ML20247R377
| ML20247R377 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 07/31/1989 |
| From: | Beck J TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TXX-89471, NUDOCS 8908070440 | |
| Download: ML20247R377 (23) | |
Text
- _ _ _ _ _ _ - - - _ _
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Log # TXX-89471
._ C File # 914.2 r = 10010 Ref. # 10CFR50.34(b) 7t/ ELECTRIC wmiam J. cahm. Jr.
Executive We President U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 l
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (Cf5ES) l DOCKET NOS. 50-445 AND 50-446 RESPONSE TO NRC CONCERNS REGARDING AC ESSENTIAL LIGHTING, CONTROL R0D WORTHS AND LOSS OF 0FFSITE POWER Gentlemen:
In accordance with the discussions between TV Electric and the NRC, during the preoperational and startup testing audit conducted by June 20-22, 1989, by Messrs. J. Zwolinski, R. Ramirez, F. Ashe, R. Gruel, and M. Field, l TU Electric hereby responds to the issues related to AC Essential Light acceptance testing, Control Rod Reactivity Worths and Loss of Offsite Power.
Per the subject audit TV Electric hereby notifies the NRC that CPSES will perform acceptance testing for Class IE AC Essential Lighting.
Attachment 1 provides an advance FSAR change submittal for Q&R 423.16 and 423.31 for further clarification of Control Rod Reactivity Worths.
Attachment 2 provides an advance FSAR change submittal for correction of Table 14.2-3, Sheet 18, Loss of Offsite Power Test Summary, and O&R 423.16.
Sincerely, I
William J. Cahill, Jr.
By: M, A-l J@n'W. Beck l Vice President.
Nuclear Engineering RSB/vid Attachments DOb c - Mr. R. D. Martin, Region IV
' \
Resident Inspectors, CPSES (3)
Mr. J. H. Wilson, OSP-NRC i 4 FDC
Attachment I to TXX-89471 July 31, 1989 iPage 1 of.11 ADVANCE FSAR CHANGE-t In the preoperational.and startup test audit conducted on June 20-22, 1989, the NRC requested TU Electric to provide additional information on Control Rod Reactivity Worths. . The following advance FSAR change for 0&R 423.16 and 0&R 423.31 clarifies the text to indicate that the verification of control rod reactivity worths will be accomplished by either bank exchanges or by boron concentration.
In order to. facilitate NRC staff review of these changes, the enclosed changes are organized as follows:
- 1. Draft revised FSAR pages, with changed portions indicated by a bar in the margin, as they are to' appear in a future amendment (additional pages immediately preceding and/or following the revised pages are provided if needed to understand the. change).
- 2. Line-by-line description / justification of each item revised.
- 3. A copy of related SER/SSER sections.
- 4. An index page containing the title of " bullets" which consolidate-and categorize similar individuel changes by subject and related SER section.
- 5. A discussion of each " bullet" which includes:
- The bold / overstrike version of the revised FSAR pages referenced by the description / justification for each item identified above. The bold / overstrike version facilitates review of the revisions by highlighting each addition of new text in bold type font and overstriking with a slash
(/) the portion of the text that is deleted.
1
- m. _..__ _ ._ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _
': - ... Attachment l'to TXX-89471 1
' ' July 31, .1989 - -
g 'y Page 2 o.f 11
~', '
- CPSES/FSAR
' DRAFT 3b. The Control Rod Reactivity Worths Test Summary
-(Table 14.2-3, sheet 15 of 35), has been revised to state that the worth of the control and shutdown banks sha11 be verified by either bank exchange or b by boron concentration exchange.
DRAFT 4. The Loss of Offsite Power Test Summary (Table 14.2- i 3, sheet 18), has been revised to state that the generator output is at approximately 10%.
DRAFT M The transient shall be initiated by a manual !
I turbine trip and startup transformer isolation in order to simulate a loss of turbine generator
. coincident with a loss of all offsite power.
1 ^&'
l 76 5. The Rod Drop Test Summary (Table 14.2-3, sheet 19).
has been revised to clarify the plant conditions at the time of the tests and to describe the additional te' ting for the fastest and slowest dropped rods.
