ML20237J901
| ML20237J901 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 02/13/1986 |
| From: | Williams J TOLEDO EDISON CO. |
| To: | Miraglia F Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-1177 1241, NUDOCS 8708260359 | |
| Download: ML20237J901 (35) | |
Text
,_
4 I
'..g.,
1 TOLEDO Docket No. 50-346 EDISON License No. NPF-3 JOE WILUAMS. JR Senor Vce Artscert-Nuclear Serial No. 1241 (41w 249 23co (419)249-5223 February 13, 1986 Mr. Frank J. Miraglia, Director Division of PWR Licensing - B PWR Project Directorate No. 6 United States Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Miraglia:
Toledo Edison has reviewed the draft Safety Evaluation Report (SER),
draft NUREG-1177, dated January 29, 1986 (Log No. 1914). The staff's effort and cooperation in reviewing the Davis-Besse Course of Action has been commendable. Toledo Edison supports the staff's overall conclusion that "... the health and safety of the public will not be endangered by the resumption of power generation..." at Davis-Besse.
To facilitate your review, TED has provided comments in the same format (i.e., the same Table of Contents and Section numbers) as in the draft SER. This submittal is provided as Revision 7 to the Davis-Besse Course of Action (COA) document, and constitute Appendix III.3.
For each numbered section or subsection of the draft SER, TEDS response, Attachments 1 and 2, indicates "no comment" or recommends revised text (i.e., words, phrases, paragraphs, etc.).
This revised text is recommended for clarification and reflects a summary of previously submitted information.
However, one section, Section 4.2, provides new information for your review.
Two recommendations were made by the NRC staff. Attachment 6 specific-
{
ally identifies both recommendations by the staff and references the
]
section in Attachment 1 that provides a description of any actions 1
implemented, or planned to be implemented with the associated completion /
I disposition date.
It is intended that these descriptions and completion /
disposition dates provide the basis for the Staff to disposition their recommendations in the final SER. provides our intentions with regards to those issues q
identified by the Staff in Section 1, Table 1.2A, requiring resolution i
prior to restart. Attachment 4 provides our intentions with regards to those issues identified by the Staff in Section 1, Table 1.2A, requiring disposition after restart.
O j
8708260359 860213 PDR ADOCK 05000346 g
E PDR l
1 THE TOLEDO EDISDN COMPANY EDISON PLAZA 300 MADISON AVENUE TOLEDO, OHIO 43652 l
l:
p
.' g,,,
Dockst No. 50-346 f,-
l Licensa No. NPF-3 Serfal No.-1241 February 13, 1986 Page 2 There are five sections in the draft SER for which the Staff identified that the content will be provided later. These are Sections 3.2.1.7, Motor-Operated Valve Operator Malfunction; 3.2.1.9, Main Steam Header Pressure; 3.3.1.4, Safety Features Actuation System; 3.3.1.5, Balance-of-Plant Improvements; and 3.4, System Review and Test Program.
It is our i
understanding that these Sections will.be forwarded individually in draft form to TED as they become available. We will provide our responses as soon as possible to minimize any delay's in the issuance in the final SER.
Very truly yours.
W=-
r
- DBI:lah cc: DB-1 NRC Resident Inspector i
______-_-____________-_____a
\\
' Docket No. 50-346 License No. NPF-3 Serial No. 1241 ABSTRACT No Comments Provided.
1 INTRODUCTION No Comments Provided.
5 2
BACKGROUND
)
2.1 Brief Description of the Event No Comments Provided.
2.2 Summary of NRC Actions No Comments Provided.
2.3 Summary of Toledo Edison Company Response Page 2-3, 3rd line states:
"... systems important to safety..."
TED Response:
There are many places the Staff utilfres the term "important to safety". Traditionally, Toledo Edison's program and activities center around nuclear safety related or non-nuclear safety l
related. A third description in our Course of Action was used
(
as "important to safe operation". Toledo Edison wanto to ensure that the scope of our use of this terminology is not mistakenly equated to "important to safety" as identified in 10 l
CFR 50, Appendix A.
l 3
EVALUATION OF TOLEDO EDISON COMPANY ACTIONS No Comments Provided.
3.1 Management and Programmatic Aspects No Comments Provided.
3.1.1 Management Restructuring Page 3-4, recommend to insert the following after last paragraph:
A new division called Information Management is responsible for records management, computer systems management and documentation and correspondence control.
' Docket No. 50-346 License No. NPF-3 Serial No. 1241 This group contains the Records Management organization which previously had not been part of the Nuclear Mission organization.
Page 3-5, 9th line states:
" Staff Recommendation that Toledo Edison Company consider the establishment of an Independent Safety Engineering Group at the i
Davis-Besse Station."
l TED Response:
Toledo Edison has organizationally established a Nuclear Safety Department with responsibility'es closely approaching those identified in Item I.B.1.2 of NUREG-0737 for an Independent i
Safety Engineering Group (ISEG). Revision 4 of the COA dated November 16, 1985, provided a description of the major areas of responsibilities for this Nuclear Safety Department and position descriptions for its staff. This organizational entity reports to the Director of the Nuclear Safety and Licensing Division.
