ML20126J719

Forwards NRC Questions & Positions Re FSAR Chapter 14. Incorporation of Resolutions of Items Into FSAR Will Reduce NRC Review Time & Ensure That Initial Test Program Design Review Will Not Impact Fuel Load or Startup Schedules

ML20126J719
Person / Time
Site:	Perry
Issue date:	03/24/1981
From:	Tedesco R Office of Nuclear Reactor Regulation
To:	Davidson D CLEVELAND ELECTRIC ILLUMINATING CO.
References
NUDOCS 8105050002
Download: ML20126J719 (24)

Text

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Distribution Docket Files NRC PDP DR bcc: NSIC MAR E 41981 $Cf p(3, itTERA^

0. Eisenhut S 06)

Docket Nos. 50-440 R. Purple and 50-441 R. Tedesco A. Schwencer D. Houston gil/ ,

M. Service O Mr. ' Dalwyn R. Davidson ELD y $

Vice President, Engineering IE (3) g '

The Cleveland Electric Illuminating Company B. Clayton, PTR8 1 P. O. Box 5000 S. Hanauer, DHFS '

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Cleveland, Ohio 441 01 9; [

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Dear Mr. Davidson:

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SUBJECT:

INITIAL PLANT TEST PROGRAM - PERK'I 1 & 2 IC We believe that it is important for us to complete our review and evaluation of FSAR Chapter 14 " Initial Test Program," as early as possible so that this review does not impact your licensing schedule. The changes to your test program or test procedures resulting from our review could impact your license ,

schedule if this action is required at a date near your proposed fuel load date, and if the changes require that you (1) increase your staffing for the

' initial test program, (2) modify and rerun preoperational test procedures which may have already been completed, or (3) modify startup test procedures. This is because increases in staffing normally require some lead time, preoperational test procedures must be written and performed prior to fuel loading, and startup test pret edures must be written and approved by the PORC 60 days prior to fuel loading.

Based upon recent reviews of the initial test programs for OL applications, we have concluded that our review time could be significantly reduced if the applicants would have addressed and accounted for staff information requests on previous applications. All of our recent reviews have contained questions and positions identical or similar to those which had to be resolved during previous reviews. If the applicants had resolved all of these staff questions

?: and positions in their original FSAR Chapter 14 submittal, we estimate the Chapter 14 review process could have been shortened by 2 - 6 months.

A copy of our questions and positions to Grand Gulf on Chapter 14 is enclosed for your consideration. Incorporating the resolution of these items into your FSAR will reduce NRC review time and should ensure that the review of your initial test program description will not impact your schedule for fuel load or startup.

THIS DOCUMENT CONTAINS  :

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810505000%- POOR QUAUTY PAGES ' <

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Mr. Dalwyn R. Davidson  !

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I The appropriate information should be submitted as an amendment to your Final i Safety Analysis Report. If you have any questions regarding these questions '

and positions, please contact Dean Houston, Project Manager (301) 492-8507. j Sincerely, t gngmal signed by

[

Robert L Todesco  !

Robert L. Tedesco I, Assistant Director for Licensing i Division of Licensing  ;

Enclosure:

As stated l;

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Mr. Dalwyn R. Davidson 4" 4 Vice Pr:sident, Engineering 198l lhe l'. Cle eland Electric Illuminating Company O. Box 5000 Cleveland, Ohio 44101 cc: Gerald Charnoff, Esq.

Shaw, Dittman, Potts & Trowbridge 1800 :'. Stree*., N. W.

Wa s ni ngton , D. C. 20036 Donald H. Hauser, Esq.

Cleveland Electric Illuminating Company P. O. Cox 5000 Clevelanc, Onio 44101 U. S. ?!uclear Regulatory Commission Resieent insoccror's Office Parmly at Center Road Perry, Ohio 44081 4

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i OUTSTANDING ITEMS i

_ GRAND GULF NUCLEAR STATION

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423.1 ll (14.2.2) Expand subsection 14.2.2 to. discuss how the plant engi-neering itial test staff program. will be utilized in conducting your in-  ;

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423.2  ;

- (14. 2. 3) Modify subsection 14.2.3 to describe the review and i and startup test procedures. approval process for preoperationa i be consistent with the responsibilities of personnelThis informa and thepreoperational for the Test Working Group stated in subsection 14.2.2 test phase.  ;

423.3 t Expand subsection 14.2.4 to describe your controls (14.2.4) provided for plant modifications and repairs. These l controls should assure that (1) required repairs or  ;

modifications will be made: (2) retesting is done,  :

as necessary, and (3) any proposed following the modifications or repairs; I facility modifications will be -

reviewed by the original. design organization or other designated design organizations. i i

423.4  !

(14.'2. 4 ) Provide a description of your method for changing test >

procedures when the scope or intent is changed. Note any differences in this method between preoperational  !

and startup tests. i 423.5 A (14.2.4) tion of the required preoperational testing" that isC!

required approval prior of testto results.

fuel loading includes review and  !

tional tests are intended to be conducted, or theirIf portions of any t results test; approved, after fuel loading: (1) list each  !

(2) state what portions of each test will be j delayed until after fuel loading; (3) provide technical i justification for delaying these portions; and (4) state  !

when each test will be completed (key to operating modes defined in your technical specifications, or to test  !

conditions defined in Chapter 14). Note that any test  !

that you do not intend to begin prior to fuel loading  ;

should be includedtest the preoperational in your startup test phase instead of phase. ,

i 7' 423.6 d t

'(14 . 2.11) .In or'er to facilitate the staff's review  !

of your individual test descriptions, provide  ;

an index of preoperational " test descriptions.