- 6. Flux Distribution Measurement Test:
1 76 A flux map shall be obtained at the all rods out (ARO) centrol rod configuration.
1 423-42
Attachment I to TXX-89471 July ~ 31, 1939 ' )
age 3 of 11
_, , CPSES/FSAR Q423.31 The response .to item 423.16, part 3, is not acceptable
'in that it only states who determines which RCCA is most reactive, and not how it is selected. Is this r.)d determined to be the most reactive one experimentally.or analytically?
R423.31 .The most reactive RCCA will be determined analytically.
If the bank exchange method is used to verify control DRAFT rod reactivity worths, no most reactive rod worth measurement can be performed.
l l
423-63
-Attachment I to TXX-39471
. Paga.4 of.11 DETAILED DESCRIPTION .
FSAR Page (as amended) Group, Description i
Q&R 423-42, 63 4 O&R 423.16 and 423.31 are clarified to provide )
consistency with a previous FSAR amendment to T14.2-3 Sheet 15.
Clarification:
R423.16, Item 3b was revised to indicate the prcoer T14.2-3' sheet reference and that the worth of the control rod and shutdown banks shall be verified by j either bank exchange or by boron concentration in lieu -
of measurement. R423.31 was revised to reflect that the most reactive RCCA can r.ot be measured if bank exchange is used to verify rod worth. This additional information has been provided per an NRC request during the audit conducted on June 20-22, 1989.
FSAR Change Request Number: 89-508 Commitment Register Number:-NL-2291 AND NL-3884 Related SER Section: 14 SER/SSER Impart: No l
i 1
Attachment 111to TXX-89471 July 31,1989 Page 5 of 11 14 ' INITIAL TEST PR0 GRAM The testing'activit'fes to be performed on safety related systems at Comanche Peak are divided into three major phases: prerequisite testing, preoperational )
i testing, and initial startup testing. ,
Prerequisite testing will be conducted to verify the integrity, proper installa- I tion, cleanliness, and functional operability of the system components.
Preoperational testing will be performed to demonstrate the capability of systems, structures, and components to meet safety-related performance require-ments. These tests will be performed on plant systems, structures, and compo-nents that are designed to perform a nuclear safety-related function. Preopera- L tional testing will be completed before fuel loading with certain limited ;
exceptions where tests or parts of tests will be deferred until the core has l been loaded. In such cases, sufficient testing will be performed before fuel
]
loading to provide reasonable assurance that the postloading tests will be 1 successful.
Initial startup tests will be performed beginning with fuel loading and ending with commercial operation. The intent of these tests is to ensure that fuel loading is effected in a safe manner; that the plant is safely brought to rated capacity; that plant performance is satisfactory in terms of established design criteria; and to demonstrate, where practical, that the plant is capable of withstanding anticipated transients and postulated accidents. The staff review concentrated on the administration of the test program and the complete-ness of the prerequisite, preoperational, and startup. tests. For example, the staff's Safety Evaluation Report issued at completion of the Construction Permit review (CP SER) was re-examined to determine the principal design criteria for the plant and to identify any specific concerns or unique design features that would warrant special test consideration. Chapters 1 through 12 of the FSAR were reviewed for familiarization with the facility design and nomenclature. Chapters 13 and 17 were reviewed for familiarization with the applicant's organizational structure, qualifications, administrative controls, and quality assurance program as they apply to or impact the initial test program. Chapter 15 was reviewed to identify assumptions pertaining to perform-
'ance characteristics that should be verified by testing and to identify all structures, systems, components, and design features that were assumed to function (either explicitly or implicitly) in the accident analysis. Licensee Event Reports for operating reactors of similar design were reviewed to identify potentially serious events and chronic or generic problems that might warrant special test consideration. Standard Technical Specifications for Westinghouse PWRs were reviewed to identify all structures, systems, and components that would be relied upon for establishing conformance with safety limits or limiting conditions for operations. Finally, the Startup Test Reports for other PWR plants were reviewed to identify problem areas that should be emphasized in the Comanche Peak initial test program.
