ML20010C093
| ML20010C093 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek, Callaway  |
| Issue date: | 08/07/1981 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Byran J NORTHEAST NUCLEAR ENERGY CO. |
| References | |
| NUDOCS 8108190112 | |
| Download: ML20010C093 (29) | |
Text
w cy iy s
f-9
,Y g'
LA g
6-
_Dist.
- Docket File-WJohnston AUG 7 1981 LB#1 Rdg WHouston BJYoungblood AChu GEdison KDempsey MRushbrook bcc:
RVollmer TERA ACRS (16)
Docket Nos.: 'STN 50-482; TMurley NRC/PDR and STN 50-483 RMattson L/PDR RHartfield, MPA NSIC TIC Mr. John K. Bryan Mr. Glenn L. Koester Vice President Vice President - Nuclear Union Electric Company Kansas Gas and Electric Company 1901 Gratiot Street 201 North Market Street Post Office Box 149 Post Office Box 208 St. Louis, Missouri 63166 Wichita, Kansas 67201
Dear Gentlemen:
Subject:
SNUPPS FSAR - Request for Additional Information -
Inttial Test Program As a result of our review of your application for operating licenses we find that we need additional information regarding the SNUPPS FSAR. The specific information required is as a result of our review of the initial test program and is listed in the Enclosure.
To maintain our licensing review schedule for the SNUPPS FSAR, we will need responses o the enclosed request by August 21, 1981.
If you cannot meet this date, please inform us within seven days after receipt of this letter of the date you plan to submit your responses so that we may review or schedule for any necessary changes.
Please contact Dr. G. E. Edison, SNUPPS Licensing Project Manager, if you disire any discussion or clarification of the enclosed request.
Sincerely, h
Robert L. Tedesco, Assistant Director for Licensing Division of Licensing
Enclosure:
4 D
As stated p
cc: See next page h
liU G 1 7 iggy,,.-
8108190112 810807 4 U.a, %,,,,
J PDR ADOCK 05000482 g
8'aw eu o,.
x m
O[ \\!
/
N
'"C 4. U.L.:.l.B f.l.
.Uk
/
t cuamp GEdison/ys BJogglood.........[P M.,,f c
M 8/,h,/8,1.,,
,,8/,
,,Q 1,,,
,8L 1,,
2nc ronu aie no em wncu o240 ceFICIAL RECORD COPY
> = = -32"24
3 Mr. J. K. Bryan Mr. Glenn L. Koester Vice President - Nuclear Vice President - Nuclear Union Electric Company Kansas Gas and Electric Company P. O. Box 149 201 florth Market Street St. Louis, Missouri 63166 P. O. Box 208
. Wichita, Kansas 67201 cc: Gerald Charnoff, Esq.
Shaw, Pittman, Potts, Dr. Vern Starks Trowbridge & Madden Route 1, Box 863 1800 M Street, N. W.
Ketchikan, Alaska 99901 Washington, D. C.
20036 Mr. William Hansen
' Kansas City. Power & Light Company U. S. Nuclear Regulatory Commission ATTN: Mr. D. T. McPhee Resident Inspectors Office Vice President - Production RR #1 1330 Baltimore Avenue S'; edman, Missouri 65077 Kansas City, Missouri 64141 Ms. Treva Hearn, Assistant General Counsel Mr. Nicholas A. Petrick Missouri Public Service Commission Executive Director, SNUPPS P. O. Box 360 5 Choke Cherry Road Jefferson City, Missouri 65102 Rockville, Maryland 20850 Jay Silberg, Esquire Mr. J. E. Birk Shaw, Pittman, Potts & Trowbridge Assistant to the General Counsel 1000 M Street, N. W.
Union Electric Company Washington, D. C.
20036 St. Louis, Missouri 63166 Mr. D. F. Schnell Kansans for Sensible Energy Manager - Nuclear Engineering P. O. Box 3192 Union Electric Company Wichita, Kansas 67201 P. O. Box 149 St. Louis, Missouri 63166 Ms. Mary Ellen Salava Route 1, Box 56 Mr. Tom Vandel Burlington, Kansas 66839 Resident Inspector / Wolf Creek flPS c/o USNRC Eric A. Eisen, Esq.
P. O. Box 1407 Birch, llorton, Bittner & Monroe Emporia, Kansas 66801 1140 Connecticut Avenue,' N. W.
Washington, D. C.