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423.0  !

OUALITY ASSURANCE 6 OPERATIONS 423.7  ;

Your response to item 423.4 is not acceptable. It appears (14.2) .that startup test procedure modifications can be made in a %s:

k manner not allowed by your technical specifications. Modify

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your method of making minor changes to approved startup test  !

procedures to make it consistent with your technical specifi--

cations requirements.  !

t 423.8 i (14.2.4) FSAR Section 14.2.4 states that the PSRC must review the test r {

results at each " power plateau" during the startup test phase WR t prior to authorizing further power testing. Define the power plateau. WU. !

423.9' e Modify FSAR Section 14.2.11 to state that copies of test (14.2.11) i procedures will be available for examination by NRC regional personnel approximately 60 days prior to the scheduled perform-  !

ance of preoperational tests, and not less than 60 days prior to scheduled fuel loading date for startup tests. (NRC l u__ j possession of the procedures should not impede the revision,  ; ,.

review, or refinement of the procedures.) i Ms; }

423.10 >

(14.2) We note your position relative to Regulatory Guide 1.80 '

' contained in Ap'pendix 3A of the FSAR and disagree with your .

position. This guide is applicable since the instrument i air system is used for a source of air for systems and l i

corponents that provide a safety function. Modify your I application to show that your test program will be consistent with the guide or show that you will conduct equivalent . . . .

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testing for the air system and supplied loads. A j 423.11 M:

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(14.2) Appendix 3A states that you comply with Regulatory Guide 1.108 b. l as discussed in FSAR Section 8., Expand the description (in Chapter 14) to show how you satisfy regulatory positions T fi  !

C.2.a and C.2.b of the guide, state explicitly any exceptions  %'

to these positions, and provide a description of equivalent testing for all exceptions. b- i '

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(14.2) , We 1.68, note that Appendix November 1973. 3A has comments addressing Regulatory Guide '

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This edition of the guide is not applicable  ;

to your facility and we have not reviewed these comments. The h' . . .  ;

applicable (January 1977). edition of the guide for your facility is Revision 1 additional industry and ACRS comments, prov than Revision 1. '^

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., Since we have compared your test program to the  !

later revision, we request that you address Revision 2 I i

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Our review of your test program description disclosed that the operability of several of the systems and components [

A listed in Regulatory Guide 1.68 (Revision 2), Appendix A

@k may not be demonstrated by your initial test program.

Expand your FSAR to. include appropriate test descriptions- ~t  !

(or modify existing descriptions) to address the following j items from Appendix A of the guide: '

1. creeperational testing i 1.a(2)(j) jet pumps. W [>

1.c demonstration of electrical independence and redundancy h(4 !

i of reactor protection systems and engineered safety i feature logic systems.

l.e(2) main steam system. '

l.e(5) steam extraction system. L-n  :

1.e(6) turbine stop, control, bypass, and intercept valves.

f l.e(7) main condenser hotwell level control system. ,

i.e(8) condensate system. ,

i 1.e(10) feedwater heater and drain systems. )

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1.e(ll) makeup water and chemical treatment systems.

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,T i 1.e(12) main condenser auxiliaries used for maintaining -

vacuum. "'

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1.f(l) circulating water sysNm.

1.f(2) cooling towers and associated auxiliaries. W m.,

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. 6 l.g(l) normal a.c. power distribution system (including t,.[ }

automatic transfers between sources). l

.3 1.g(2) emergency a.c. power distribution system.

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tests of structures and equipment (e.g., watertight hatches, walls) that protect engineered safety features ['(

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f from flooding (internal or external).

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1.h(1)(d) demonstration of operability of interlocks and l isolation valves provided for overpressure protection of low pressure core spray and low pressure core injection (RHR) systems,  !

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1.h(3) tanks and other sources of water used for ECCS (e.g., condensate storage tank, suppression pool.,

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and suppression pool makeup pool). '

1.i(1) containment design overpressure structural tests >

(drywell and containment).  ;

t 1.i(2) containment isolation valve closure timing tests U (should verify that closure times, including i' i time delays of initiating logic are in accordance t."

with design assumptions).

1.i(3) containment isolation valve leakage rate tests. -

1.i(4) containment penetration leakage tests.

1.i(5) containment airlock leak rate tests. u l l.i(6) integrated containment leakage tests. Q 1.i(10) containment vacuum-breaker tests (drywell/ containment).

1.i(15) containment penetration pressurization system tests, l.i(19) bypass leakage tests on pressure suppression containments. . _.

i . i(21 ) containment penetration cooling system tests. hI

, 1.j(3) m. \

1 steam pressure contro{ system. g iP"  !

1.j(6) loose parts monitoring system.

1.j(9) pressure control systems designed to prevent leakage b r

across boundaries (feedwater leakage' control system). f,1 1.j(10) seismic instrumentation. Vbj

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1 1.j(16) hotwell level control system. ,% 1 tt ph  !

1.j(17) feedwater systems.

heater temperature, level and bypass control i r,;,...

1.j(18) auxiliary startup instruments (neutron response checks).

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1.j(20) instrumentation used to detect internal flooding

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t as fluid system piping failures. #;r 1.j(23) post-accident hydroom monitors and analyzers used "  !

in the combustible L.o control system. .

i 1.l(8) plant sampling systems.

1.m(3) operability and leak tests of sectionalizing devices ,

and drains in the refueling canal or fuel storage pool. .