The object of the staff review of FSAR Chapter 14 is to determine whether the acceptance criteria stated in the Standard Review Plan are met. The staff review covered several ;spects of the initial test program including the following major considerations:
14-1
'Tb '?AttachmentlitoTXsB9d1 l
' july' 31,;1989 :
1Page 6 offil
<j. ,.
-(1) 'The appliqant's organization-and staff for performing the. initial test
- program were reviewed. . The staff concludes that an adequate number of-appropriately qualified personnel are assigned to develop test procedures, conduct the tests, and review the results of-the tests. Plant staff
. personnel are utilized to maximize the training benefits of the test
- program.
(2) The applicant stated that the test procedures were developed using input from the NSSS vendor, the architect-engineer, the applicant's engineering staff, and:other equipment suppliers and contractors as needed. We applicant also stated that a review of operating experiences'at similar plants was factored into the development of the test procedures.
~(3) The applicant stated that the tests are being conducted using approved test procedures and that administrative controls cover (a) the completion of test prerequisites, (b) the completion of necessary data sheets and other documentation, and (c) the review and approval of modifications to test procedures. The applicant stated that administrative procedures also cover implementation of modifications or repair requirements identified as being required by the tests and any necessary retesting. _
(4). The applicant stated that the results of each test are reviewed for
~
technical adequacy and completeness by qualified personnel, including NSSS vendor and architect-engineer personnel as appropriate. Preopera-tional test results are reviewed before fuel loading, and the startup .
test results from each activity or power level will be reviewed before.
the next activity or power level.
(5) TheapplicantstatedthatnormalplantoperatNgandemergencyprocedures are used in performing the initial test program, thereby verifying the correctness of'the procedures to the extent practical.
(6) The applicant's schedule for conducting the initial test program allowed I adequate time to conduct all preoperational tests and startup tests. The sequential schedule for performing the startup tests established (a) that systems required to prevent, limi", or mitigate the consequences of .
postulated accidents will be tested before 255 of rated power is exceeded I and_(b) that the safety of the plant will not be dependent on the perfor-mance of untested systems, structures, and components. Preoperational test procedures will be available for IE review at least 30 days before j the exsected performance of the test, and startup test procedures will be <
availasle~at least 90 days before fuel loading.
! (7) The abstract of each test procedure presented in Chapter 14 of the FSAR ,
was reviewed. The staff verified that there are test abstracts for those J structures, systems, componants, and design features that: (a) will be used for shutdown and cooldown of the reactor under normal plant condi-tions ad for maintaining the reactor in a safe conditbn for an extended shutdown period; (b) will be used for shutdown and cooldown of the reactor under transient (infrequent or moderately frequent events) conditions and postulated accident conditions and for maintaining the reactor in a safe condition for an extended shutdom period following such conditions; a (c) will be used for estabifshing conformance with safety limits of '
14-2 i
^
. _ f Attachmen't 1 to ~ TXX-89471 1 ; July ~ 31,1989 Page 7 of 11- a
- limiting' conditions for o
' technical specifications;peration'that will be included in the facility (d) ar
'or will be-relied on to support or ensure the operations of engineered ~
. safety features within design Ifmits; (e) are assumed to function or for which credit is: taken in the accident analysis of the facility, a.s described in the FSAR; and (f) will be used to process, store, control, or limit the release of radioactive materials.
'(8). The test objective, prerequisites, test methods, and acceptance criteria
. for each test abstract were reviewed in sufficient detail to establish that the functional' adequacy of the structures, systems, components, and design' features will be demonstrated.
(9) The test program's conformance with applicable Regulatory Guides was reviewed. The review included: Regulatory Guides 1.E0,1.41,1.52, 1.68, 1.68.2, 1.79, 1.80, and 1.108.
The applicant made a number of changes to the initial te.st program because of the staff's comments. Examples of these changes' include (1) Administrative controls were added to ensure that all test procedure 4 modifications that alter the acceptance criteria or test intent would be appropriately reviewed.
(2) Additional (five consecutive, cold quick starts) testing of the steam-driven auxiliary feedwater pumps was added to further demonstrate. system reliability.
