ML20073D436
| ML20073D436 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 04/18/1991 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20073D429 | List: |
| References | |
| NUDOCS 9104290064 | |
| Download: ML20073D436 (2) | |
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UNITED STATES 4
NUCLEAR RE3ULATORY COMMISSION e
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WAEHtNoToN O, C. 20666 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENOMENT NO. 56 TO FACILITY OPERATING LICENSE NO. NPF-47 GULF STATES UTILITIES COMPV RIVER BEND STATION, UNIT,
DOCKET NO. 50-458 1.0- INTRODUCTION By letter dated March 1, 1991, Gulf States Utilities Company (GSU) (the licensee) requested an amendment to Facility Operating License No. NPF-47 for the River Bend Station (RBS), Unit 1.
The proposed amendment would revise Technical Specification (TS) Table 3.3.2-1, " Isolation Actuation Instrumentation,"
to correctly identify actuation of the emergency mode of the main control room area ventilation system (MCRAVS) at reactor vessel water low, low level 2
(-43 inches) instead of low, low, low level 1 (-143 inches), _which is currently reflected in the TS table.
By letter dated February 28, 1991, GSU discussed the plant modification to be_made during the forced outage beginning on February 27, 1991.
The control circuitry for the charcoal filter start logic was modified so that the MCRAVS will start on reactor water low, low level 2.
The proposed TS amendment will revise the TSs to reflect this modification.
2.0 EVALUATION Tha-MCRAVS is part of the control building air. conditioning system and an-engineered safety feature.
The MCRAVS consists of two full capacity redundant air handling units, each with a charcoal filter train and dampers.
To protect y
contro1Lroom personnel against airborne radiation during accident conditions the MCRAVS automatically closes.the dampers and diverts the exhaust air through
' the charcoal filters.
This occurs on high drywell' pressure, low reactor water level, cF high-radiation in-the local air intake.
The plant was: originally designed and TSs we'e written for initiation of the r
MCRAVS in the emergency. mode at low, low, low level 1 (-143 inches).
- However, on February 12, 1991, GSU discovered calculations for a main steam line break assumed the MCRAVS received a-start signal at reactor vessel water low, low' 1
level 2 (-43' inches).
Because-of the conflicting information, RBS placed one charcoal-filter train in continuous service to bypass the automatic function of the MCRAVS and' performed a review of Updated Safety Analysis Report (USAR)
' Sections 7.3.1.1.9 and 7.3.2, the TSs, and radiological dose calculations to determine the' correct initiation level.
RBS determined that if the MCRAVS initiated;on reactor water low, low, low level 1,. the initiation would occur too_ late and the, regulatory doce limit to the thyroid would be exceeded for co_ntrol_ room personnel.
Therefore, RBS concluded the MCRAVS should start on a
' low, low 11evel 2 reactor water level and a plant modification during shutdown twould be:needed.
On February 27,-1991,-RBS entered a forced outage to repair a
~ recirculation pump seal and made plans to modify the control circuitry for the MCRAVS.
By letter dated February 28, 1991, GSU discussed the modification and resulting TS change.
.9104290064 910418 PDR =ADOCK 05000458
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. Because the plant modification resulted in an initiation signal for the MCRAVS
]
which was more conservative than current TSs, the modification was performed without an immediate change to the TSs.
The emergency mode of the MCRAVS now actuates on low, low level 2 reactor water level.
Licensee Event Report (LER)91-001, dated March 14, 1991, discussed why the discrepancy occurred, provided GSU's design analyses, described the control of design documentation, discussed the root cause analysis, corrective action and GSU's safety assessment.
GSU concluded that no other control room
< habitability analyses or LOCA related accident analyses were impacted by the low, low, low level 1/ low, low level 2 discrepancy.
Additionally, as a result of GSU's investigation of the discrepancy, a number of other discrepancies were identified and resolved.
Based on the staff's review of the l w ensee's application and LER, and the applicable sections of the USAR, Saiecy Evaluation Report (SER), and Standard Review Plan (SRP), the plant modification and proposed TS changes are acceptable.
- 3. 0 STATE CONSULTATION in accordance with the Commission's regulations, the Louisiana State official was notified of the proposed issuance of the amendment.
The State official had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no si increase in individual or cumulative occupational ri.diation exposure. gnificant The Commission has previously issued a proposed finding thGt the amendment involves' no significant hazards consideration, and there has been no public comment on such finding (56 FR 10582).
Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
- 5. 0 CONCLUSION The Commission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contrioutor:
Claudia M. Abbate, PDIV-2 Date:
April 18, 1991 l
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INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.2.1 Each isolation actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.2.1-1, 4.3.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.*
4.3.2.3 The ISOLATION SYSTEM RESPONSE TIME of each isolation trip function shown in Table 3.3.2-3 shall be demonstrated to be within its limit at least once per 18 months.
Each test shall include at least one channel per trip system such that all channels are tested at least once every N times 18 months, where N is the total number of redundant channels in a specific isolation trip system.
l l
- Logic System Functional Testing period may be extended as identified by notes C and D on Table 4.3.2.1-1.