20036 Mr. Michael C. Keener Wolf Creek Project Director Sta te Corporation Commission Ms. Wanda Christy State of Kansas 515 N. Ist Street Fourth Floor, State Office Building Burlington, Kansas 66839 Topeka, Kansas 66612
ENCLOSURE SNUPPS - Page 1 of 27' QUESTIONS - SNUPPS
~
640.0 Procedures and Test Review Branch 640.1 Certain exceptions to regulatory guides as listed in Appendix (14. 2.7) 3A are not acceptable or require further justification.
Provide the following infonnation:
(1) Regulatory Guide 1.68 Describe existing tests that verify acceptable plant response for a loss of turbine-generator coincident with a loss of offsite power, or delete this exception and include the appropriate test description.
(2) Regulatory Guide 1.80 State which tests demonstrate that safety-related valves fail-safe on loss-of-instrument air.
(3) Regulatory Guide 1.118 The discussion states that nuclear instrumentation sensors are exempt from time response testing since their worst
' case response time is not a significant portion of the total overall system response (i.e., less than 5%).
Given that this exemption is no longer permitted by IEEE-338 (1977 version), delete this exception or provide expanded technical justification for not conducting time response testing.
640.2 Your initial criticality description should be expanded to (14. 2.10. 2 )-
include:
1.
A source range count of at least 1/2 count per second should be visible on the startup channels prior to commencing the startup.
a 2.
The signal to nois.e ratio should be known to be greater than 2.
. 3.
Criticality predictions for boron concentration and control rod positions should be provided, and criteria Tnd c tions to be taken should be established if actual plant conditions deviate from predicted values.
4.
The approach to criticality shuula De slow enough to limit start up rate at criticality to less than 1 decade per minute.'
~
640.3 Section 14.2.11 of SNUPPS states that insofar as practicable, (14.2.11) test requirements will be completed prior to exceeding 25-percent power for all plant structures, systems and components that are relied upon to prevent, limit or mitigate the consequences of postulated accidents. 'According to Table 14.2-5 the following,startup tests are perfomed after exceeding 25-percent power:
(1) S070012 - Rod Drop and Plant Trip
. (2) S07AB01 - Automatic Steam Generator Level Control (3) S07SF05 - Automatic Reactor Control System (4) S07SF07 - Startup Adjustments of Reactor Control System.
Perform these tests at 25% power or less, or provide technical justification for not fulfilling the testing requirements of Section 14.2.11.
640.4 Saction 14.2.11 of SNUPPS states that startup test procedures (14.2.11) will be available for NRC review at least 60 days prior to fuel loading.
Table 14.2-5 indicates that twenty of thirty-eight 3tartup tests will be in the procedure preparation, review and approval stage at that time.
Modify Table 14.2-5 to indicate by a note or legend alteration that complete procedures will be available for review in the time frame stated in Section 14.2.11.
ShUPPS - Page 3 of 27 4
640.5 Provide a commitment to include in your test program the (14.2.12) design features to prcvent or mitigate anticipated transients without scram (ATWS) that may now, or in the future, be incorporated.into your plant design (Subsv. tion.15.8).
640.6 List those tests that will only be performed on the first (14.2.12)
SNUPPS unit.
In addition c,ite the criteria that will be used during subsequent unit testing programs to ensure that follow-on units perform in an identic.al manner regarding those tests to be deleted.
640.7 Identify any of the post-fuel loading tests described in (14.2.12.3)
Section 14.2.12.3 which are not essential towards the demonstration of conformance with design requirements for structures, systems, components, and design features that meet any of the following criteria:
(1) Will be relied upon for safe shutdown and cooldown of the reactor under normal plant conditions and for maintaining the reactor in a safe conditicn for an extended shutdown period.
(2) Will be relied upon for safe shutdown anc ccoldown of the reactor under transient (infrequent or j
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.
l (3) Will be relied upon for establishing conformance with
- safety limits or limiting conditions for operation j
that will be included in the facility technical specifications.
(4) Are classified as engineered safety features or will be relied upon to support or assure the operation of engineered safety features within design limits.
SNUPPS - Pag) 4 of 27 a
(5) Are assumed to function or for which credit is taken in the accident analysis for the facility (as described-in the Final Safety Analysis Report).
(6) Will be utilized to process, store, control, or limit-the release of radioactive materials.