(3) Testing to verify that the reactor cooled pipe penetration cooling system will maintain the pipe tunnel concrete temperature within design limits was added.
(4) The acceptance criteria for certain preoperational and startup tests involving reactor protection system hardware instrumentation delay time, remote _ plant shutdown validation, diesel generator starting requirements, snubber inspection, battery charger capacity, containment spray flowpath verification, sampling system flowpath and holdup, and power-operated relief valve capacities were modified. Modified acceptance criteria were required to more accurately reflect the actual test conditions and to provide assurance that system performance will be in conformance with design predictions.
(5) The minimum qualifications for personnel who direct or supervise the conduct of testing were upgraded to conform with acceptable industry standards. .
(6) Certain system tests were expanded to ensure that comprehensive system and component testing was scheduled. Example systems included station service water, component cooling water, vent and drain, spent fuei pool cooling, residual heat removal, chemical and volume control, safety injection containment ventilation, diesel generator compartment ventila-tion, ac a,nd de power distribution, reactor protection, rod control, l steam generator safety and relief valves, main steam and feedwater isolation valves, auxiliary feedwater, pressurizer safety and relief valves, and engineered safety features.
14-3 L
' Attachment b to TXX'-89471 .v l' Quly 31c 1989
' Page 8 o,f 11 Based on its review, including the' items discussed above, the staff concludes that the initial test program described in the application meets the acce)t-ance criteria of SRP Section 14.2 and that the successful completion of tie program will demonstrate the functional adequacy of plant structures, systems,
.and components. The staff also has concluded that the initial test program ),
described meets the test requirements of GDC 1 and Section XI of 10:CFR 50 Appendix B. '
l 14-4
asesumes a av usewne t
. , July 31,1989 i '
Page 9 o.f 11 SECTION 14.0 - INITIAL TEST PROGRAM l! POEB 13. The FSAR has been revised to clarify 0&R423.16 and 423.31 to (77) l' provide consistency with a previous FSAR amendment to T14.2-3, Sheet 15, Control Rod Reactivity Worth Test Summary. No SER impact.
o l
I
AEtachment~1to5XX-89471- Il
,. ;)uly 31,:1989 y .Page 10,of 11'
. CPSES/FSAR U? .
3b. The Control Rod Reactivity Worths Test Summary (Table 14.2-3, sheet 15 3 of 35 23), has been revised to state.that the worth of the control and shutdown banks shall be verified by either bank exchange or by boron concentration exchange
.ddddd/dd.
-4. The Loss of Offsite Power Test Summary (Table 14.2-3,' sheet 18 16 6f 23), has been revised to state NG+ Port 0F .that the generator output is at approximately 10%.
his $ubrn60(M The transient shall be initiated by a manual turbine trip and startup transformer isolation tidit6f ttif 4HitH dill ir\itidtd d taf61ridI
\ dddd/dtst t/ip in order to simulate a loss of
~
turbine generator coincident with a loss of all offsite power.
- 5. The Rod Drop Test Summary (Table 14.2-3, sheet 19). 76 has been revised to clarify the plant conditions at the time of the tests and to describe the additional testing for the fastest and slowest dropped rods.
- 6. Flux Distribution Measurement Test:
A flux map shall be obtained at the all rods out 76 (AR0) control rod configuration.
423-42
y^
.LAtt'achment 1:to;TXX-89471 -
- 4 '
- July' 31,1989. ~
m Page 11Lof 11
~CPSES/FSAR:
,. Q423.31' 'The response to item.423.16, part 3, is not acceptable
,,t in that it only; states who determines which RCCA is most
-reactive, and not how it'is selected. Is this rod-determined to be'the most reactive one experimentally or analytically?.
R423.31- The most reactive RCCA will be determined analytically.
If the bank exchange method is used to verify control rod reactivity worths, no most reactive rod worth 'i -
measurement'can be performed.
. i 1
L f
423-66 L. _ - .___-________-_-_-__a
l .- :
l J,
' Attachment 2 to TXX-89471' July 31, 1989 Page 1 of 11 1, ?