The Nuclear Safety Department has been established to be intrinsically independent of both the Nuclear Engineering and Plant Management organizations. This ensures independence in reporting, as well as in its functioning, from organizations directly in the chain for power production. Attachment 2 Figure I highlights the Nuclear Safety Department's reporting I
relationship and independence from other major divisions.
Through the recent appointment of Mr. Sushil C. Jain as the Nuclear Safety Manager, new leadership and direction has been provided to the Nuclear Safety Department. Under Mr. Jain, the emphasis of the Nuclear Safety Department will be to develop and l
institute programs for more in-depth technical reviews, assess-ments and evaluations in the areas which may affect nuclear safety.
It is noted that there are varying degrees to which an ISEG function can be performed while still meeting the requirements of Item I.B.I.2 of NUREG-0737.
It is Toledo Edison's intent to I
further strengthen this function within the Nuclear Safety Department. To that end, steps will be taken to further augment the Nuclear Safety Department manning beyond that specified in l
Revision 4 of the COA, thereby providing for increased independence and wider coverage in areas of nuclear safety.
It is emphasized that although staffing is not complete, the organization has already sponsored and/or supported such post-June 9 activities as the Decay Heat Removal Task Force, Safety Evalua tion Review Group and the Independent Process Review Committee. Toledo Edison staffing and formalization of the day-to-day role of this group is continuing at the present time and is expected to be completed with the remainder of the _
_a
. _ - _ _ _ _ _ = _ _ - _ _ _ _.
4
- Dackst No. 50-346-
)
[-
Liesnso No. NPF-3 L
Serial No. 1241-L l
t L
Nuclear Mission organization in 1986. Further staff augment-l ation as referenced above is expected to be' completed by 1987.
- 3. l ~. 2 Maintenance No Comments.Provided.
'3.1.3 Procedures and Training i
No Comments Provided.
3.1.3.1 Plant Operating and Emergency Procedure No Comments Provided.
3.1.3.2 Role of Shift Technical Advisor No Comments Provided.
3.1.3.3 Reporting of Events No Comments Provided.
3.1.3.4 Security No Comments Provided.
3.1.3.5 Training No Comments Provided.
3.1.4 Operating Experience Feedback and Post-Trip Review Page 3-23. TED recommends inserting the following after the first paragraph:
The responsibilities of the Operations Assessment Program have been transferred to the Operations Engineering Department of the Nuclear Engineering Division as of January, 1986. The program's scope will. continue to encompass those elements of the current program and will focus on additional topics of operating experience interest.
Important items will be tracked internally by the group responsible for the programs operation. Operations Assessment Program Procedures based on Operations Engineering standards will be in place by June, 1986.
3.2 Plant Review 3.2.1 Event-Specific Investigations L_
L_- _
Dickst No. 50-346 Liesn2a No. KPF-3 Serial No. 1241 3.2.1.1 Auxiliary Feedpump Turbine Overspeed and Control 3.2.1.2 Auxiliary Feedpump Turbine Trip Throttle Valve Page 3-27, TED recommends revising the second paragraph as follows:
l The OTM consists of a spring-loaded tappet in the turbine casing. The tappet is struck by a spring-loaded weight and I
when the weight is pulled sufficiently away from the turbine shaft by centrifugal force. Once the tappet is struck, it moves away from the turbine shaft striking the leaf spring / tappet and releases the spring-loaded trip linkage. The linkage releases the latch on the T&T valve, thereby allowing the spring in the T&T valve to close;the valve. Resetting the AFPT overspeed trip involves manually moving the linkage, resetting the OTM, resetting the latch on the T&T valve, and re-engaging the valve operator to the valve internals. If the linkage is not moved f ar enough, the OTM will not reset and, if the T&T valve latches, the latch will hold only because of the friction between the parts of the linkage.
3.2.1.3 Spurious Steam and Feedwater Rupture Control System Actuation and Spurious Main Steam Isolation Valve Closure i
Page 3-31, TED recommends revising the last two paragraphs on I
Page 3-31 and the first paragraph on Page 3-32 as follows:
The licensee has reviewed data available from Davis-Besse and from other nuclear plants to determine the effects of sudden TSV closure on steam generator level-sensing instrumentation.
Data recorded during a pre-operational turbine trip test from-75% power at Davis-Besse show that oscillations occurred in the sensed / indicated steam generator level (by the startup range level transmitters). The oscillations caused the falsely indicated levels to be 50 in. or more below the actual indicated level immediately following turbine trip. The oscillations were of short duration, less than 200 milliseconds (msec), and the amplitude of the oscillations decreased significantly after several cycles.
The licensee reviewed transient reports from three other nuclear plants that revealed oscillatory behavior in the level transmitter outputs following reactor / turbine trips, apparently caused by pressure oscillations in the main steamlines caused by TSV closure.
Bailey BY level transmitters were installed during the Davis-Besse turbine trip pre-operational test.
During the fourth (1984) refueling outage, these transmitters were placed with Rosemount' 1111 transmitters. This caused a change in the hydraulic configuration and responsivencan of the Rosemount 1152 startup range level transmitters th' t feed the SFRCS. The a
Licensee, therefore, believes that this increased _
Dockat No. 50-346 License No. NPF-3 Serial No. 1241 responsiveness of the Rosemount 1152 transmitter outputs (resulting in more amplified oscillatory output conditions that i
were caused by steam line pressure oscillations from TSV closure on a turbine trip) was the root-cause of the spurious SFRCS actuation during the June 9, 1985 event.
