURE INJECTION SYSTEM FCR SUCTION FROM REACT
' BUILDING SUMP AND PLACE INTO SERVICE.
A.4.3 CONTROL RC TEMPERATURE WITH DECAY HEAT COOLERS.
IF ONLY ONE LPI PUMP IS AVAILABLE, CROSS CONNECT DISCHARGES AND BALANCE FLOWS.
A.4.4' ISOLATE CORE FLOOD TANXS.
A.4.5 GO TO STEP A.1.1 AND FOLLOW THE PROCEDURE GIVEN THERE, IGNORING THE JESTRUCTIONS RELATING TO RC PUMP OPEPATION.
i I
//
~
Appendix 5 - BG Assessment of " Decay Heat Removal During A Very Small Break LOCA for BW 205 Fuel Assembly PWR", January,1978, C. Michelson.
The information in Appendix 5 of Volume II of "Ivaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 F al Assembly Plant", dated 5/7/79 is applicable to Davis Besse. The only difference betveen the 177 FA plant described in item 1 on page A 5-3 and Figure A-5-1 and Devis Besse is that, Davis Besse b.ss the dual advantages of the raised steam generators (similar to the 205 FA plants) and the injection location for at d'4=*y feedvater near the top of the steam generator similar to the other 177 FA plants.
I k
4 l
l I
(
l i
i 4
l f