, )7g l.m(4) dynamic and load testing of cranes, hoists, and 'IAL-Pt '

associated lifting and rigging equipment, including j the fuel cask handling crane. i L

1.m(6) irradiated fuel pool or building ventilation system i' tests (fuel handling area ventilation). ,

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, 1.n(l) plant service water system. ,

1.n(2) ik?N I turbine building cooling water system. i  ;

1.n(5) sampling systems.

1.n(6) chemistry control (condensate demineralizers and  !

fil te rs) . $

1.n(7) fire protection systems. _..

1.n(11) compressed gas systems.

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1.n(13) communication systems.

7e i 1.n(14)(e) auxiliary buildingf turbine building, radioactive ,'$

waste handling building heatins , cooling, and '  !

ventilation systems. -, t

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L@;: 1 1.n(14)(f) control room habitability systems: leaktightness of ducts and flow rates in systems, and proper 59 j control of space temperatures. rd. ,

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1.n(18) heat tracing and freeze protection systems. '4; '

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1.o reactor components handling systems (e.g., polar' crane),

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~4. low oower testing.

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4.e. flux distribution measurements. i 4.1. demonstration of operability of rod inhibit or block '

functions. '

'T i 4.m. demonstration of operability of the t'SlV leakage control $l system at hot standby conditions. '

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5. power ascension tests.  :

t 5,. a . demonstration that power vs. flow characteristics are  !

in accordance with design values, u_

5.c. control rod pattern exchange demonstration.

5.g. demonstration that all control rod sequencers, rod -

worth minimizers, and rod withdrawal block functions operate in accordance with design.

5.m. demonstration that reactor coolant system flows, pressure  !

drops, and vibrations are in accordance with design for j various operating modes. --i nw <

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5.o.

baseline data for loose parts monitoring system.

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calibration of instrumentation and demonstration of proper *** 1 response of reactor cooQnt leak detection systems. t 5.s. verification of performance of plant control systems (e.g., hotwell level control, steam pressure control),

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5.t. verification of response times of main steam relief valves, turbine bypass valves and turbine stop, control, and @ l g?.'

intercept valves; also, verification of the capacity rM ' '

of the turbine bypass valves. W

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5 w. demonstration of adequate performance margins of shielding and penetration cooling systems. -

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5.x, demonstration of adequate performance margins for auxiliary systems required to support the operation, i[b.. I of engineered safety features or to maintain the '

1 environment in spaces that house engineered safety fea tures . [

5.c.c. demonstration that gaseous and liquid radweste processing, storage, and release systems operate -

in accordance with design. ,

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5.f.f. demonstration that the ventilation system that serves ~!

the main steam line tunnel and the containment cooling  !

system maintain temperatures within design limits.

423.13 t We could not conclude from our review of your preoperational (14.2.12) test descriptions provided in Section 14.2.12 that comprehensive i

f"; l testing is planned for several of the listed systems, structures, and components. g.C. j Therefore, clarify or expand the description '

t of the preoperational tests to address the following. i 1.

Low Pressure Core Spray System Preoperational Test (7) -

Verify that the response time of the LPCS system is in l accordance with accident analysis assumptions.

2.

Reactor Recirculation and Flow Control System Preoperational ~'

Test (10) - Expand the description to demonstrate that flow control valve analysis stroke times are in accordance with accident assumptions.

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Control Rod Drive Hydraulje System Preoperational Test (11) -

Clarify prerequisite numbel 7.  :. -

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4. .I Reactor Protection System Preoperational Test (16) - d i

Expand the description to state how you will accomplish ' i Test Procedure Item 5, and to include the remaining A. . r.i !

pressure, level, or AP sensors in this item.

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Trave.stng Incore Probe. (TIP) System Preoperational Test (18) - hk Test Procedure Item 2 does not appear to be consistent with r

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your design. Chapter 7 states that there are no ball or

- shear valves in the TIP system. Clarify this inconsistency. .[ t

6. 6 i Process Radiation tionitoring System Preoperational Test (19) - -

Expand Test Objective Item 2(e) to state which systems are monitored by the liquid process radiation monitoring sub-j  ;

sys tem. -  !

Item 6 of Prerequisites states that check sources  !

are in place "where required." If any radiation monitors {

are to be tested without a check source, describe how they j will be tested.  ;

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7. Combustible Gas Control System Preoperational Test (24) -

Clarify the test description to show that the post-LOCA i hydrogen monitors will be tested.  %, ,

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Standby Service Water (SSW) System Preoperational Test, (26) - -

Expand this description or reference another test description  ;

that will demonstrate the following: a) correct flows to '

supplied components; b) leaktightness of valves which isolate SSW from other systems (e.g., plant service water, i component cooling water); c) water loss from cooling towers  !

and basins (e.g., spray drif t, evaporation) is in accordance . F---

with design assumptions; d) heat removal capability of N*..,

towers; e) absence of vortexing and adequate NPSH for the f!l SSW pumps and HPCS service water pump over the range of l basin level from maximum to the minimum calculated 30 ,

days following LOCA; and f) that the overflow lines are l unobstructed.

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9. Control Room HVAC System Preoperational Test (29) - Expand  :

the test description to include a demonstration that y i inleakage to the control room is in accordance with design gy  :

assumptions when the system is in the recirc mode (when ,*  ;

control room is not pressurized). -

10 Class lE 125 Volt DC System Preoperational Test (39) -

State your plans for demonstrating the following: i a) that emergency loads are in accordance with battery sizing assumptions; and b) that each emergency load can operate at the minimum voltage level at which it F.;~; L can be postulated to operate. I 11 . Solid Radwaste System Preoperational Test (41) - Expand ~l the test description to demonstrate that the solidification M.

equipment and process operate properly by processing a sample batch. The acceptance criteria should require that p .