A review of Figures 3.2 and 3.3 of NUREG-1154 (plots of steam generator level as a function of time during the event) indicates that the transmitter output oscillations would have to be approximately 70 to 90 in, in amplitude, only slightly greater than the oscillations exhibited by the Bailey transmitters, to cause the spurious SFRCS low level actuation.
An analysis performed for the licensee by MPR Associates has estimated that the apparent magnitude of level swings shown by the Rosemount 1152 transmitters following turbine trip from 100% power could be increased several times because of the change in its hydraulic configuration caused by replacement of 1
Bailey transmitters with Rosemount 1153 transmitters. This is 6
caused by the increased sensitivity of the Rosecount 1152 transmitters and the change in the instrument sensing line hydraulic configuration required for installation of the Rosemount 1153 transmitters.
It was estimated that the effects for the SFRCS actuation channel No. 2 would be more pronounced a
because of the level transmitter configuration. The licensee believes that the SFRCS full trip control room annunciator point did actuate at the time of the trip, but that, because the trip was present for only a short duration and because of several annunciator acknowledgement and reset actions by the control room operator, as is normal following any reactor trip, the annunciator had returned to normal by the time the operators looked to see if an SFRCS trip had occurred.
On the basis of a review of the licensee's analysis, the staff concurs with the licensee's determination of the root-cause for the spurious SFRCS actuation.
3.2.1.4 Main Feedpump Turbine and Control Failure No Comments Provided.
3.2.1.5 Turbine Bypass Valve, SP 13A2, Actuator Failure No Comments Provided.
3.2.1.6 Power-Operated Relief Valve Malfunction During the Event on June 9, 1985 No Comments Provided.
3.2.1.7 Motor-Operated Valve Operator Malfunction No Text. No Comments Provided.
- 1-
.D:ckat'No. 50-346 Licanaa No, NPF-3 Serial No. 1241 3.2.1.8 -Source Range Nuclear Instruments l
No Comments Provided.
j i
3.2.1.9 Main Steam Header Pressure No Text.- No Commet$ts Provided.
3.2.1.10 Starting Feedwater Valve, SP-7A
'No Comments Provided.
3.2.1.11 Spurious Transfer of Auxiliary Feedvater Suction to Service Water Page 3-45, 19th line states:
{
"Therefore, the staff continues to recommend removal of the strainer in the common suction Ifne, eliminating a possible common mode failure."
TED Response:
The mesh size of the strainer in the common suction lines has been increased, and the baskets in the two pump specific suction lines have been removed.
The strainer in the common suction line remains to protect the pumps from large pieces of debris which could potentially damage both pumps. The recent seismic event at the Davis-Besse site highlights'the possibility of generating debris in the common suction line.
In the event the suction strainer becomes clogged and restricts flow, a low suction pressure will be created and the pump suction will automatically transfer suction to the safety grade service water system. This would bypass the restricted strainer, thus avoiding a common mode system failure.
Without this strainer, debris could affect both pumps which would not be correctable by the auto-transfer.
Note:
This same response addresses the Staff's recommendation in Section 3.3.1.2, Page 3-63, Line 10.
- 3. 2,. 2 Thermal Transient Effects en Reactor Coolant System 3.2.2.1 Reactor Vessel No Comments Provided. 1
D2ckst Ns. 50-346 Liesnsa N=. NPF-3 Serial.No. 1241 3.2.2.2 Pressurized Thermal Shock Page.3-47. TED recommends the following new section:
In accordance with 10 CFR 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock (PTS)' Events, an evaluation of the D-B reactor vessel was performed.
10 CFR 50.61 provides screening criteria to indicate that th risk from PTS events is acceptable when the calculated valve of RT is less than 270' for longitudinal welds and less than 30 b for circumferential welds. The limiting weld in the Davis-Besse reactor vessel is the middle (beltline) circumferential weld. Hence, the RT r this PTS weld must be less than 300*F to satisfy the screening criteria. Additionally, a method for the evaluation of PTS events is. contained in the staff's generic evaluation of PTS as documented in SECY 82-465.
In accordance with the prescribed methods of 10 CFF 50.61, paragraph b, the 1_icensee had submitted the results of the regt. ired RT ca cu ati ns on January 20, 1986, reference PTS Licensee correspondence Serial No. 1236. As a result, it was determined that the RT at the inside surface of the limiting PTS veld when the Davis-Besse transient occurred was 156'F. This calculated valve of RT is substantially less than the establishedscreeningcr5teria.
PT Figure D-9 in SECY 82-465 provides a generic evaluation of the effect of PTS transients on the critical values of RT final water temperature T, pressure and the cooldown rates,cause g
crack initiation in a reactor vessel. The cooldown rate is expressed as Beta, the reciprocal time constant.