640.8 The objectives specified for-several tests are inappropriate.
(14.2.12.3)
In general, appropriate test objectives are:
- to neasure to calibrate e
to obtain data e
e to document to verify performance.
Provide appropriate objectives for the following tests:
14.2.12.3.1 3.2 3.3 3.8 3.22 3.33 3.35 a
n
~
SNUPPS - Pag 5 of 27 i
640.9 It is unacceptable to reference test instructions for (14.2.12.3)
. test prerequisites.
Provide acceptable prerequisites-for the following tests:
14.2.12.3.1 3.4 3.5 3.6
- 3. 7 -
3.8.2.a 3.13 3.14 3.21 3.22 3.23 3.24 3.25.2.a 3.26 3.27 3.29 3.30 4
3.31 3.32 3.33 3.34.2.b 3.35 640.10 Certain terminology used in the individual test descriptions (14.2.12) does r.ot clearly indicate the source of the acceptance criteria to be used in determining test adequacy.
An acceptable format for providing acceptance criteria for test results includes any of the following:
Referencing technical specifications e
Referencing specific sections of the FSAR Referencing vendor technical manuals Providing specific quantitative bounds (only if the information cannot be provided in any of the above ways).
Modify the individual test description subsection presented i
below or, if applicable, add a paragraph to Subsection 14.2.12 tha,t provides an acceptable description of each of the unclear terms.
3 r
~
SNUPPS - Page 6 of 27 1
(1) Within. design specifications 14.2.12.1.3 1.4 1.5 1.7 1.9 1.10 1.11 1.12 1.15 (2 times) 1.18 (2 times) 1.21 (2 times)'
1.23 (2 times) 1.24 1
1.25 (2 times) 1.26 (2 times) 1.27
~
1.28 (3 times) 1.29 (3 times) 1.30 1.32' (2 times) 1.33 (4 times) 1.34 (3 times) 1.36 1.37 (3 times) 1.39 1.41 (3 times).
1.42 (2 times) 1.43 1.44 (2 times) 1.45 (2 times) 1.46 1.47 1
1.48 1.49 1.50 (2 times) 1.51 (2 times) 1.52 1.53 1.59 1.60 (2 times) 1.61 (2 times) 1.62 1.64 (6 times) 1.65 1.66 (2 times) 1.68 (2 times) 1.71 1.72 l
l I
t
,-n.
SNUPPS - Paga 7 of 27
-2.1
~
2.2' (2 times)
~2.3 (2 times) 2.4
- 2. 5 2.6 (2 times) 2.7 2.8 2.10.
2.11 (2 times) 2.14~(2 times) 2.15-2.16 2.19 2.22 (2 times) 2.25 3.15 3.18 (2 times) 3.20 (2 times)-
(2)
In accordance with design, in accordance with syste.1 design 14.2.12.1.1 (2 times) 1.6 (2 times) 1.8 l.44 1.45 1.46 l.48 1.51 1.54 1.55 1.56 (2 times) 1.57 1.58 (2 times) 1.59 1.63 (2 times) 1.64 (4 times) 1.65 (2 times) 1.66 i
1.68 1.69 1.70 1.71 (2 times) 1.72 1.73 2.15 2.16
.e f
-SNUPPS --Paga 8 of'27
- (I)
In.accordance with design specification, in accordance with ' system design specificatioas 14.2.12.1.39
-2.1 2.9
'2.11 2.12 2.13
=2.20 2.21- (2 times )
2.24 2 26
.(4) Design 14.2.12.1.10 1.11 1.17 1.35' i'
-1.42 1.65 (3 times) 1.67 (5 times) 1.70 4
1.80 2.17 2.18 1
3.15 3.17 3.37 (5)' Within design limits, without exceeding design limits, within the limits predicted by design analyses, within design requirements 14.2.12.1.16 (2 times) i 1.29 i
1.32 1.35 1.37 1.41 1.62 1.64 1.73
!~
1.78 1.79 3.16 i
s i
.... U
Sf;UPPS - Paga 9 of-27 (6) Within allowable limits, within required limits
~
14.2.12.1.22 1.38 1.62 (7) Required 14.2.12.1.10 1.22-1.64 (10 times) 1.65 (2 times) 1.85
^
(8) Rated.
