ML20050A516
| ML20050A516 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 03/25/1982 |
| From: | Davidson D CLEVELAND ELECTRIC ILLUMINATING CO. |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| TASK-2.K.1, TASK-2.K.3.25, TASK-2.K.3.31, TASK-2.K.3.45, TASK-TM NUDOCS 8204010364 | |
| Download: ML20050A516 (12) | |
Text
Y e
T H E C L E V E L A N D E L E C T R ! C i! L U f,il l1 AT I N G C O M P A t! Y P O BOX f/X)O e CLEVELAND. OHlo /.4101 e TELEPHONE (216) 622-9800 e ILLUMINATING BLDG e 55 PUBLICSOUARE crving The Best Location in the Nation Daiwyn R. Davidson VICE PRESID(NT
$7Sitb ENGINECTUNG AND CONSTRUCTION March 25, 1982 Mr. A. Schwencer Chief, Licensing Branch No. 2 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Perry Nuclear Power Plant Docket Hos. 50-M 0; 50-M1 Response to Draft SER Reactor Systems Branch Dce.r Mr. Schwencer:
This letter and its attachment is submitted to provide draft responses to the concerns identified in the Draft SER for Reactor Systems.
It is our intention to incorporate these responses in a subsequent amendment to our Final Safety Analysis Report.
Very Truly Yours,
/
/
iN s
Dalwyn Davidson Vice President System Engineering and Construction DRD: mlb cc:
Jay Silberg John Stefano to g
Max Gildner P
A DEccp;go
~
e7 l'?/lR 3.7 79826 tus=m cen
(-
D 2
1
+
p0t
,,3 s
$5??o???oyggggg9 PDR
0 i
l Item No. II.K.1.5 l
Safety-Related Valve Position REQUIREMENT Review all valve positions, positioning requirements, positive controls, and related test and maintenance procedures to ensure proper ESF functioning.
RESPONSE
Perry Nuclear Power Plant is equipped with valve position status monitoring that satisfies the J:qairements of Regulatory Guide 1.47 as discussed in FSAR Section 7.1.
Perry Plant procedures for tagging, maintenance, and surveillance will assure verification of valve position status on the affected portions of system to verify ESF systems are functional after the performance of sur-veillance tests, and maintenance activities.
These plant procedures will be available for review by Region III Division of Inspection and Enforcement, approximately six months prior to fuel load.
i Item No. II.K.l.lO Safety-Related System Operability Status Assurance REQUIREMENT Review and modify, as required, procedures for removing safety-related systems from service (and restoring to service) to ensure that operability status is known.
RESPONSE
Perry Plant procedures for removing safety-related systems from service and restoring to service will assure the operability status is known.
Release of all ESF equipment from service will require Shift Supervisor's approval.
Plant procedures will include verification of operability.of safety-related equipment after restoration following surveillance and maintenance activities.
These procedures will be available for review by Region III Division of Inspection and Enforcement, approximately six months prior to fuel load.
d 4
l
.=
l Item No. II.K.3.25 Effect of Loss,of Alternating-Current Power on Pump Seals REQUIREMENT The licensees should determine, on a plant-specific basis,-by analysis cnr experiment, the consequences of a loss of cooling water to the reactor recirculation pump seal coolers.
The pump seals should be designed to withstand a complete loss of alternating current (ac) power for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Adequacy of the seal design should be demonstrated.
RESPONSE
The Cleveland Electric Illuminating Company has participated in the BWR Owners' Group evaluation of the effect of the loss of pump seal cooling for a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
This evaluation was submitted in a letter from D. B. Waters to D. G. Eisenhut, dated May 1981.
The study indicates that the loss of pump seal cooling for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is not a safety problem, but may require seal repairs prior to resuming operating. - Even in the case of both seal cooling systems failing, followed by extreme degracation of the pump seals, the primary coolant loss is analyzed to be less than 70 gallons per minute.
Consequently, no hazard to the health and safety of the public will result from total loss of recirculation pump seal cooling water.