For the Davis-Besse transient, the most rapid cooldown occurred during the firy five minutes. This results in a Beta of approximately
.04 min The staff's most current method of predicting the increase in RT resulting from neutron irradiation damage is NDT documented in proposed Regulatory Guide (RG) 1.99, Rev. 2
" Radiation Damage to Reactor Vessel Materials." Using the method documented in RG 1.99, Rev. 2, the RT at the inside surfaceofthelimitingweldwhentheDavis-Esetransient occurred was 168'F.
Figure D-9 indicates for the transient 'ne final water temperature must be below the RT for the weli metal to cause crack initiation. BecausethMTy3, for the limiting veld is 168'F and the lowest water temperature dusing the transient was 545'F, the water temperature in the vessel would have had to drop an additional 377'F to cause crack initiation and to be a significant PTS event.
3.2.2.3 Once-Through Steam Generator No Comments Provided. !
__ _ _ __ _m
{
. D:cke t No. 50-346 Licenso No. NPF-3 Serial No. 1241-I 3.3 leprovement Programs and System Modifications l
No Comments Provided.
j 3.3.1 Evaluation of Facility Modifications
?
3.3.1.1 Steam and Feedwater Rupture Control System
)
Page 3-53. TED recommends replacing Page 3-53 with the J
following:-
The short-term recommendation proposed by the licensee'for implementation before restart to resolve the staff's concern on I
ccmplete isolation of feedwater from both steam generators is
~
to modify the SFRCS logic'to prevent isolation of AFWS flow to both steam generators if steam generator low pressure conditions were to be sensed in each steam generator; only the first steam generator with a low pressure condition will be isolated.
Additional modifications to the SFRCS to be implemented before restart to improve system performance and reliability include (1) modifying the SFRCS logic to prevent the unneeded isolation of the main steamlines and main feedwater lines when steam generator low level conditions are sensed (2) filtering of the steam generator low and high level SFRCS actuation signals to prevent spurious actuations caused by pressure transients (e.g., turbine stop valve or MSIV closures) which are not indicative of changes in steam generator inventory (3) providing a dedicated manual reset feature for the SFRCS 4
full trip control room annunciator point, which requires the operator to perform a deliberate and separate action q
in order to clear (reset) the annunciator (4) providing additional cooling capability for the cabinets housing the SFRCS electronic power supplies For low pressure in one steam generator, the SFRCS will l
continue to isolate the associated main steamline, main feedwater line, and the auxiliary feedwater line to that steam l
generator, and align AFWS flow to the other steam generator.
However, with the modified SFRCS logic, if pressure in the second steam generator should fall below the trip setpoint value, AFWS flow will continue to be provided to the second steam generator. Upon isolation of the first steam generator, a signal is generated to block (prevent) automatic isolation of the second steam generator. Therefore, only one steam
f Decket No. 50-346 License No. NPF-3 Serial No. 1241 i
generator may be isolated at a time by the.SFRCS in response to steam generator low pressure conditions ensuring 3.3.1.2 Auxiliary Feedwater System Although there have been no quantitative analyses to determine the improvement in the availability of the AFWS for some of these changes, these change, except for the removal of strainers in the AFP suction lines, will tend to provide a more available AFWS and are, therefore, acceptable.
3.3.1.3 Motor-Driven Pumps Toledo Edison basis for delaying provision for automatic initiation,'is to prevent uncontrolled overcooling, of the steam generators. This issue will be evaluated as part of the more comprehensive reliability study of the AFWS. Provisions for a safety-related source of water has been addressed by the licensee by their commitment to provide a connection to the SWS for.the MDFP prior to Fuel Cycle 6 cperation.
3.3.1.4 Safety Features Actuation System No Text..No Comments Provided.
3.3.1.5 Balance-of-Plant Improvements i
No Text. No Comments Provided.
3.3.2 Ongoing Improvement Programs No Comments Provided.
3.3.3 Control Room Review and Improvement Page 3-72, TED recommends the last on Page 3-72 to read as follows:
Programs to Reduce the Likelihood of Inadvertent Isolation of Auxiliary Feedwater to Both Steam Generators The licensee's proposed work plan, shown in Exhibit 5 of the licensee's submittal dated September 30, 1985, should result in adequate human factors improvements to the steam and feedwater rupture control system panel. These improvements would minimize the likelihood of inadvertent isolation of auxiliary feedwater to both steam generators. The retraining of control room operators with respect to these modifications and their impact on the manual actuation of the SFRCS are addressed in Section 3.1.3.5.
This retraining will compliment the training _
i l
G
'Dsekst Ns. 50-346 Liesuso No. NPF-3
- Serial No. 1241 -
required for licensed operators on all plant modifications as a part of the plant modification process.
Detailed Control Room Design Review (DCRDR)
Page 3-75, TED recommends the following additional paragraph be inserted at the end of para 6raph (5):
Assessment of HEDs The HEDs to be corrected during the present outage will be identifie.' and justifications for those HEDs not corrected or only partially corrected will be provided no later.than February 28, 1986.
Page 3-76, TED recommends the following additional paragraph be inserted at the end of paragraph (6):
Selection of Design Improvements Items (a) and (b) will be provided no later than one month following the end of the current outage. Items (c), (d), and (e) are linked to the completion of the special studies which are currently scheduled to be completed by the end of the next (fifth) refueling outage.