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there be no free liquid in the solidified sample.

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12 . Primary and Secondary Containment Isolation System M I

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Preoperational Test (43) - Clarify the acceptance criteria M i to show that the valves, dampers, and systems respond in ,. , l the required times.  ;

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13 ECCS Integrated Initiation During Loss of Offsite Power Preoperational Test (44) - Clarify the test description h

to'show that you will demonstrate independence of emergency W".

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i buses and correct assignment of loads by performing the  %

loss of offsite power coincident with LOCA signal three "

,. l times, each time allowina only one diesel generator to 54.'  !

start, and having only its associated DC system energized. '

If this test is to be used to demonstrate independence t'  :

and proper load group assignments in accordance with Regulatory Guide 1.41, expand the description to state - )

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how this will be accomplished. If this test will not  !

accomplish this demonstration, provide (or reference) descriptions of tests which will. s..

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(M.2.12). Our review disclosed that several of your startup' test "

descriptions are not acceptable. Modify the following. .

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. Control Rod Drive System (5) - The acceptance criteria  :

-for rod insertion times (scram times) appear to be _ .-

nonconservative with respect to your accident analysis. -

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a) Modify the criteria to be consistent with Figure '

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15.0-3 or justify exceptions; b) provide the bases for the less conservative acceptance criteria for @['7 i'

j scram times with vessel dome pressure less than 950 psig; c) verify correct buffer action for each rod; and d) correct the numbers in the table in

Figure 15.0-3. '

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2.  ;

RCIC System (14) - Clarify the test description on Figure 14.2-4 to show the power level restriction if any Level 1 L

i acceptance criteria are not met. ' Also, note that the h;' , ! '

time that you are allowed to operate with RCIC inoperable -

will be limited by your technical specifications. I

3. t Initial Pressure Centroller (IPC) (22) - Provide assurance that the Level 2 acceptance criteria for system response to ,

setpoint changes are based on the expected performance for  !

the existing test conditions (e.g., beginning-of-life r-reactivity feedbacks).  !

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4. Feedwater System (23) - Expand the description to describe how the heater loss will be effected. Verify that this is **>-

the limiting heater loss that could result from a single equipment failure or single operator error. Verify that the Level 2 acceptance criteria for system response to

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setpoint changes are based on expected performance for p;,, j gg' .

the actual test conditions.

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5. Main Steam Line Isolation VaIves (25) - a) modify the test f,.  !

description to measure full closure time of the MSIV's or provide technical justification for extrapolating the full N p,

closure times from 90f, closure data; b) mcdify the Level 1 l p,.; . i acceptance criterion to be consistent with your technical 4

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specifications (they will not allow averacino closure times);

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c) provide assurance that the acceptance criteria for pressure and flux are based on realistic predictions for h i l

the test conditions, not on accident analysis (worst case) assumptions; and d) provide acceptance criteria for relief valve reset pressures.

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Relief Valves (26) - Pro' vide the bases for the capacity {

acceptance criteria. Verify,that the demonstrated relief ._

capacities are in accordance wi'th your overpressure  ;

protection design bases and the accident analysis 4" . -

assumptions for minimum and maximum capacities. Also, @

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i provide a description of test methods and acceptance ' j criteria-(and their bases) which will verify capacity 1 of the ADS valves. '

7. Turbine Trip and Generator Load Rejection (27) - Provide separate test descriptions for the two tests. For each ,

test, verify that the Level 2 acceptance criteria for , p -

reactor pressure and heat flux are based on realistic predictions for the test conditions, and not on accident 4;. ,

analysis (worstcase)' assumptions. Provide acceptance

' criteria, and their bases, for capacity of bypass valves; ,

response times of turbine stop, intercept and control valves, and recirculation pump trip; and relief valve reset pressures. For the generator load rejection, ,

state how the generator will be tripped and verify ,

that this method causes maximum turbine overspeed.

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Shutdown From Outside the Main Control Room (28) - Modify the test description to show that the test will be conducted in accordance with Regulatory Guide 1.68.2, Revision 1. or provide technical justification for any exceptions. State what method of cooldown will be initiated. If this test does not demonstrate initiation of cooldown using the shutdown cooling mode of RHR, reference the test in which --

this will be demonstrated. Provide acceptance criteria for the performance of plant equipment and the variables @ ".

to be monitored during the test. @

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9. Recirculation System (30).- Modify the acceptance criteria to a) verify that flow coastdown rates are in accordance with maximum and minimum accident analysis assumptions; h%'

b) verify that pu*p startup rates are within the bounds

'g4 of the accident analysis assumption; and c) verify that ;f

' no cavitation occurs in the system in all allowable modes *"

of operation.

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Loss of Turbine-Generator and Offsite Power (31) - Modify p,nb the test description to a) provide assurance that the loss 9.,

of offsite power condition will be maintained for at least I *** ,

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30 minutes to demonstrate that necessary equipment, controls, and indication are available following station blackout to l remove decay heat from the core using only emergency power supplies and distribution systems; b) show that the test will be initiated with electrical systems aligned for h,

normal full-power operation; c) verify that the acceptance -:

l criteria include proper start and load times for the l s'

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I diesel generators; .and d) verify that the reactor pressure acceptance criterion is based on a realistic prediction ,

for the test condition. ,

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Isolated Reactor Stability (36) - This is a new test that kp  :

has not previously been reviewed by the staff. Provide d_.  :

background information which shows this testing to be' i desirable or necessary. Modify the test description to  !

better describe the test conditions (e.g., rod positions,  ;

core flow). Provide the bases for the acceptance criteria.  :'

Note that current standard technical specifications do not

• allow " black" startup (e.g. startup without offsite power). ,

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C3.15 Table 14.2-3 indicates that the !4SIV full isolation may be Y.i,i c  !

