ML19326C779
| ML19326C779 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 08/14/1974 |
| From: | Bernero R Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8004280714 | |
| Download: ML19326C779 (36) | |
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THIS DOCUMENT CONTAINS AUG 1 s u74 P0OR QUALITY PAGES DOCKET NO.: 50-313 APPLICK3T : ARKANSAS PO'42R AND LIGHT COMPMTY (APGL)
FACILITY : EriSAS NUCLEMt ONE, UNIT 1 REPORT OF MEETING AT DETHESDA, MARYL\\ND ON JULY 30, 1974 Introduction This meeting was held to discuss details of the following outstanding electrical, instrumentation, and control natters on Arkanoas Nuclear One, Unit 1 (ANO-1); these matters are undergoing specific post-licensing review:
1.
Reactor Protection System Modification I
2.
Emergency Feedvater System Control 1
3.
Steam Line Dreak Instrumentation & Control System (SLBIC) i The applicant responded to all questions raised at this meeting. The i
applicant informed the staff that there will be additional delay in the l
delivery of a pair of key components for the SLBIC, delaying completion of tha system at least until the end of 1974.
Discussion
,I 1.
Reactor Protection System Modification There was some general discussion of this rodification and discussion of drawing details. This modification consists of some minor circuit changes to produce an automatic change of the overpower trip setpoint from 105.5%
to 57. whenever the shutdown bypass mode is selected _ Previously, the Reactor Technician had to adjust the tirip setpoint manually each time.
The applicant presented a handout (Attachment 2) which summari::es the principal features of thia modification.
2.
Emergency Feedwater System Control There was brief discussion of this matter which involves the acceptability of isolating devices which separate a non-safety grado automatic control system from safety grade motor operated valves which must remain manually operable during emergency conditions. AP&L clarified some of the explana-tions given in the April 29, 1974 letter on this subject.
l arrec s w
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- omra w Forma AEC 313 (Rev. 9-$D AECM 0240 W u. s. oovenmusur pontino orrics s e74.sas.s eo 8004280 M
/..
. 3.
Steam Line Break Instrumentation & Control System (SLBIC)
The applicant gave a detailed presentation reviewing the design basis and details of the system. A 30-page handout, including copies of the slides and pertinent electrical schematics, was presented; a copy of this handout is attached to the docket file copy of this meeting summary. The presenta-i tion covered the safety aspects of the system and included an itemized review of how the applicant feels the system satisfies the specific requirements of applicable design standards (IEEE-279, etc.).
The following
~
items were noted in particular:
a.
The valvos controlled by the SLBIC, the nain steam block valves (air open-spring shut), and the main feedwater block valves (motor-operated), can be cycled through partial stroke for surveillance testing while the plant is operating.
1 b.
The individual sensors and logic trains are testabic while the plant is operating.
c.
All SLBIC equipment is seismic Category I.
d.
The SLDIC cabinets are located in the Elcetrical Equipment Room; the pressure signal is brought to that roon by small diacet--
(3/8 inch 0.D.) steam lino.
)
e.
The pressure switches cloce contacts on decreasing pressure.
f.
The SL3IC uses General Elcetric HrA relays. Scismic tent data 1
(5-33 rtz) is available for the relays, but the SLBIC cabinets are computer analyzed for scianic response.
1 g.
The SL3IC pancis have been built; the critical component delivery is the tuo new, high-speed (20-second stroke) operators for the main feedwater block valves whose delivery date has slipped to December 1, 1974.
Subject to AEC raviev and acceptance the SLBIC can be installed quickly except for the feedwater valve operators. The existing operators en those valves are slow acting (90-second stroke).
AP&L and 3&W will datormine whether a steam line break calculation can be made to establish how far into core life the plant could operate with reliance on such slow-actinF valves..They agreed to notify the staff within a week of the results of this determination.
l
2.
3-Conclusion The~ staff noted to the applicant that some of the documentation of these matters appears inadequate, an FSAR acendment to cover them appears to l
be appropriate. The schedule for installation of the SLBIC will bc determined along with the conclusion of the staff review.
I ortsinal sused W Robert M. Bernero, Project Manager i
Light Water Reactors Branch 2-3 Directorate of Licensing Attachments:
1.
List of' Attendees 2.
F.eactor Protection System DISTRIBUTION:
Docket Files AEC PDR LPDR L Reading LWR 2-3 Reading AGiambusso RSBoyd RP ads RP BCs SVarga DEisenhut.
3 RKlecker i
FSchroeder TR ads 4
RS (3) l
-OcC 1
RMBernero EGoulbourne ACRS (16) l RSchool FAnderson i
x7886/ LWR 2-3 o,,,c.,
pg RMBerhero:cj,b
- e.....,.