These remaining items will, therefore, be provided prior to the restart from the next scheduled refueling outage.
Page 3-77, TED recommends the following new paragraph at the end of paragraph-(8):
Coordination of the DCRDR with Other Improvement Programs This documentation will be provided by the licensee no later than two months following restart from the current outage.
i 3.4 System Reviews and Test Procedures No Text. No Comments Provided.
3.4.1 Component and System Testing Before Restart l
3.4.2 Integrated Systems Testing at Power (input from contractors assistance also being considered) 4 EVALUATION OF DECAY HEAT REMOVAL RELIABILITY AND CAPABILITY 4.1 Auxiliary Feedwater System Before June 9, 1985 No Comments Provided. e
r L
4 L
4 l,
s.
Dockst No.'50-346 L
Licensa No. NPF-3 Serial No. 12411
- Attachment 1 l
l 4.2 Auxiliary Feedwater System Before Restart After the Event on June 9, 1985 Page 4-3, line 2 states:
"... that thg calculgted AFW system unavailability does not meet the 10 to 10 per demand criterion of SRP Section I
10.4.9..."
TED Response:
l ToledoEdisonhasreviewedthecagculatiogsandtheAFW unavailability does mee_t the 10 to 10 per demand criteria of SRT Section 10.4.9.
4.3. Power-Operated Relief Valve /High Pressure Injection / Makeup f
System for Makeup /High-Pressure Injection (MU/HPI) Cooling TED Comment:
Further analysis has modified the specific results of the previously submitted computer analyses. Therefore, the Section 4.3 should be revised to reflect the new information in.
This revises Appendix C.3.1 - Transient Analysis Program Results. NOTE: This section supersedes Revision 6 to
]
Appendix C.3.1 of the Davis-Besse Course of Action. The results i
are similar, therefore, the conclusions should be the same.
5 CONCLUSIONS J
i No Comments Provided.
l
)
I l
l 1
l
- L
m T
. N T E N D E N M E T P - L R EEC g
A iHU I
n P NTN ri r
E I
arpo D
eeut E
l eoc S
iTN l
cnre I
TNE uiG r E
l i
N g i
T FMI n
D OES E
Y G
NGI V ENV g
O PI rnr R ERR aie P DEO esg NEI l na E
I NN cen B
I uca EGS nim O
iN l
L T TEE S TDT I
CN nn I A oo N P t
ii O ET n
d t t I
DNG e
ng aar I
T A
d r ranr ugo C OLI ria a
io l i s e
N TPT ose eyst ati r
U i el l t nc vsv u
F NMO nrc c eee E er g
WOP eP u u f cr ve i
G 0RE S
N N
aii ynp F
N 1
F 1
- e SLD t I u I
S c
e S
R N
i V
fd E SOO an E
I I I S a N
T I
TCA G RNZ -
N AUI E
iFN l
C r
r Y
G aye T NNRA et g E
OI O l ea F
I R cf n A TEER uaa S
AEl NSM ZNTS T
I I N NGFN E ANOO D GE N R t r E
OYC no P
T r
ye s E SEAS t mi D
l F
I t e es v N
A ng i
f sr I
TS B aa l n aee S s p P a s u M
1 2
AS E E T T O O N N
(
Y Docket'No. 50-346 License No.-NPF-3
' Serial No. 1241.
' Attachment 3 TABLE 1.2 UNRESOLVED ISSUES A.
Resolution Required - Before Restart Authorization NRC Item TED Pesponse 1.
. Provide justification for not This justification shall be correcting those outstanding safety provided.no later than significant HEDs and for those that February 28, 1986.
See will'only be partially corrected.
insert at end of paragraph (5)
Section 3.3.3. Page 3-75.
Section 3.3.3, Page 3-75.
2.
Confirm cospliance with the concerns Compliance with these concerns identified in IE Bulletin 85-01 with will be achieved through revi-respect to. steam binding of MDFP.
sion to SP.1106.06, prior to Section.3.3.1.2,'Page 3-66.
Restart.
3.
The staff will conduct another survey No response-required at this at the site before restart to judge time.
4 the effectiveness of modifications.
to maintenance programs and organization..." Section 3.1.2, Page 3-8.
'4.
(Potential Open Issues-on MOVs)
No response required at this time.
5.
(There may be some issues for No response' required at this restart from the system review and time.'
. test program.)
l L
i' l
Dockat No. 50-346 Licansa No. NPF-3 l
Serial No. 1241 TABLE 1.2 UNRESOLVED ISSUES B.
Resolution Note Required for Restart Issues NRC Item TED Response 1.
Submit training program descriptions TED will submit training of infrequent, critical and difficult program descriptions prior tasks for review by the Staff, when to Restart.
completed. Section 3.1.3.5, Page 3-20, 2.
Submit, within 60 days after Restart, TED will submit proposed proposed technical specifications for Technical Specifications the MDFP similiar to those applicable no later than 60 days to the AFPs.
Section 3.3.1.2, Page 3-64.
after Restart 3.
Submit for review a comprehensive TED will submit study no reliability study of the AFW system later than 90 days after to determine if further improvement Restart.
is required. Sections 3.3.1.2 and 4.2, Pages 3-55 and 4-2.