(14.2) demonstrated at any power level greater than 75%. Table 14.2-1 indicates that it may not be demonstrated at all. {

Either modify the tables to show that the test will be conducted at 100% power as stated in Regulatory Guide l 1.68 or provide technical justification for the exception. .

423.16 Table 14.2-1 indicates that a turbine trip may be conducted L. {

(14.2) at 100% power and that a generator trip will not be conducted c-k.

a t 100% power. Table 14.2-3 indicates that either of the i two will be conducted at test condition 6 (which is  :

approximately100% power). Modify the tables to show that  :

you intend to conduct both a turbine trip and a generator l trip at 100% power as stated in Regulatory Guide 1.68 or provide technical justification for any exceptions.

423.17 The test conditions shown in Table 14.2-2 for some.of the stability *-' '

(14.2) tests.do not agree with the ones in Table 14.2-3 this inconsistency.. ~

Eliminate 5-

,l 423.18 (14.i)

Provide a description of the electrical lineup for Unit No. 2 during preoperational t6sts that will be conducted - M

,r y to satisfy regulatory positions in Regulatory Guide 1.41 ,gg for Unit No. 1. Provide a description of the lineup f'6'r" both plants during similar preoperational testing on w" '-

y, 7h- ..

Unit No. 2 subsequent to initial criticality of Unit No.  : $

l. The descriptions should address both normal and -

emergency AC and DC power distribution systems. Provide assu,ran,c,e i'V that crossties will not exist which could cause loss of ' *(  ;

emergency bus power to one unit due to testing of the other unit. k. -

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. 423.19 Provide a commitment to include in your test program any design 8 (14.2) features to prevent or mitigate anticipated transients without '

scram (ATWS) that may be incorporated in your plant design. "

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423 ?0 provide preoperational test descriptions (or modify existing  !

(14.2) i descriptions) to assure that each engineered safety feature -

pump operates .in accordance with 'the manufacturer's head-flow  :

curve. Include in the' description the bases for the acceptance c ri te ria . (The bases provided should consider both flow require- Qt ments for ESF functions _ and pump NpSH requirements.) [@ ;;

i 423.21 Because of the high frequency of reported occurrences of snubber (14.2)  !

fcilures in operating plants that are being disclosed by routine  ;

~ surveillance tests, it appears that preservice inspections and/or t functional tests of snubbers are warranted during either the ,

l preoperational test phase or startup test phase.- Describe your -

plans for conduct of tests and/or inspection of snubbers installed' W9 ;;

on ASME Section 111, Classes 1, 2, & 3 systems and high energy $ -l systems at your facility. '$ s t?3.?2 Identify any of the post-fuel loading tests described in Section t

i14.2) 14.2 which are not essential towards the demonstration of conform-1 L

ance with desian reouirements for structures, systems, comoonents, l and design features that meet any of the following criteria: '

'" k (1 ) Will be relied upon for safe shutdown and coo'ldown of I" 'I i the reactor under normal plant conditions and for main- -- I taining the reactor in a safe condition for an extended  !

shutdown period. r t

(2) Will be relied upon for safe shutdown and cooldown of the reactor under transient (infrequent or modera tely . h '

[

frequent events) conditions and postulated accident  ;  !

P.  !

conditions, and for maintaining the reactor in a safe condition for an extended shutdown period following {.;': i i such conditions. .

. F na.

(3) .

Will be relied upon for eltablishing conformance with .  %  :

safety limits or limiting conditions for operation that will be included in the facility technical specifications.

(b., 0.,a  !

j I% s.

(4) Are clas'sified as engineered safety feature's or will' be >

relied upon to support or assure the' operations of

'5

• engineered safety features within design limits.

!(()

' l (5) '.

}

Are assumed to functicn or. for which credit is taken,q . W  :

in the accident analysis for the faci.lity (as described,,, yI.*

in the Final Safety Analysis Report).' ' l' u

{

(6) Will be utilized to process, store, control, or limit the release of radioactive materials. l

'N h;.. '

)

p . .

• e e

- ,l 0 3.23 Provide or modify test descriptions 1) that will verify that i the plant's ventilation systems are adequate to maintain all ESF equipment within its design temperature range during normal operations; and 2) that will' verify that the emergency ventilation systems are capable of maintaining all ESF g.

e**

• equipment within their design temperature range with the P'  ;

equipment operating in a manner that will produce the ' '

maimum heat load in the compartment. If it is not i practical to produce maximum heat loads in a compartment, 3 describe the methods that will be used to verify design i heat removal capability of the emergency ventilation systems. l 423.24 Our review of licensee event reports has disclosed several (a h (14.2) instances of RCIC pump failure to start on demand. It appears

.j ,

that many of these failures could have been avoided if more thorough testing had been conducted during the plant's initial test programs, in order to discover any problems affecting pump startup and to demonstrate the reliability of your  :

RCIC system, state your plans to demonstrate at least five consecutive, successful, cold, quick RCIC pump starts during L_

I your initial test program.  ;

• N.

423.25 Our review of licensee event reports has also disclosed that k. '

(14.2) many events have occurred because of dirt, condensed moisture,  ;

or other foreign objects inside instruments and electrical cm.;,onents (e.g. , relays, switches, breakers). Describe any ,

tests or inspections that will be performed or any administrative controls that will be implemented during your initial test y i

program to prevent component failures such as these at your facility. --

[

4 23.25 Modify Figure 14.2-4 to locate points A through F on the map.