ADVANCE FSAR CHANGES l'
In the preoperational and startup test. audit conducted on June 20-22, 1989, the NRC requested that TU Electric provide additional information on' Loss of Offsite Power. The following advance FSAR changes submitted for T14.2-3.
- Sheet 18 and O&R 423.16 provides text corrections which provide. consistency ;
with the planned method of test performance. '
In order to facilitate NRC staff review of these changes, the enclosed changes-are organized as follows:
- 1. Draft revised FSAR pages, with changed portions indicated by a bar in the margin, as they are to' appear in a future amendment (additional pages immediately preceding and/or following the ;
' revised pages are provided if needed to understand the change). l
- 2. Line-by-line description / justification of each item revised.
i
- 3. A copy of related SER/SSER sections. !
i 4. An index page containing the titie of " bullets" which consolidate i .and categorize similar individual changes by subject and related l SER section. !
1
- 5. A discussion of each " bullet" which includes:
The bold / overstrike version of the revised FSAR pages referenced by the description / justification for each item J identified above. The bold / overstrike version facilitates ,
review of the revisions by highlighting each addition of I new text in bold type font and overstriking with a slash
(/) the portion of the text that is deleted.
l l
l l
1 i
l;
L ' Attachment 2fto;TXX-89471 CPSES/FSAR i _ ; doly; 31,1989 Table 14.2-3 L
Page 2* of 11*
(Sheet 18)
LOSS OF OFFSITE POWER l1 TEST
SUMMARY
CPSES OBJECTIVE Q423.16 l To demonstrate the proper. plant response following a plant trip with 6 no offsite power available.
PRERE0VISITES Q423.16 The Turbine-Generator output is approximately 130 MWe (greater than 76 10% reactor power) with non-Class IE buses being supplied from the unit auxiliary transformer and the Class IE buses being supplied from their offsite power source.
. TEST METHOD
- 1. Manually generate a main turbine trip and isolate offsite power DRAFT sources.
- 2. . Verify proper starting and load sequencing of diesel generators and transfer of power supplies fer all re:;uired equipment.
Q423.16 0010.18
- 3. Verify the Reactor Coolant Sy, tem can be maintained in a shutdown 6 condition for a minimum or 30 minutes utilizing the power operated atmospheric relief valves to remove decay heat from the reactor core.
ACCEPTANCE CRITERIA 0423.11 The on-site power supplies (i.e., diesel generators) shall auto-start DRAFT and operate the necessary controls, equipment and indication to remove i decay heat and maintain the Reactor Coolant System in a shutdown condition for the duration of the test.
~
li : Attachment 2'to(TXX-89471 l( Nuly. 31,1989' -
E [Page 3 of 11: .
' *- x CPSES/FSAR
. DRAFT 3b. .The Control Rod Reactivity Worths Test Summary
. N W PNJ.TOF (Tabie 14.2-3 sheet is of 3s), has been revised to TU NNb state that the worth of the control and shutdown banks shall'be verified by either bank exchange or by boron concentration exchange.
V ._-
DRAFT 4. The Loss of Offsite Power Test Summary (Table 14.2-3, sheet 18), has been revised to state that the generator output is at approximately 10%.
DRAFT The transient shall be initiated by a manual turbine trip and startup transformer isolation in order to simulate a IcIs of turbine generator coincident with a loss of all offsite power. !
- 76. 5. The Rod Drop Test Summary (Table 14.2-3, sheet 19),
has been revised to clarify the plant conditions at the. time of the tests and to describe the additional testing for the fastest and slowest dropped rods.
- 6. Flux Distribution Measurement Test:
76 'A flux map shall I..e obtained at the all rods out (ARO) control rod configuration.
423 42 an_________________. -
'kitachm'ent2'to.TXX-89471- CPSES FSAR AMENDMENT i
- July 31,19.89- DETAILED DESCRIPTION Page 4 of 11
'FSAR Page (as amended) GroUD Description' Table 14.2-3 3 See Sheet No(s):18 Corrects the Test Method and Acceptance Criteria sections of the Loss of Offsite Power Test Summary.