14.2.12.1.62 1
1,64 (2 times) 1.65 1.82 (3 times)
(9) Responds, resnonds properly, properly respond 14.2.12.1.12 1.34 1.36 1.48 1.49 4
1.51 (10) In accordance with test instruct.ans, +is provided in test instructions, meets the requirements of _the test instructions, consistent.with the acceptance criteria given in the test procedure, agrees with the acceptance criteria given in the test procedure, as required by the test instructions 14.2.12.1.74 1.75 1.76 3.2 3.6 3;7 3.8 3.11 3.13 3.14 i
.~..
E SNUPPS - Page 10 of.27
- 6..
.3.23 3.30
~3.31 3.32 3.33 (11)' Shall not exceed code-allowable stresses, must not exceed their code-allowable limits at the test or design conditions 14.2.12.1.80 1.81 3.37.(2 times)
(12) Setpoint tolerances.
14.2.12.1.2 (13) Acceptable 14.2.12.1.14 1.64 (2 times) 2.17 i
2.18 (14) Adequate 14.2.12.1.37 1.83 7
(15) Approximate 14.2.12.1.14 1.80 3.37 (16) Predicted 14.2.12.1.14 (17) Verified 14.2.12.1.14 1.22 (18) Fails safe 14.2.12.1.73
,1
. SNUPPS - Pag 2 11 of 27 c
(19) Operate satisfactory per design 14.2.12.1.83 (20) Impair design functions 14.2.12.1.83 (21) Slightly above 14.2.12.1.20 640.11 Our review of your initial test program description disclosed (14.2.12) that the operabil.ity of several of the systems and components-listed in Regulatory Guide 1.68 (Rev. 2), Appendix A, may not be demonstrated.
Expand your FSAR to include appropriate test descriptions (or identify existing descriptions) that address the following items from Appendix A, or provide technical justification for any exceptions to the guide in Subsection 14.2.7:
(1)
Preoperational Testina 1.a.(2)(i)
RCS safety valves 1.b.(1)
Control rod drive system test 4
1.e.(5)
Steam extraction system 1.e.(6)
Turbine stop, control, and intercept valves 1.e.(10)
Feedwater heater and drain systems 1.h.
Test of protective devices such as leak-tight covers, structures, or housings provided to protect Engineered Safety Features from flooding 1.h.(8)
Tanks and other sources of water used for ECCS 1.i.(5)
Containment airlock leak rate test 1.1.(12)
Containment air purification and cleanup system 1.1.(15)
Containment penetration pressurization system tests
SNUPPS - Pag 212 of 27-1.j.(6)
Loose parts monitoring system 1.j (7)
Leak detection systems for ECCS and containment spray systems outside of containment 1.J.(8).
Reactor control system 1.J.(9)
Pressure control systems designed to prevent leakage across boundaries l.j. (11 )
Traversing incore probe system 1.j. (13)
Incore nuclear instrumentation 1.J. (14)
Instrumentation and controls that affect
. transfers of water supplies to auxiliary feedwater pumps, ECCS purps, and containment spray pumps 1.j. (16)
Hotwell level control system 1.J.(17)
Feedwater heater temperature, level, and bypass control systems 1.j.(18)
Auxiliary startup instrument tests 1.j.(20)
Instrumentation used to detect internal and external flooding, 1.j.(22)
Instrumentation than-can be used to track.
i the course of postulated accidents such as containment sump level monitors, and humidity l
monitors 1.j.(24)
Annunciators for reactor control and engineered safety features 1.j.(25)
Process computers 1.1.(4)
Isolation features 'for steam generator blowdown 1.1.(7)
Isolation features for liquid radwaste effluent systems 1.m.(4)
Dynamic and static load testing of cranes, hoists, and associated lifting and rigging equipment, including the fuel cask handling 1
m SNUPPS - Page 13 of 27 O-crane.
Static testing at 125% of rated load and full operational testing at 100%
of rated load 1.n.(2)
Closed loop cooli..g water systems 1.n.(6)
Chemistry control systems.for the reactor coolant and secondary coolant systems 1.n.(9)
Vent and drain systems for contaminated or potentially contaminated systems 1.n. (10)
Purification and cleanup systems for the reactor coolant system 1.n.(12)
Boron recovery system 1.n. (14)(c)
Battery room ventilation 1.n.(16)
Cooling and heating systems for the refueling water storage tank 1.0 Reactor components handling systems (2)
Initial Fuel Load and Precritical Testing 2.a.
Shutdown margin verificacion for the fully loaded core 2.b.