N

" i (14.2) Your operating license will not allow operation at a power 'N l

level greater than rated thermal power. Therefore, modify i i

your definition of Test Condition 7 :r.d locate this test .

condition on the map. p'[?

1 423.27 L.A' {

We have reviewed your minimum qualifications for personnel .;.

(14.2.16) participating in the initial test program and find them i

• Y  !

acceptable. However, Section 14.2.16 references Regulatory l Guide 1.58. This Regulatory Guide is not applicable for {^E

, ;' .' r the testing which is described in Chapter 14. There fore, n/M remove this reference from Chaoter 14.  ; 4' ',  !

.e l

m  ;

n i

• a

4

- /

- s .

':n 114- -

n

'LA 'TY 055URANCE AMD 5? ERAT 10NS i:.~3 The response' to item 423.7 is not acceptable.

Your olant's :sc fi:a1 a s

ia.-i speci{::aticns wi.ll recuire that minor :ehp:rary changes to :r;;g:gre's covertrg test activities of safety-rela:ed 'ecui:~en: rus- e a: r:.e:  !

by r..o ~e-tcrs of the :lant managemen; s:af', a- less; : e ce ..; , ech

noics a Senior Reactor Operator's License en

ne af'ec e: unit.

a nte nost, it not all, startup tests affect safety-rela.ec sys s s,

[

nis re,uirement a: plies to startuo test pr:cet.ees. (;; :ess ne:  !

3: ply :0 creo:eraticral tests concutted befcre #;ei Icacin:.)

iN.e /er, ,

S'R Section 10.2.4 indicates that min:r changes .c 2 s tar up '

}es; 15 procecures may be made "on-the-srot" by :ne tes t su:erviser (who '

ngt required to hold a Senior Reactor Opara:ce's License). :cdify seg W1  ;.1cn ia.2.4 to show that minor changes to s:2-tu: test :r:cedures i i te mace in accorcance with your echnical 1:ecification requirements.  ;

413.29

('i.El The resp:nse to item 423.10 indicates that :eiting will be c:nducted I in acc rdance with Regulatory Guide 1.30, June '974, with the exception listed ;n F5p; page 2A/1.30-1. Provide a css:r'o:icn of :nts pre-operational test. .

4'.3.20 f (ic.2, The responses to some sub-parts of item 423.12, regarding Re;ulatory Guice 1.68, are nct acceptable. provide the in'ccmation rec.es ted j balow.  ;

l .e(2), i.e(5), l .e(6), i .e(7), l .e(8), i .e(10), l .e(ll), i .e(12), .

l .f(l ), l .f(2),1.g (i) ,1.j (3), l .J(16), l .J(17 ),1.l(8),1.n(l ),

_ ._ 1.n(2),1.n(5),1.n(6),1.n(ll),1.n(14)(e) t Your response states that these systems and ccm:onents will be tested

' as a part of your Acceptance Testing Program. It is our position that  :

these be systems and comoonents are important to safety and should, therefore, incluced in your preoperational test program. Provice a summary description of each of these tests and either (1) include these tests in your preoperational test program, or (2) provide a description of the administrative controls for the-Acceptance Testing Program. If you decide on option (2), provide a sufficiently detailed description to enable us to determine that the review of acceptance tes: procedures, conduct of the tests , and review of the test results are comnensurate  :

witn those of your preoperational test program, t s

1(h)

Your resconse to this iten does not address the leaktightness  !

cf structures ethich protect ergir. sered sa'ety 'et .res ' rom '

'::ci.;. I- is our ::si:i:n r : ,:. ;er#: - a:D ::

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P

.r l.h(3) Your response to this item regarding tanks ,.nich sut:tr:

is garbled. Clarify your response'. 'CCS

' . i(1) , l. i(3), l . i(4), l..i(5), l. i(6), l . i(15 ) ,1.i (19)  !

' erify that the structural integrity test as descri:ed ir

~!*? Cection 3.3 and the centairment teit: :escribed in lection or: gram.

6.2.6 are done as part of ycur ; rec:erational test >

i 1.'s21) Your response to this item states that no ctntainment r penetration coolers are used in your clint cesign. 3rovide ,

a description of a startue test that .till demonstrate tnit concrete tenceratures surrouncing not :inetestions (e.g.,

cain steam lines) do not exceed cesign imits,  !

i 1.j (9 }

The retconce states that testing of tne feecaater leikage )

c0ntr:l system will.be addec to pre:peratioral test cescription NO.

5, " Residual Heat Removal System Prec;erational Test."  ;

Mocify the test description to describe this testing. i l.7(3) Your response states that fuel pool leak detection and  !

sectionalizing devices test will be con: acted as part cf the I fuel cool cooling and cleanuo system ;reocerational test (22).

Modify tne test description to cescribe tnis testing. l 4.m I Your response states that a leakace cemonstration of the MSIV f leakage control system is conducted as precoerational test No.

9.

Accarently this test will be conductec at cold conditions, 1 j

Provice PS:V-L:

a description of a test which cemonstrates tnat tne

"' t 5 c:moonente operate proce*1y when handling steam anc

'nat the system can handle the amount of leakage tha- is present when the main steam system is at operating temperature, i, 5.c.c. 4  :

Your response does not addrers the liquid raowaste systen, 1 Provide a description of your startup test wnich demons

• rates i the processing of liquid racicactive wastes.  !