Correction:
The T6st Method has been corrected as a result of E
previous system modifications and the identification of a more preferable method of test performance. The Acceptance Criteria has been corrected to be consistent with the planned method of test performance.
FSAR Change Request Number:-89-510.1 Related SER Section: 14 SER/SSER Impact: No O&R 423-42 4 Editorial change provides consistency with FSAR Table 14'.2-3. Sheet 16 of 23 has been changed to sheet 18.
Clarification:
- FSAR Change Request Number: 89-510.3 Related SER Section: 14 SER/SSER Impact: No LO&R 423 42 3 Corrects the description regarding the performance of the Loss of Offsite Power Test.
Correction:
The correction is necessary as a result of previous system modifications and the identification of a more preferable method of test performance.
FSAR' Change Request Number: 89-510.4 Related SER Section: 14 SER/SSER Impact: No '
MGiMER 3)uly 31,- 1989 Paije 5 of 11 14 INITIAL TEST PROGRAM
~
The testing activities to be performed on safety-related systems at Comanche Peak are divided into three major phases: prerequisite testing, preoperational testing, and initial startup testing.
Prerequisite testing will be conducted to verify the integrity, proper installa-tion, cleanliness, and functional operability of the system components.
Preoperational testing will be performed to demonstrate the capability of systems, structures, and components to meet safety-related performance require-ments. These tests will be performed on plant systems, structures, and compo-nents that are designed to perform a nuclear safety-related function. Preopera-l tional testing will be completed before fuel loading with certain limited exceptions where tests or parts of tests will be deferred until the core has been loaded. In such cases, sufficient testing will be perfonned before fuel loading to provide reasonable assurance that the postloading tests will be successful.
l Initial startup tests will be performed beginning with fuel loading and ending '
with commercial operation. The intent of these tests is to ensure that fuel loading is effected in a safe manner; that the plant is safely brought to rated capacity; that plant performance is satisfactory in terms of established design criteria; and to demonstrate, where practical, that the plant is capable of withstanding anticipated transients and postulated accidents. The staff review concentrated on the administration of the test program and the complete-ness of the prerequisite, preoperational, and startup, tests. For example, the staff's Safety Evaluation Report issued at completion of the Construction Permit review (CP SER) was re-examined to detemine the principal design criteria for the plant and to identify any specific concerns or unique design features that would warrant special test consideration. Chapters 1 through 12 of the FSAR were reviewed for familiarization with the facility design and nomenclature. Chapters 13 and 17 were reviewed for familiarization with the applicant's organizational structure, qualifications, administrative controls, and quality assurance program as they apply to or impact the initial test program. Chapter 15 was reviewed to identify assumptions pertaining to perfom-ance characteristics that should be verified by testing and to identify all structures, systems, components, and design features that were assumed to function (either explicitly or implicitly) in the accident analysis. Licensee i Event Reports for operating reactors of sinflar design were reviewed to identify potentially serious events and chronic or generic problems that might warrant special test consideration. Standard Technical Specifications for Westinghouse PWRs were reviewed to identify all structures, systems, and components that would be relied upon for establishing conformance with safety limits or limiting conditions for operations. Finally, the Startup Test Reports for other PWR plants were reviewed to identify. problem areas that should be emphasized in the Comanche Peak initial test program.
The object of the staff review of FSAR Chapter 14 is to detemine whether the acceptance criteria stated in the Standard Review Plan are met. The staff review covered several aspects of the initial test program including the following major considerations:
1 14-1
IAttachment 2 - to . TXX-BM71 J.uly 31,:1989 Pa'ge 6. df 11-(1) The appliqant's organization and staff for performing the initial test program were reviewed. The staff concludes that an adequate number of appropriately qualified personnel are assigned to develop test procedures, conduct the tests, and review the results of the tests. Plant staff personnel are utilized to maximize the training benefits of the test program.
(2) The applicant stated that the test procedures were developed using input from the NSSS vendor the architect-engineer, the applicant's en staff,andotherequIpsentsuppliersandcontractorsasneeded.gineering The (
applicant also stated that a review of operating experiences at similar 1 plants was factored into.the development of the test procedures.