In addition, a supplement of the BWR Owner's Group evaluation was submitted in a letter from T.
J. Dente to D.
J.
Eisenhut dated September 21, 1981.
This supplement describes three tests per-formed on Representative BWR reactor recirculation pumps in which all seal cooling water was loot.
The test results show that pump seal leakage is acceptably low following a loss of seal cooling water for as long as two hours.
These test results are representative and bounding for 'he' Byron Jackson reactor re -
circulation pumps utilized at Perry.
I s
.I l
Item No. II.K.3.31 Plant-Specific Calculations to Show Compliance with 10CFR Part 50.46 REQUIREMENT Plant-specific calculations using NRC-approved models for small-break loss-of-coolant accidents (LOCAs) as described in Item II.K.3.30 to show compliance with 10CFR 50.46 should be submitted for NRC approval by all licensees.
RESPONSE
The results of a typical LOCA analysis have been provided in FSAR Section 6.3.3.
This analysis uses the currently approved Appendix K methodology.
A Perry Plant specific analysis using NRC approved models will be submitted in April 1982.
If any model changes are required by the NRC as a result of Item II.K.3.30, an evaluation will be made of the impact of the change on the Perry plant specific analysi s, to determine the need for any additional analysis.
l
1 Item No. II.K.3.45 Evaluation of Depressurization with Other Than ADS REQUIREMENT Analyses to support depressurization modes other than full actuation of the automatic depressurization system (ADS)
[e.g.,
early blowdown with one or two safety relief valves (SRVs)) should be provided.
Slower depressurization would reduce the possibility of exceeding vessel integrity limits by rapid cooldown.
RESPONSE
The Cleveland Electric Illuminating Company participated in the BWR Owners' Group generic evaluation of alternate modes of de-pressurization other than full actuation of the ADS.
The results of this program were submitted to the NRC in a letter from D.
B. Waters to D. G.
Eisenhut dated December 29, 1980.
The BWR Owners' Group evaluation showed that vessel integrity limits are not exceeded for full blowdown, and slower depressurization rates have little benefit to vessel fatigue.
LR2 1. Foii. ion Fa:er 1/25/E2 4-RSB OPERATOR ACTION RE0VIRED 10-20 MINUTES FOLLOWING A LOCA ISSUE:
y
(
The LRG II participants must clearly state and justify all operator actions assumed in the LOCA analysis in the 10-20 minute time frame following a LOCA.
LRG II RESPONSE:
No operator actions will be r? quired, before 20 minutes following the LOCA, to maintain the safety of the LRG II Plants.
The original BWR/6 design would have required the operator to initiate ADS at 10 minutes after steam line break outside of containment to ensure adequata core cooling.
ADS logic modifications proposed by the BWR Owners' Group in response to TMI Action Plan Item II.K.3.18 will eliminate the need for this action.
LRG II plants will implement a modification acceptable to the Staff.
. Operator actions assumed in the ECCS analysis are not required actions c
but are actions that make the analysis more conservative.
The FSAR core cooling analysis assumes the operator diverts ECCS to containment cooling
.at 10 minutes; this assumption was made to add conservatism to the core cooling calculation.
Diversion at 10 minutes is not needed for either containment or core cooling. No operator action is required to adequately cool the containment until 30 minutes after a LOCA.
MJA: rf/11SA25
d 421.41 The statement in Section 7.3.1.1.1.1 that the HPCS provides makeup water to the reactor until the vessel water level reaches the high level (trip level 8) conflicts with the statement in,
Section 6.3.2.2.1 regarding the HPCS system.
Please indicate which discussion-is correct.
Response-Current design includes a high drywell pressure interlock as discussed in Section 6.3.2.2.1.
There is a design change in progress to remove the 'high drywell pressure interlock for the HPCS trip. The FSAR will be modified to reflect this change when the design'has been finalized.
