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- 75 NUCLEAR REGULATORY COMMisslON 1% E WASHINGTON. D. C. 20555 5.iM Mu. l [
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AN 0 2 1984 Docket No.:
50-416 MEliORANDUM FOR:
Thomas M. Novak, Assistant Director for Licensing Division of Licensing 3
FROM:
R. Wayne Houston,' Assistant Director for Reactor Safety Division of Systems Integration t
SUBJECT:
GRAND GULF TECHNICAL SPECIFICATION CHANGES Plant Name:
Grand Gulf Nuclear Station Unit 1 Docket No.:
50-416 Licensing Stage:
OL Responsible Branch: Licensing Branch #4 Project Manager:
M. D. Houston l
DSI Review Branch:
Reactor Systems Branch TACS Nos.:
54329, 54335, 54318, 54332, 54478 Enclosed is the Reactor Systems Branch safety evaluation report on technicalspecificationchangesproposedbythelicenseeforGrandGy]lf Unit 1.
The proposed changes for TACS 54329, 54335, 54318' and 54478 are acceptable.
The change proposed with TACS 54332 must be modified as discussed in the enclosure.
(/V W
R. Wayne Houston, Assistant Director for Reactor Safety Division of Systems Integration
Enclosure:
As stated cc:
R. Mattson D. Houston L. Kintner D.[
CONTACT:
T. Collins X24478
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,L Technical Specification 3.3.3, Tabl'e 3.3.3-3 " Emergency Core Cooling System Actuation Instrumentation Setooints" Item b.2.e, LPCI Pump B&C Discharge Pressure - High, I',^CS 54335 K Mt D N y s vu n-Q l
.Ih: 1ker.seehe5p.vgusedch=ghgNeLPCI'BandCPumpDischarge 4 cA -p/
Pres sure-High,gfrom greater than or equal to 122 psig, increasing, to 115-135 psig, increasing.
1,u, Safety Evaluation"
.We-purp0:e of Ee lower setpoint limit on this trip h t; provide--
et assuranee that a low pressure ECCS pump is operating prior to ADS actuation. The u per setpoint limit assures that ADS is not prevented from. initiating due to a setpoint higher than pump capability.
Pump characteristic curves in the FSAR indicate that the 115-135 psig range is within the design limits of the LPCI pumps.
i The pf'o,'p;(sIlf change does not increase the probability or consequences of an accident previously evaluated nor does it create the possibility of a r.ew or different accident from those previously evaluated.
It does not cer.stitute a significant hazards consideratien.
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Technical Specifications 6.9.1.5 " Annual Reports" and 6.9.1.12
" Prompt otification with Written Followuo'i (T^.CS 543297CL
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!d Item II.B.3.3 of NUREG-0666, the THI Task Action Plan, required that safety / relief valve failure be reported promptly and challenges be reported annually, GGNS committed to meeting these requirements in FSAR Section 18.1.30.1.
- The pgp'ohe changes incorporate these commitments into the technical specifications.
Saf.ety Evaluation The'psosqset changes are. administrative in nature.
They do not cause a
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.significant increase in the probability or consequences of an accident previously evaluated nor do they create the possibility of a new or
.different accident from those previously evaluated. The profds1!Il changes are therefore acceptable.
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Technical Specification Table 3.3.2-3 Item 5.a, "RCIC Steam Line Flow-High." (TACS 54332)
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The licensee has proposed a change to the footnote on " Isolated System
. Response Time" for RCIC St'eam Line Flow-High.
The change would note i
- that.a time delay of 3-7 seconds is included in the response time l
i listed. The licensee has proposed this change for compliance with GenericLetter83-02(GL-83-02), Item 9.
i Safety Evaluation l
The change proposed by the licensee is not sufficient to" satisfy the
~ requirement of GL-83-02, Item 9.
The specification proposed would be satisfied by any isolation which takes less than 13 seconds to complete, and'hence makes no provision to assure that the time delay is operiible.
Attached are. sample specifications which will satisfy the reouirement
. for the time delay relay at Grand Gul.
Implementation of the sample specifications is acceptable to us for closing out GL-83-02, Item 9.
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TAELE 3.2.2-3 (0:ntinuec)
ISOLATIONSYSTEMINSTRUMENTATIONRNSPONSETIME TRIP FUNCTION RESPONSE TIME (Seconds)#
5.
REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION a.
RCIC Steam L.ine Flow - High
< 13(*)
? c.
, RCIC Steam Supply Pressure - Low.
513(*)
b.
RCIC Turbine Exhaust Diaphragm Pressure - High NA t
d.
RCIC Equipment Room Ambient Temperature - High NA RCIC Equipment Room a Temp. - High-NA e.
f.
Main Steam Line Tunnel Ambient Temp.
. igh NA H
g.
Main Steam Line Tunnel A Temp. - High NA h.
Main Steam Line Tunnel Temperature Timer NA 1.
RHR Equipment Room Ambient Temperature - High NA j.
RHR Equipment Room A Temp. - High NA k.
RHR/RCIC Steam Line Flow - High NA 1.
Manual Initiation NA m.
Drywell Pressure - High (ECCS Division 1
_ 13(,)
and Division 2)
RHR Equipment Room A Temp. - High NA b.
c.
Reactor Vessel Water Level - Low, Level 3
< 13(,)
d.
Reactor Vessel (RHR Cut-in Permissive)
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Pressure - High NA e.
Drywell Pressure - High NA f.
Manual Initiation NA (a)'The isolation system instrumentation response time shall be measured and
. recorded as a part of the ISOLATION SYSTEM RESPONSE TIME.
Isolation system instrumentation response time specified includes the delay for diesel generator starting assumed in the accident analysis.
(b) Radiation detectors are exempt from response time testing.
Response time shall be measured frem detector output or t'he input of the first ele:tronic component in the channel.
" Isolation sy' stem instrumentation response time for MSIVs only.
No diesel generator delays assumed.
"" Isolation system instrumentation response time for associated valve's except MSIVs.
- solation system instrumentatien response time specified fer the frio l
% ncti:n actuating each valve group shall be adcec tc isclatien ti a sne.n in Taoies 2.6.4 *. anc 3.6.5.2-1 fer valves in ea:n valve gr:cp :: ::.ain ISOLATION SYSTEM RESPONSE TIME for ea:n valve.
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