(3) The applicant stated that the tests are being conducted using approved test procedures and that adelnistrative controls cover (a) tlw completion of test prerequisites, (b) the cogletion of necessary data sheets and other documentation, and (c) the review and approval of modifications to test procedures. The applicant stated that administrative procedures also cover implementation of modifications or repair requirements 3' identified as being required by the tests and any necessary retesting.
l (4) The applicant stated that the results of each test are reviewed for technical adequacy and completeness by qualified personnel, including l NSSS vendor and architect engineer personnel as appro Proopera-l
.tional test results are reviewed before fuel loading,priate.
and the startup .
test results from each activity or power level will be reviewed before l
the next activity.or power level.
i (5) The applicant stated that nomal plant operating and emergency procedures are used in perfoming the initial test program, thereb correctness of the procedures to the extent practical. y verifying the 1 (6) The app 1tcant's schedule for conducting the initial test program allowed adequate time to conduct all preoperational tests and startup tests. The sequential schedule for perfo ming the startup tests established (a) that systems required to prevent, limit, or sitigete the consequences of postulated accidents will be testad before 2 5 of rated power is exceeded and (b) that the safety of the plant will not be dependent on the perfor-eence of untested sys'tems, structures, and cosponents. Prooperational test procedures.will be available for IE review at least 30 days before the egocted performance of the test, and startu l available at least 90 dqys before fuel loading. p test procedures will be (7) The abstract of each test procedure presented in Chapter 14 of the FSAR was reviewed. The staff verified that there are test abstracts for those I structures, systans, components and design features that: (a) will be used for shutdown and cooldown o,f the reactor under nomal plant condi-tions t.nd for maintaining the reactor in a safe condition for an extended shutdown period; (b) wil be used for shutdown and cooldown of the reactor under transient (infrequent or moderately frequent events) conditions and postulated accident conditions and for maintaining the reactor in a safe condition for an extended shutdown period following such conditions; (c) will be used for establishing conformance with safety limits of 14-2
_ _ _ - _ _ _ _ _ _ _ _-- _ _ )
Attachment 2 of TXX-89471 l
. July '3},1989 Page 1 of 11 ,
limiting conditions for operation that will be included in the facility technical specifications; (d) are classified as engineered safety features or will be relied on to support or ensure the operations of engineered safety features within design limits; (e) are assumed to funct'on or for which credit is taken in the accident analysis of the facility, ?,s described in the FSAR; and (f) will be used to process, store, control, or limit the release of radioactive materials.
(8) The test objective, prerequisites, test methods, and acceptance criteria for each test abstract were reviewed in sufficient detail to establish that tk functional adequacy of the structures, systems, components, and design features will be demonstrated.
(9) The test program's conformance with applicable Regulatory Guides was 4 reviewed. The review included: Regulatory Guides 1.20, 1.41, 1.52, 1.68, 1.68.2, 1.79, 1.80, and 1.108.
J The applicant made a number of changes to the initial test program because of the staff's comments. Examples of these changes include (1) Administrative controls were added to ensure that all test procedure modifications that alter the acceptance criteria or test intent would be appropriately reviewed.
(2) Additional (five consecutive, cold quick starts) testing of the steas-driven auxiliary feedwater pumps was added to further demonstrate system reliability.
(3) Testing to verify that the reactor cooled pipe penetration cooling system will maintain the pipe tunnel concrete tasperature within design faits was added.
(4) The acceptance criterfa for certain preoperational and startup tests involving reactor protection system hardware instrumentation delay time, remote p ant shutdown validation, diesel p nerator starting requirements, snubber inspection, battery charger capac<ty, containment spray flowpath verification, sampling system flowpath and holdup, and power-operated relief valve capacities were modified. Modified acceptance criteria were required to more accurately reflect the actual test conditions and to provide assurance that system performance will be in confonnance with design predictions.
(5) The minimum qualifications for personnel who direct or supervise the conduct of testing were upgraded to conform with acceptable industry standards. .