223. 0  :

(14.2) Your response to item 423.13 is not totally acceptable. i infdrmation requested below. Provide the {

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Traversing Incore Probe ('TIP) System Preo:eca:itnal Tes: O 3; -

"cdify the test description to delete the accep:2r.:e cri;2ri:3 accressing the squib charge.  !

3. t StanCby Service Water (SSW) System ?reo: era:icnii Tes (26; - i Yc;r response indicates that the leaktigh*.ntss :f va'.ves f a: '

is;lete the SSW system from other systems is ad:ressed in t*,e  !

revi;sd test description. '

the tas; description. However, this is not adcressed by i Expand the descriction :: state h:'.. y:u will demonstrate ne leaktightness of tne valves which isolate ,

the SS'.i system from the plant service water systen and the c:n:cnent cooling water system, ,

4. Class lE 125 Vol OC System Preoperational Test (29) - Ycur resconse states that preoperational tes: No. 04 demonstrates ,

that the DC system, loads are in accordance witn bat:ery si:in; as s umo tien s .

stratec.

However, it is not clear now :*.is will be ten:n-  !

icad will be measured and verified to be consistentModify with ba ttery the sizing assumptions. Your respense also refers : D the response  ;

to item 0:1.31(f) for the minimum voltage demonstration. The esponse to item 041.31(f) does not state Inat the DC loads will >

i te demonstrated to be operable at minicum voltage levels; it only states that the DC equipment is "specified" for operation over ,

a range of 105 to 140 volts. It is our position that you demen-strate the ocerability of the DC system loacs at the minimum  ;

voltage level at which they can be postulated to ccerate. Moci fy the test description to describe this cemonstration.  !

423.32 ',

- (14.2} Your response to item 423.14 addressing several startup test descriptions is not totally acceptable. Provice the information requested belcw. ,

1.

. Main Steam Line Isolation Valved (25) - Your resp nse stated that acceptance criteria for relief valve reset pressures are .

, contained, "if applicable," in STI-25. It is cur position that  ;

you demonstrate during your initial test program that the relief i i

valves reclose at the correct pressures. Modify tnis tes

' that the test procedure will contain acceptance criterial relief valve reset pressure. j

2.  ! ? ; * ' '. M ~rCm Cu;;ide t".e Main C;nt^ci 0  !

I'..~' 0 :~i (~3 V :'. *

~:?

. . . . ~ ; * : ." * " ; . Oid:  :;'#'i' ** ~';

c. 'ia 3 ' ith pegula:Ory Guide 1.53.2.

= .

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s *. 3. .t .E- ,s. .t ' ". .a t *. P e '* : ! * *. 0

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c:'*.r 1 rocm).  !

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t e .. . , t a: . - 10. 0 F

'3.

Recirculation System (30) - Modify th'e test description to (a) pr0 vide acceptance criteria for speed of the recirculation pumo [

(folicwing trip of the normal supply breaker) ,.nen the LFMG set breater closes, (b) provide acceptance criteria for pumo startup {

ra:es t.nat are consistent witn. assumotlers ci.er...s :ection 15..,.,,

. i an: ,

al'.; acie power an:1c; cescr1:e new you w1,41 ver1:y that no cavi*ation occurs t,a ficw conditions.

4 Lc:s of Turbine-Generator and Offsite power (31) - The loss of e t

turbine. genera:Or and offsite power shoul.' be maintained for a j

peri d of time sufficient to demonstrate .nat :ne P.ecessary ecui: ment, controls, and instrumentatior statien tiackcu :: are availa:le fcilowing i emergency acwer su plies. rem;ve decay heat f.:m the :cre using enly the 1:55 It is our position that ycu maintain l i

of #fsite ;cwer for at least 30 minutes in order to der nstra e inis. Y cify the tes descrioti:n c state that I cffsite power will nc: ce res*cred for at leas: 30 minutes.  ;

I 023.23 t n Your respense s,*4sto.*a. item 423.15 indicates that FSAR Table 14.2-3 has been

- L L . ; . c ,- . ,. 4 . 1 c-

.m .., v..e.u.

.. .~. . i e.ui, -

4..e.. a.< . . n. .iei a se ,..ne ses: C ndi:10n o. . . .... .. . a ,. .,

ms :sm:ns: ation may un tne be ;erf:rmecc:ntrary,ainis taole s:1;,i indicates that .. .

s

ne *a;,le 0

. . . "anywnere) 75% power." Modify

.. To. s e. incica*e that tne Tull 1 solation wi,ll be cemenstrated e.n. e.s 4. *. a. C n G. .

t 423.3; .

,14.2) fcur res::: ente *o i em 423.22 addresses prec;eratier.ai *ests. The item refers a23.22 : ; cst.#uel leading tests. ::evise ycur res;cnse to item accress only :ost-fuel icacine star:co tests (precritical  !

l tests, initial criticality, icw-power tests, and p wer ascensi0n tests)

_. _,23...;:

Ycur res: nse to 1:en ,2;..:: .. does n0:

(la.2) ments regar:in; tne nea: complete,y satisfy our require- .

t serving areas that h0use engineered-safety features. It is notremoval capabilit EEF equiprenap;arent that post-accident design heat leads will be ; reduced in recms during the pcwer ascension test phase; tnerefore,  !

s '~-* '.. v .S .e .