4.
Submit outstanding information Paragraph (6), Page 3-76.
related to the DCRPR to enable Items (a) and (b) will Staff to complete its review as be provided no later than a separate licensing action.
one month after Restart.
Section 3.3.3, Pages 3-72 through Items (c), (d), and (e) 3-77, will be provided prior to start of Cycle 6.
Paragraph (8) Page 3-77.
Documentation to be provided no later than two months after restart.
j I
5.
Ccmplete power supply modifications TED is attempting to j
to SFAS to ensure redundant channel complete these modific-independence. Section 3.3.1.4, ations prior to Restart.
Page 3-67.
These modifications will be completed prior to the start of Cycle 6.
1 i
l I
m
!Dockst No. 50-346-
- Licanso No. NPF ' ;..,
. Serial No. 1241' d'
' Attachment 5-
.r v
i i
' Appendix C.3.1 T_ransient Analysis Program Results Note:.This'section supersedes Revision 6 to
-Appendix C.3.1 of the Davis-Besse Course of Action.
I y
{
)
l APPENDIX C.3.1 - TRANSIENT ANALYSIS PROGRAM RESULTS To calculate a "best estimate" response for a LOFW event, the RELAP5/ MOD 2 computer program was used for all ECCS analyses. These analyses were performed by Babcock & Wilcox. The RELAPS model developed for Davis-Besse is shown in Figure C.3.1.1.
The acceptability of using the RELAP5/ MOD 2 1
program and of the modeling techniques utilized for the ECCS analyses was established by successfully benchmarking RELAPS results to Integral System Test (IST) Program experimental results from Once Through Integrated System (OTIS) Test 230299.
It is noted that the use of systems to support the makeup /HPI cooling mode depends upon the use of RCS ccde safety valves, the PORV, and high point and pressurizer vents. These systems are safety related except for the PORV (the PORV block valve is safety related). For injection capability at RCS operating pressure, the makeup system is utilized. Although not originally classified as a safety grade system, the quality attributes of the system and its support systems compare well with other safety grade systems. The required continuous service of one pump at all times, coupled with the procedural immediate action of starting the second pump following a reactor trip results in a high system reliability. The operability of both makeup pumps is required in the Technical Specifica-tions.
All analyses performed incorporated the following set of assumptions.
These analyses confirm that the June 9 conditions would not have resulted in core uncovery given initiation of feed and bleed within the first 30 minutes.
Appendix C.3.1 1
' Reactor _at 90% of full power for Case A and 102% of' full power for t'
Case B prior to LOFW LOFW initiated by 10 second ramp down of main feedwater flow l7 Realistic decay heat based upon 1979 ANS 5'1 Standard Reactor. trips on high RCS pressure of 2300'psig by RPS Turbine trips I second after Reactor trip r
o' l Main Steam.Line pressure controlled at 1050 psig by safety valves Make-Up Pump flow, prior to. operator action, maintains pressurizer 17
. level at 198 inches
,RCS Pumps trip on loss of sub-cooled margin Letdown flow isolated throughout the transient AFW System unavailable throughout the transient
~
Start-Up Feed Pump unavailable throughout the transient Additionally,' based upon the present Davis-Besse ATOG Procedures, all analyses. assume the following. operator actions upon determination that both lack of heat transfer and lack of feedwater conditions exist.
Open PORV and PORV block valve Open pressurizer and hot leg high point vent lines Actuate both Make-Up Pumps Align HPI Pumps in piggyback mode i
i These assumptions conservative 1y' bound the actual plant response to a total loss of feedwater.
In particular, they provide for no decay heat removal other than that provided by operator actions. For example, Appendix C.3.1 2
Rev. 7
O letdown'f'lew would probably not be isolated and would provide some heat removal capability.
The following cases were analyzed:
Case A - June 9 transient with no recovery of ' Secondary Feedwater.
Makeup /HPI cooling is initiated 30 minutes after Reactor Trip.
Case B - Operator action 10 minutes after RCS hot leg temperature reaches 600*F for LOFW at 102% power.
Analyses were performed using Case A assumptions for two different PORV 7
flow capacities. The two capacities assumed are a minimum flow of 160,000 lb/br at 2500 psia pressure in the pressurizer and a maximum capacity of 226,000 lb/hr at 2500 psia pressure in the pressurizer.
Based upon testing of the Davis-Besse PORV design performed by Crosby at its facility and testing performed at the Duke Power Marshall Station, the actual th Davis-Besse PORV capacity at the time of the June 9 event was 195,000 lb/hr at 2500 psia pressure in the pressurizer. Therefore, the Davis-Besse June 9, 1985 transient is bounded by the two cases analyzed.
Both analyzed cases demonstrated that core uncovery would not have occurred. The results for the 160,000 lb/hr PORV flow case is presented for Case A in this report since it represents a more conservative analysis.
l l
l l
l Appendix C.3.1 3
Rev. 7 i
i i
Case B' analyzes the feed and bleed transient associated with a complete loss of feedwater from full power for plant operation after June 9, 1985.
For operator action feed and bleed cooling of the RCS is to be initiated when hot leg temperature exceeds 600*F.
PORV flow capacity for this case l
.is based upon design changes made to the PORV nozzle following the June 9, 1985 event.