Control rod withdrawal and insertion speeds, sequencers and protective interlocks 2.d.
Final reactor coolant system leak rate test (4) Low Power Testing 4.b.
Confirm by analysis that rod insertion limits will be adequate to ensure a shutdown margin consistent with accident analysis assumptions, with the greatest worth control rod stuck att of the core 4.c.
Pseudo-rod-ejection test 4.e.
Flux distribution determination 4.f.
Neutron and gamma radiation surveys 4.g.
Determination of proper response of process and effluent radiation monitors
SNUPPS - Paga 14 of 27 4.h.
- Chemical and radiochemistry tests 4.1.
Demonstration of the operability of control rod witlidrawal inhibit or block functions over the reactor power level range during which such features must be operable
- 4.,j.
Demonstration of the capability of the primary containment ventilation system 4.n.
Demonstrat' ion of the operability of the control room computer system 4.r.
Demonstration of the operability of reactor coolant system purification and cleanup systems 4.t.
Performance of nttural circulation tests of the reactor coolant system to uetermine that adequate heat removal capability exists.
NUREG-0694 "TMI Ralated Requirements for New Operating Licenses," Item I.G.1, requires applicants to perform "a special low power testing program approved by NRC to be conducteo at power levels no greater than 5 percent for the purposes of providing meaningful technical information beyond that obtained in the normal startup test program and to provide supplemental training." To comply with this requirement new PWR applicants have comitted to a series of natural circulation tests. To data such tests have been performed at the Sequoyah 1, North Anna 2, and Salem 2 facilities.
Based L
or the success of the programs at these plants, the staff has concluded that augmented natural circulation training should be performed for all future PWR operating licenses.
Includes descriptions 1
of natural circulation tests that, in addition to validating the operating procedures, fulfill the following objectives:
SNUPPS - Page-15 of 27 Testing The tests should demonstrate the following plant characteristics:
lur.gth of time required to stabilize naturar circulation, core flow distribution, ability to establish and maintain natural circulation with or without onsite and offsite power, the ability to uniformly borate and cool down to hot shutdown conditions using natural circulation, and subcooling monitor performance.
Training Each licensed reactor operator (R0 or SR0 who performs R0 or SR0 duties, respectively) should participate in the initiation, maintenance and recovery from natural circulation mode.
Operators should be able to recognize when natural circulation has stabilized, and should be able to control saturation margin, RCS pressure, and heat removal rate without exceeding specified operating limits.
If these tests have been performed at a comparable prototype plant, they need be repeated only to the extent necessary to accomplish the above training objectives.
(5)
Power-Ascension Tests 5.b.
Determine that steady-state core performance i
is in accordance with design 5.d.
Demonstrate the capabilities of plant features and procedures for controlling core xenon transients u_
SNUPPS - Paga 16 of 27 5.e.
Pseudo-rod-ejection test
- 5. f.
Single rod insertion and witiArawal 5.g.
Demonstrate operation of the control rod sequencers, and rod withdrawal blo::k functions 5.h.
Check rod scram times from data recorded during the startup test phase.
5.1.
Demonstrate the capability of incore and excore neutron flux instrumentation to detect a control rod misalignment equal to or less than the technical specification limits 5.1.
Demonstrai;e design capability of all systems and components provided to remove residual or decay heat from the reactor coolant system 5.m.
Demonstrate that reverse flows through idle loops and differential pressuras across the core are in agreement with design values 5.n.
Obtain baseline data for reactor coolant system loose parts monitoring system 5.r.
Verification of input to, and output from control room process computer 5.s.
Verify the performance of the auxiliary feedwater control system, the hotwell level control system, steam pressure control system, and the reactor coolant makeup and letdown control systems
- 5. t.
Verify the response times, relieving capacities, and reset pressures for the pressurizer relief valves; main steam line relief valves; atmospheric steam dump valves; and the turbine bypass valves 5.u.
Verify operability and response times of main steam line isolation and branch steam line isolation valves
SNUPPS - Page 17 of 27 5.v.
Verification of main steam system and feedwater system performance 5.w.
Demonstrate that concrete temperatures surrounding hot penetrations do not exceed design limits 5.y.
Verify the proper operation of the incore nuclear instrumentation, and instruments and systems used t' perform a heat balance 5.z.
Demonstrate that process and effluent radiation monitoring systems are responding correctly 5.aa.