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rem . a. . - s *. .* .= *. e *. .*.

o c a.a.s i a, n .r. a. a *. i to ,cval ca; ability Of these systems. Mocify your test descriptions  ;

19. .awinclude

e. 2 . e. * " .a measurement Of air and cooling water te :eratures an: i

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- r OUTSTANDING ITEMS '

GRAND GULF NUCLEAR STATION '

423.39 The response to some sub-parts of' item 423.12, regarding Regulatory Guide 1.68, are not totally acceptable. Provide the information requested below. .

r 1.i(2) Your response states that the logic time delays will be included in acceptance criteria which are currently '

being' developed. At present no reference to the acceptance criteria is contained in Subsection

• 14.2.12.1.4.d. Include the requirement to meet the  ;

acceptance criteria for logic time delays in the test I abstract. i 423.40 .

ine response to item 423.13 is not totally acceptable. Item 423.13, 8f, states that the test description should be expanded i

to demonstrate that the overflow lines are uncbstructed. The I

. response refers to revised Subsection 14.2.12.1.26, which includes '

no mention of the overflow lines. EhandtheStandbyServiceWater I

(SSW)SystemPreoperationalTest(26)orreferenceanothertest '

abstract that will demonstrate that the overflow lines are s '

unobstructed. ,

i c23.cl The response to item 423.29 is not totally acceptable. Include test descriptions that fulfill requirements C.9 and C.10 of '

Regulatory Guide 1.80 as part of the Instrument Air System preoperational Test (14.2.12.1.59), or provide technical 1

justification for this exception to the guide. i i

t

c i

p 423.42 The response to item 423.30 is not totally acceptable. Modify  ;

the response to 1.i(21) to provide a startup test abstract that will evaluate the adequacy of cooling for those selected penetrations where fin: were analytically determined to be necessary.

For all '

other high-temperature penetration lines where analysis indicated no cooling wcs required, either include those penetrations in the  !

aftrerectioned startup test abstract or provide .

- Evaluation Fe:hodology 1

- Maximum Concrete Temperature Criteria

- Assumed Heat Transfer Coefficients '

- :escri: tion of Similar Applications Wnere, for Comparable i Temperatures and Materials, No Cooling Was Used Successfully 423.03 The resp;nse to item 023.33 is n0 totally acceptable. Modify Subsecticn 10.2.12.3.22.2.c to indicate that the simultaneous full r

iosure of all MS:Vs will be performed at TC-6 (100 percent test plateau). .

Mocify the response to item 423.33 to show that the test r

will be performed at TC-6 for GGNS Unit 2 as stated in Regulatory  ?

Guide 1.53, or pr: vide technical justification for this exception to the guide.

22.~2 The res::rse :: item 023.37 re:uires -i r : ari#ica:icn. The res:Onse states that a :uroine trip will te : nducted during 70-3 at 50-50 per:en: poner, and a:

greater than Or ecua; to 95 percent core flow.

Ta:1e '.4.2-3 has been :. van;ed to in: uce the :grtine tri:

.r'r; 7:-3; n:aever, 70-3 is :sf'*ed as 5: E per:en: c:aer anc '

>i: per:en: :: e f :..

rt'..;e an additi:nal f:::n::e to Table 14 .- 23 stating :ne ::aer 3..: fi c o-

r.:'.:icrs :nat wil. e establisned for

.e... . , ..

G.

r 423.45 Certain data that was to be provided in 'the FSAR Chapter 14 has not been made available. Supply the missing information described below: i

a. p. 14.2-133 Section 14.2.12.3.14.d  !

Tolerance data that was to be provided in June 1980 is not given.

b. p. 14.2-169, 170 Section 14.2.12.3.29.d '

riteria and allowable novement information that was to be  !

provided in June 1980 is not given.

4 423.46 Figure 14.2-4 (Startup Test Condition Power Flow Map) is of major im:ortance to the understanding of FSAR Chapter 14. I There are a number of significant problems with the figure <

j t ht : neec to be resolved. Revise Figure 14.2-4 to address the felicsing deficiencies:

i

a. TC-2 .

The 50 cercent control rod line is not labeled on the figure.

The remaining nemenclature (analytical lower limit of master flow control mode, bypass valve capacity) do not match 4

the nomenclature used for points A, B, C, D, E, F labels or terms used on tre figure. Clarify the area to be represented i

by TC-2.

b. TC-3 The arrow points to the Region IV area, which is si;nifi:an:1y lar;er : nan :ne 7 -3 ori :en cefinitien

. (a::ve 20 ;e :ent : ore fi:w, etc.).

423.46 c. TC-4 (CONTO)

The- + 5 percent criteria does not specify whether it

~

applies to rated thermal power, or rated core flow rate, or both. ,

"LINE WITH" should be "LINE WITHIN".

d. TC-6 .

The hatched area indicated for this test condition does not match the written definition, e.

Wnat does the hatched region IV represent? Also what does the r

hatched region to the immediate left of region IV represent?  !

Define regions I and II.

423.47 Subsections 14.2.10.2.e and 14.2.12.3.2.c state that at each major power level (25%, 60%, and 100%), the Local Power Range '

Monitors (LPRMs) are calibrated and radiation measurements will ce made at significant locations throughout the plant. Modify the pcwer levels for LPRM calibration to match the test power levels,' or provide technical justification for using power levels other than the power plateaus cited in Subsection 14.2.4.5 (10%,

25%, 50%, 75%, and 100%).

-t 4 23. t. 8 Modify test abstract 14.2.12.1.5 to verify that paths for air-flow t

test of containment spray no:zles overlaps the water-flow test paths of the pumps to demonstrate that there is no blockage in the flow i path.

3 423.49 List any tests, or portions of tests, described in Section 14.2.12 1

}

which you do not intend to perform on each unit and provide technical i Justification for deletion of each.  !

I i

423.50 Provide assurance that all sources of power supply to vital buses are capable of carrying full accident loads. If some portions of the power supplies cannot be full-load tested, provide justification. .

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