Based upon testing performed by Crosby at its facility t
and testing performed at the Duke Power Marshall Station, the flow capacity for the revised Davis-Besse PORV is ~ 211,000 lb/hr at 2500 psia pressure in the pressurizer.
The results of Cases A & B are summarized on the following pages
.7 and are reflected in " collapsed" water level above the core. This i
level determination is therefore conservative in core cooling consider-ations.
S_UMMARf of RESUL't FOR CASE A CASE DESCRIPTION (Refer to Figures C.3.1.2 to C.3.1.5)
TITLE: June 9,1985 Transient without Recovery of Secondary Side Feedwater INITIAL CONDITION:
Power - 90% of 2772 MW KEY ASSUMPTIONS:
l Appendix C.3.1 4
Rev. 7 i
l ICS controls SG Water Level @35" for 4\\ minutes
- 12 Make-up pumps on until pressurizer level reaches 200"
- PORV actuates 3 times After 3rd actuation PORV closed when RCS Pressures reaches 2080 psia Pressurizer Spray actuates @2200 psia
initiate make-up flow @ 30 minutes.
l PORV:
I.D. = 1.54 in., flow set to 160,000 lb/hr steam @ 2500 psig Open @ 30 minutes.
7
)
MU:
2 pumps @ 30 minutes s.
HPI:
2 trains in piggy-back @30 rinutes i
SEQUENCE:
1 TIME 0-1800 s~ec Sequence as experienced June 9 without restoration of startup or auxiliary feedwater.
t 1800 sec Operator initiates makeup /RPI cooling.
)
i l
Appendix C 3.1 5
Rev. 7 i
/
- 14)
L u.
..g,i
' (,
W p
7
- a p.
.),)), h, l'
h; >
e l
4000 sec.
Minimum reactor vesseY water level at 3 feet o
,,, + '
i above the core.
,l
./
/
,o y
5400'see Reactor coolant system max 6ne pressure of
.s.
2500 psi' starting to depress'urize.
4
~
)
9000 see Reactor. coolant system pressure 2100 psi and N
declining c;, j RESULTS: No core uncovery
, i
}
k,.p S_UMMARY OF RESULTS FOR CASE B 7
CASEDESCRQfION (Refer to Figures C.3.1.6 to C.3.1.9)
,o r
A
}
TITLE:
100% Power with New Procedures
.4 I
M. ',
Loss ~of Feedwater. No Auxiliary Feedwater. No startup j
1 Feeddater. Operator initiation of makeup /HPI cooling 10 minutes
.g after RCT;-Hot leg 3 temperature reaches 600*F.
j
.,[.
\\
p A, ' rY.$
j i
.i.,
INITIAL CONDITION: Poher -= 102% of 2772 MW
- y L
J y ',.*
~>
,v i
1 AppendidC.3.E t,
'6 Rev. 7 j f a,,.,
_ __ - - _. - - - = - - -
=
Y f
KEY ASSUMPTIONS:
- Decay heat = 1.0 x 1979 ANS-
- High' point vents, PORV and make up flow initiated 10 minutes after T
= 600*F g
PORV: Flow set to 211,000 lb/hr steam @ 2500 psia MU:
2 pumps i
HPI: 2 trains in piggy-back 7-SEQUENCE Time:
O see LOFW 15.25sec Reactor trips on high RCS pressure 96 see Hot leg temperature reaches 600 F 696 sec Operator initiated feed and bleed 705 sec RC Pumps tripped on Loss-of-sub-cooled margin 1200 see Reactor Coolant System Pressurizes to 2400-2500 psi.
2850 see Depressurization of the Reactor Coolant System begins.
Appendix C.3.1 7
Rev. 7
_ _ _ _ _ _ _ _, _ _ _. - - - - - - - - - ' ' - ^ ^ ^ ' '
l L
3730 see Minimum Reactor Vessel collapsed water level reached is 0.1 feet below the top of the core.
7-RESULTS: No core uncovery due to froth in core region.
Key RCS parameters are shown in figures C.3.1.2 through C.3.1.9 for each of the cases analyzed. -The parameters shown are:
Collapsed Liquid Level in Core 7
Loop A Hot and' Cold Leg Temperature Loop B Hot and Cold Leg Temperature Hot Leg Pressure in Loop with Pressurizer
)
Appendix C.3.1 8
Rev. 7
\\ll g$
l 1_ 0 86 h
r e N
-]
s_
gl "N
s 5
f
)
5 j_i s
s e
6 N
j e
5 e
Nf5_
4
~
2 o m5 l
4 z
i h'
E
,M s,
I j
l l ]
l I
yt gM(
[F l
l l
r l
f-y o
o "l
v s
c e
e j
l 7
y,
-l 6
82 l
a "
rr
{
v a
l l
1 5
4l f,
3
~
0 n1 2 ~j o
3__
o 1
l J{
u3 jI.