RIVER BEND - UNIT 1 3/4 3-11 Amendment No. 8
11 TABLE 3.3.2-1
- o -
h ISOLATION ACTUATION INSTRUMENTATION h
VALVE GROUPS MINIMUM APPLICABLE OPERATED BY OPERABLE CHANNELS.
OPERATIONAL o
TRIP FUNCTION SIGNAL *** PER TRIP SYSTEM (a)
__ CONDITION ACTION b
1.
PRIMARY CONTAINMENT ISOLATION a.
Reactor Vessel Water Level-1,7,8,9(b)(c)(j)
[
Low Low, Level 2 15, 16 2
1,2,3 20 b.
Drywell Pressure - High 1,'3,8(b)(c)(j) 2 1,2,3 20 c.
Containment Purge 8
1 1,2,3 21 Isolation Radiation -
High r
2.
MAIN STEAM LINE ISOLATION R
a.
Reactor Vessel Water Level-Low Low Low, Level 1 6
2 1,2,3 20 3
b.
Main Steam Line Radiation - High 6,9(d) 2 1,2,3 23 c.
Main Steam Line Pressure - Low 6
2 1
24 d.
Main Steam Line Flow - High 6
2/MSL 1,2,3 23 e.
Condenser Vacuum - Low 6
2 1, 2**, 3**
23 f.
Main Steam Line Tunnel Temperature - High 6
2 1,2,3 23 g
g.
Main Steam Line Tunnel g
A Temperature - High 6
2 1,2,3 23 h.
[
Area Temperature
.o High (Turbine Building) 6 2/ area 1,2,3 23 8
~
TABLE 3.3.2-1 (Continued)
ISOLATION ACTUATION INSTRUMENTATION l
VALVE GROUPS MINIMUM APPLICABLE l
6 OPERATED BY OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION SIGNAL *** PER TRIP SYSTEM (a)
CONDITION ACTION 1
i 3.
SECONDARY CONTAINMENT ISOLATION a.
Reactor Vessel Water Level-Low Low, level 2 11, 12, 2
1,2,3 25 13(b)(c)(e)(h)(i) b Drywell Pressure - High 11, 12, 2
1,2,3 25 13(b)(c)(e)(h)(i) c.
Fuel Building Ventilation 13(e)(h) 1 28 Exhaust Radiation - High g
d.
Reactor Building 12(b)(e)(i) 1 1,2,3 29 A
Annulus Ventilation Exhaust l
Radiation - High l
4.
REACTOR WATER CLEANUP SYSTEM ISOLATION a.
A Flow - High 7, 15, 16 1
1,2,3 27 b.
A Flow Timer 7, 15, 16 1
1,2,3 27 c.
Equipment Area Temperature -
7, 15, 16 1
1,2,3 27 p
High d.
Equipment Area A Temperature -
I R
High 7, 15, 16 1
1,2,3 27 z
e.
Reactor Vessel Water l
Level - Low Low, Level 2 7, 15, 16 2
1,2,3 27
\\
f.
Main Steam Line Tunnel 7, 15, 16 1
1,2,3 27 l
Ambient Temperature - High
TABLE 3.3.2-1 (Continued) 2 99 ISOLATION ACTUATION INSTRUMENTATION VALVE GROUPS MINIMUM APPLICABLE c' '
OPERATED BY OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION SIGNAL *** PER TRIP SYSTEM (a)
CONDITION ACTION E::
4.
REACTOR WATER CLEANUP SYSTEM ISOLATION (continued) g g.
Main Steam Line Tunnel a Temperate e - High 7, 15, 16 1
1,2,3 27 h.
SLCS Initiation 7(f), 16 1(I) 1,-2, 3 27 5.
REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION a.
RCIC Steam Line Flow - High 2
1 1,2,3 27 u,
1 b.
RCIC Steam Line Flow - High 2
1 1,2,3 27 u,
J.
Timer
+
c.
RCIC Steam Supply Pressure - Low 2
1 1,2,3 27 d.
RCIC Turbine Exhaust Diaphragm Pressure - High 2
2 1,2,3 27 e.
RCIC Equipment Room Ambient Temperature - High 2
1 1,2,3 27 f.
RCIC Equipment Room A Temperature - High 2
1 1,2,3 27 g.
Main Steam Line Tunnel Ambient Temperature - High 2 1
1,2,3 27 h.
Main Steam Line Tunnel A Temperature - High 2
1 1,2,3 27
ATTACHMENT TO LICENSE AMEN 0 MENT NO. 56 FACILITY OPERATING LICFNSE NO. NPF-47 DOCKET NO. 50-458 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.
The revised pages are identified by Amendment number and contain a vertical line indicating the area of change.
The overleaf pages are provided ti maintain document completeness.
REMOVE INSERT 3/4 3-12 3/4 3-12 3/4 3-13 3/4 3-13
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