8/f,)/74 o.v.,
Fena AEC.)l3 (Rev. 9 5)) AECM 0240 W u. s. oovsammsnr,mmtime o,ricas se,..eae.see
4 ATTACHMENT l~
MEETING WITH APJJWSAS POWER AND LIGHT COMPANY ARKANSAS NUCLEAR ONE, LTIT 1
-HELD JULY 30, 1974 LIST OF ATTENDEES Atomic Energy Commission R. Bernero R.-Scholl' F. Anderson Arkansas Power & Light-Company
.W. Cavanaugh D. Rueter J..Grisham D. Admas D. Smither Babcock-& Wilcox E. Willingham, Jr.
R. Williamson H. Baker E. Patterson Bechtel Corporation E. Smith-W.
!ehegan K. Bailey G. Smith
.)
TTACIC!E:1T 2 Reactor Protection System (RPS) s I.
Subject:
Shutdown Bypass Power Level Trip Bistable II.
References:
- 1) Arkansas Nuclear One, Unit 1, Final Safety Analysis Report 2)
Babcock & Wilcox Topical Report SAW-lC003, Revision 2, Qualification Testing of Protection System Instrumentation III.
Discussion A.
Present Shutdown Bypass High Power Level Trio Bistable 1.
There are four high power level trip bistables, one in each of tne RPS's four protection channels as shown in Raference l's Figure Number 7-1.
Each high power level trip Distable 4 set to trip its protection channel whenever the power level reaches or excceds the trip setpoint (nominally 1105.5% of full power as noted in Reference 1, Section 15's Table 2.3-1) during plant operation.
2.
There are four shutdown bypass circuits, one in each of the RPS's four protection channels as shown in Reference l's Figure Number 7-l.
As discussed in Reference l's Section 15.2.3 F., whenever the shutdown bypass circuit is used tvio conditicas are imposec:
"1.
By administrative control the nuclear overpower trip set point must be reduced to a value 1 5.0 percent of rated power during reactor shutdown.
2.
A high reactor coolant system pressure trip set point of 1720 psig is automatically imposed."
Condition 1 is met by manually adjusting the four high power level trip b'istables, discussed in 1. above, from their normal trip set-point (< 105.5%) to the shutdown bypass trip set:oint (1 5.0%).
Condition 2 is met by having four high reactor ccolant pressure trip bistables, one in each of the RPS's four protection channel shutdown bypass circuits, previously set to the shutdown bypass trip setpoint (1 1720 psig).
3.
When the shutdown bypass circuit is returned to normal, tne four high power level trip bistables, discussed in 1. and 2. above, must be manually readjusted from the sL tccwn cy ass trip set-point (15.07,) back to their normal trio setpoin- (1 105.57.) in i
order to return to full pcwer operation.
B.
Proposed Shutdown Bypass High Power Level Trip Bistatie 1.
Four additional high power level trip bistables.could be added to the present four for a total of eight, two ir each of the RPS's
~a
,l four protection channels.
The first, and present, high power level trip bistable would be set to trip its protection channel whenever tne power level reaches or exceeds the trip setpoint (nominally
.1 105.5% of full power as noted in Reference 1, Section 15's Table
'2.3-1) during plant operation as it presently is.
The second, and addedshigh power level trip bistable would bein the shutdown oypass circuit and would be set to trip its protection channel whenever the power level reaches or exceeds the trip setpoint (nominally 1 5.0L of full power as noted in Reference 1, Section 15's Table 2.3-1) during plant shutdown (shutdown bypass circuit actuated).
2.
Conditions 1 and 2 of Reference l's Section 15.2.3 F would be ret by having the four sets of high reactor coolant pressure and hign powsc level trip bistables, one set (pressure and power) in each of the RPS's four protection channel shutdown bypass circuits, previously set to the shutdown bypass trip setpoints (11720 psig and 1 5.0%).
3.
When the shutdown bypass circuit is returned to normal, no manual trip setpoint adjustments are required in order to return to opera-tion.
C.
Summary of How the Proposed Modification Meets IEEE-279 See Reference l's Section 15.2.3 F.
D.
Design Bases of the Modification See Reference l's Section 15.2.3 F.
E.
Design Bases of the RPS See Reference 1; specifically, Sections 7.1.1, 7.1. 2 and 7.3.1.1 for design basis information.
See Reference 2 for qualification testing _information.
IV.
Summary There is one high power level trip bistable in each protection channel of the RPS presently.
It has to be adjusted to a lower trip setpoint when the shutdown bypass is used.
It then has to be readjusted to the normal (higher) trip setpoint when normal operation is desired.
There would be two high power level trip bistables in each protection channel of the RPS as proposed.
The first (presently installea) would be adjusted to the normal trip setpoint.
The second (proposed to be installed) would be adjusted to the shutdown bypass trip setpoint.
The first bistable (normal trip setpoint) would always produce a protection channel trip re-gardless of the presence or absence of the shutdown bypass actuation.
The second bistable'(shutdown bypass trip setpoint) would only produce a pro-tection channel trip in-the presence of the shutdown bypass actuation.
As' proposed, no adjustments need to be made to any trip setpoints when switching to or from shutdown bypass.
l
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