(6) Certain system tests were expanded to ensure that comprehensive system and component testing was scheduled. Example systans included station service water, component cooling water, vent and drain, spent fuel pos1 cooling, residual heat removal, chemical and volume control, safety injection containment ventilation, diesel generator com tion, ac a,nd de power distribution, reactor protection,rodpartment control, ventila-steam generator safety and relief valves, main steam and feedwater isolation valves, auxiliary feedwater, pressurizer safety and relief valves, and engineered safety features.
14-3
~ W hac @enf f To M -WbS7F ~ l
. . July 31,1989 Page 8 of'11. j
'liased on its review, including the itses discussed above, the staff concludes that the initial test progras described in the application meets the accept-ance criteria of SRP Sect on 14.2 and that the successful completion of the i
program will demonstrate the functional adequacy of slant structures, systems, !
and components. The staff also has concluded that t w initial test progras described meets the test requirements of GDC 1 and Section XI of 10 CFR 50 l l
Appendix 8.
l
{
1 i
. I i
i 14-4 t'
ww e w w-zum u July'3},'1989
3 f m f Pa'ge!9 of 11'
^
,; . 7; . .
SECTION 14.0.- INITIAL TEST PROGRAM POEB 14. The FSAR has-been revised to correct T14.2-3.' Sheet 18. and _
(77)
< O&R 423;16 as a result of-previc4s, system modifications and the identification of a more preferti.ble method of-test performance.
No SER~ impact; l , 10, '
l i
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Page 10 of il .
Table 14.2-3 (Sheet 18)
LOSS OF 0FFSITE POWER TEST
SUMMARY
CPSES OBJECTIVE Q423.16 To demonstrate the proper plant response following a plant trip with 6 no offsite power available.
PRERE0VISITES Q423.16 The Turbine-Generator output is approximately 130 MWe (greater than 76 10% reactor power) with non-Class IE buses being supplied from the unit auxiliary transformer and the Class IE buses being supplied from their offsite power source.
TEST METHOD
- 1. Manually generate a main turbine trip and isolate /gggggi gf fff / 75 fjff M M fffff $$1f$1 $ M offsite power sources.
- 2. Verify proper starting and load sequencing of diesel generators and transfer of power supplies for all required equipment.
Q423.16 0010.18
- 3. Verify the Reactor Coolant System can be maintained in a shutdown 6 condition for a minimum of 30 minutes utilizing the power operated atmospheric relief valves to remove decay heat from the reactor Core.
ACCEPTANCE CRITERIA Q423.11 The on-site power supplies (i.e., diesel generators) shall auto-start 6 and JH ffMjftf ff fM MJJ ffMf $$ffff fff# fM fgff MMfffff 16 /
IM fM11H ffM M/ M Mf $Mfif 11 /H1/fttiffill MfMtifM M/ 7H i 1
- //M #fMM/ ff#f /#f// operate the necessary controls, equipment and indication to remove decay heat and maintain the Reactor Coolant System in a shutdown condition for the duration of the test.
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, .AttachmbntL2 to TXX-89471 '
( fJuly 3111989 ' '
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fage 11 of sL3L CPSES/FSAR' !
13b. .The Control Rod Reactivity Worths Test Summary
. NOT PART OF .
-(Table 14.2-3. sheet 15 3 of 35 23), has been
. % l 3 C )4 A f\ L E - revised to state.that the worth of the control and shutdown banks shall .be verified by either bank exchange or by boron concentration exchange
- ddisfid.
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- 4. The Loss of Offsite Power Test Summary (Table 14.2-3, sheet 18 16 ff 23), has been revised to state that the generator output is at approximately 10%.
The transient shall be initiated by a manual turbine trip and startup transformer isolation-finitst tfif 4KilK dill initidtd 6 taf6fddl dddd/df6/ flip in order to simulate a loss of turbine generator coincident with a loss of all offsite power.
l l 5. The Rod Drop Test Summary (Table 14.2-3, sheet 19), 76 has been revised to clarify the plant'conditices at the time of the tests and to describe the additional testing for the fastest and slowest 1
dropped rods.
- 6. Flux Distribution' Measurement Test:
1 A flux map shall be obtained at the all rods out 76 (AR0) control rod configuration.
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l 423-42
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