O t
g C
s g-E mU gf oI R
I G
q l
n F o'
y1 mf u
N_ k ]sC
~I I1 i
e s
1 3
~
1 l
l f, 0-t O
5
}
s i
e:>
7 l
5 li
- ]*
.g N
0 I
l-l l
U
" u, lr a
o I'
i l
1 i
'r lI Ig
~
1
(
'I
[
1
- oS tG o, c L E
, oE 1
0 i
0 5
8T 1
I N.$ d 0
4 l
3
. ~
s
$I@ j. 9 X*I l
im 5
5 7
5 7
7 1
l l34 Nt g
[-
5 L.7' l
I 2
3 2 5
0 4
o 5
5 9 i irf lll
_y 8t
?
- z 15 9
D M c E o
(
2 o
g. *C B g
/
ag a
G l
8 7
v g
.o k
h o
am i?i e
r o
c4 od
~O*09 QTC 0n 0*95 0'g O'O
- - ~ ~
~.
~ _ -
f~
8 M
s b
- f 8
^
o d
4
\\
2 5
.Co a
x o 88 g 's m
e oo 8 iE d 50
^
g 2 9 eb 8
m a
7 8
.itit x
a w
o W $
0 2
3 g
j P
In
$ i 9 3 *B
\\
m o 4
4 3
Q 3
~M m
\\,-
'f s
O
-d 4
0
's
..g_
u
.6 oon oon oon.
o cos o'045 cm pngy;;g en rmmm m e_ L 2
0 1
0 8
N M
03 O
R S
E T
W C
P.
O A
M
/
00 0
P P
E
'00 6
T 00
'o
% O G
00 t
0 41 9
V E
.f 00 h
L R
05 C
W O D 17 E
F P
L 22 S
~
D
(
O K C
.L 0
PP E
E 6
D MM M
I S
1 N
EE T
S A
TT E
N a
B O
T 0
O I
S T
A H
4 IV L
B A
U D
M IS 9
E N
U J
0 t
2
~
y k
J, 0
,I O
.OC a$
a o R*
oO3 ow8 k $
E o.P[
- + = = * * " * "
- W O
-e 25 on C
h b k
[,,, i
-e n
64 8
]
T a
h 7
N O
-d.$ d p
d i $
N to O x
\\
o h b l
28 l
m*
W 23 l
1 d
o i
4 I
-d g
5 i
6 i
V
{
l C
1 o
i
\\
0'000%
0'00PT 0'000E 0'0096 0'00Z1 0'003 l
o.
o._s-------..
~
C ES 6
96 R OT E
F S
WT L OC V P A L
%P O
~
I 2 O L 0
.D 1
P E WiSH P FON A L
L L O L
E? O
(
C C
i S RT E E
R R B
O S
C IV T
A O
D T
\\
m,-
e
';ii'i
',,I
,,,,,'i',,l 9E Og 0 *e O.n ee-e 2'3g Yd ba aCBEzo g C.
<~
04 0
0053 C
ES 0
0 6
0 9
0 6
3 R O E
S W T P 0
O C M 0 0 0
P A E 0 0 50 T
% P 0 0 2
2 O G 0 0 E
4 1 0
L P
0 0
)
1 D
0 5 C
W H L 7
- 0. E i
1 1 1 0S F
O 0(
O N FF C
0 L O PP 2 E D
MM M
E P N
EE T
I S
I S R A
.TT E T ao R T 0
B O
0 S
H 0
e c 5
IV A
1 AD 0
0 0
0 1
r ro 0
t w
e 0
n F
R 0
i T
5 3
f 4
Wi '
J n
i i i
' i
,t z
r 0
a<
0
\\
Od=
- o. =
- 8*
2*a O,
Coavu. h "
n 3o nLL l
e 4
y d
I o
a o
-R
/
m Ow O
M 1
Wa o
43 9
C f5*
!d a-e o
O 2
oo o
g Oo S
A 4
g a
oo a
N O
O 00 L'
O o
-s a w.
-I J
NN OM 7
O ov b b O
kk EW O
22 2
w g
$ E k dd e
oo e
w r,3 o
a
~S I
co l
m I
~
4 C
8s I
.e o
{
E{
d 8
g k
o 0'099 0*0r9 0'009 0*099 0'0ZS O*097 m
(3030) 3dN31 -
a FIGURE C.3.1. 8 Rev 7
]
I
?
~
o S
S o
I Q
l 0
w M
O e
_8 CD
/
o W
f.
M i
[\\
l 0
5 3:
10 o
O O
/
d A
4 E
w N
1 m
cN O
O L' $
6 w
ow 7
C I g om d
Z J
O O
EW
.h '
W 5 s m
O w
g o
i G3 1
f
-l E
Q O
o o
o I
y -- *g l
I E
IP p
qo g
O'009Z O*00FE 0'000Z 0*0091 0'00E1 0'009 (ENI/J81) 0000t004 d l
I FIGURE C.3.1.3 Rev 7
\\
y..
- ;h $,
.Dackst No. 50-346
~'
l Liccuss No.-NPF-3--
I1 Serial No.'1241 Atcachment. 6 '.
NRC STAFF RECOMMENDATIONS SER.
TED Response Recomt:endation Page No.(s) in Attachment 1
^ Remove the'AFW Common Suction.
3-45 Section 3.2.1.11 Strainer.
3-63 Section 3.3.1.2 Formation of Independent Safety 3-5 Section 3.1.1 Engineering Group i