9 l,
i Vd,
L $a i*
^ '
~ -
r.. c ua x..,,
q _. m ;c.u_,.-
b w
i
~
421.23 From the discussions provided in Sections 6.3.2.2.3, 6.3.2.2.4, 7.3.1.1.1.3, and 7.3.1.1.1.4, it is not clear whether or not the LPCS and LPCI injection valves are interlocked to prevent them from opening unless reactor pressure is low enough for injection to be possible. Provide more information concerning the operation of these valves. Also, there are discrepancies in the FSAR as to whether differential or gage (absolute) pressure transmitters are used for the interlocks. For example, Section 7.3.1.1.1.3 and 7.3.1.1.1.4 imply differential pressure transmitters are used.
However, the P&I diagram for the LPCS system, Figure 6.3-8, does not show a differential pressure transmitter near the injection valve.
Response
The present Perry design of the high pressure / low pressure interface of the low pressure ECC lines is illustrated in Figure 5.4-13, Sheet 2.
The pressure interlock (a pressure transmitter / trip' unit between the testable check valve and the motorized injection valve) is designed to be functional during test opening of the MOV.
Automatic initiation of the low pressure ECCS by the LOCA signal will bypass the interlock and immediately open the M0V'S.
The differential pressure transmitters mentioned on pages 7.3-8 and 7.3-10 were not used in the Perry design and the FSAR. Revised pages are attached to reflect current design.
Reactor vessel water level (Trip Level 1) is monitored by two redundant level trans'mitters. Drywell pressure is monitored by two redundant pressure transmitters. The vessel level trip unit relay contacts and the drywell pressure trip unit relay contacts are connected in a one-out-of-two twice logic arrangement so that no single instrument failure can prevent initiation of LPCS.
4 The LPCS components respond to an automatic initiation signal simultaneously (or sequentially as noted) as follows:
1.
The Division 1 diesel generator is signaled to start.
~
2.
The norma 11y closed test return line to the suppression pool valve t10F012 is signaled closed.
3.
When power (offsite or onsite) is available at the LPCS pump motor bus, the LPCS pump is signaled to start.
2 N
I.
s' 4.
The LPCS pump discharge flow is monitored by a differential pressure transmitter. When the pump is running and discharge flow is low enough to cause pump overheating to occur, the minimum flow return line valve M0F011 is opened.
The valve is automatically clos,ed if flow is normal.
The LPCS pump suction from the suppression pool valve M0F001 is normally open, the control switch is keylocked in the open position, and thus requires no automatic open signal for system initiation.
TheLPCSpumpanbinjectionvalveareprovidedwithmanualoverride controls. These controls permit the operator to manually control the system subsequent to automatic initiation.
I 7.3-8
The Division 1 LPCI (Loop A) receives its initiation signal from the LPCS logic.
The.LPCI system components respond to an automatic initiation signal simultaneously (or sequentially as noted) as follows (the loop A components are controlled from the Division 1 logic; the loop B and C components are controlled from the Divisica 2 logic):
The Division 2 diesel generator is signaled to start from the loop B 1.
and C initiation logic.
When the offsite power or the diesel generators are providing power to 2.
This is the pump motor buses, sequential loading is provided.
accomplished by delaying the start of LPCI pumps A and B by 5 seconds while allowing the IPCS and LPCI pumps to start immediately.
n 3.
The following normally closed valves are signaled closed to ensure
[
[
proper system lineup:
l The RHR heat exchanger discharge to RCIC valves M0F026 A, B, and (a)
AOF065 AB.
(b) The RER heat exchanger flush to suppression pool valves M0F011 A, B.
l 1
l l
7.3-10 i
d
-a.
I i
Issue RSB-15 AWS procedures are to be submitted by the applicant for staff review and approval.
-Response to Issue RSB-15 The preparation of an Emergency Procedure for the AWS event shall be in accordance with the General Electric generic Emergency Procedure Guidelines and the BWR Owner's Group. This procedure shall be completed about six months prior to fuel load and submitted for NRC review.
j i
5 i.
Ic:4Isc/c/pjn
+
- -....