Demonstrate the operation of the chemical and radiochemical control systems 5.bb.
Conduct neutron and gamma radiation surveys to establish the adequacy of shielding 5.cc.
Demonstrate the operation of the gaseous and liquid radioactive waste processing, storage, and release systems 5.ff.
Demonstrate that ventilation systems maintain design temperatures 5.11.
Demonstrate that the dynamic response of the plant is in accordance with design for lim'iting reactor coolant pump trips 5 kk.
Demonstrate that the dynamic response of the plant is in accordance wir1 design for the loss of or bypassing of the feedwater heaters 5.mm.
Demonstrate that the dynamic response of the plant is in accordance with design for the case of sutomatic closure of all main steam line isolation valves at 100% reactor power 5.nn.
Demonstrate that the dynamic response of the plant is in accordance with design for the case of full load rejection (tripping of the main generator breakers) a
SNUPPS - Pag) 18 of 27 640.12 We could not conclude from our review of your. individual test (14.2.12) descriptions that comprehensive testing is scheduled for several systems and components.
Therefore, clarify or expand the appropriate test descriptions to address the following items:
(1) 14.2.12.1.1 - Clarify, of reference the FSAR section which clarifies, the purpose of a decreasing condenser pressure signal.
(2) 14.2.12.1.5 - Provide acceptance criteria for steam generator feedwater pump operation.
(3) 14.2.12.1.7 - Subsection 10.4.9.2.3 indicates four separate actuation signals can cause an automatic start of the motor-driven auxiliary feed pump.
Ensure these four are included in your test description acceptance criteria.
(4) 14.2.12.1.8 - Our review of licensee event reports has disclosed several instances of turbine-driven auxiliary feedwater pump failure to start on demand.
It appears that many of these failures could have been avoided if more thorough testing had been conducted during the plant's initial te:t programs.
In order to discover any problems affecting pump startup and to demonstrate the reliability' of your emergency cooling system, state your plans to demonstrate at least five consecutive, successful, cold quick pump starts during your initial test program.
(5) 14.2.12.1.9 - Commit to verifying operation of any pump
_ permissive interlocks which serve to prevent cold water addition accidents or serve to protect RCS components from excessive differential pressures at low temperatures.
~
SNUPPS - Paga 19 of 27 4
(6) 14.2.12.1.17,18 - State that flow and coastdown testing will be performed for all permissible combinations of pump operation.
(7) 14.2.12.1.29 - Verify that the maximum obtainable baron dilution rate is less than or equal to that assumed in your accident analysis (Subsection 15.4.6).
(8) 14.2.12.1.34 - Ensure that the interlocks and isolation valves for overpressure protection of the RHR system are tested (Subsection 5.4. 7.2.5).
(9) 14.2.12.1.39 - State which safety signals are used to test boron recirculation pump and valve response.
(10) 14.2.1?. 1.40 - Verify that paths for the air-flow test of containment spray nozzles overlap the water-ficaw test paths of the pumps to demonstrate that there is no blockage in the flo'w path.
(11) 14.2.12.1.41 - State which safety signals are used to test containn.ent spray pump and valve respon'e.
s (12) 14.2.12.1.48 - Verify that the cooling fans can operate in accordance with design requirements at the containment design peak accident pressure.
(13) 14.2.12.1.64 - (a) Verify that the transfer pump flow capacity (Subsection 14.2.12.1.53) is sufficient to satisfy the fuel oil consumption rates.
b) Ensure that the 2 hr. and 22 hr. load tests are accomplished within a 24 hr. period.
(14) 14.2.12.1.73 - (a) Account for process-to-sensor hardware (e.g., instrument lines, hydraulic snubbers) delay times; b) Provida assurance that the response time of each primary senscr is acceptable; and l..
i.-
SNUPPS~- Paga 20 of 27 a
c) Provide assurance that' the total reactor protection system response time is consistent with your accident analysis assumptions.
Note:
Item 2 can be accomplished by measur,ing the response time of each sensor during the preoperational test, ensuring that the response. time of each sensor will be measured by the manufacturer within two years prior to fuel loading, or describing the manufacturer's certification process in sufficient detail for us to i
. conclude that the sensor response times are in accordance with design, d
(15) 14.2.12.2.6 - Verify that the operability of your liquid radwaste system will be demonstrated by actually processing representative chemical waste streams.
(16) 14.2.12.3.7 - Ensure that the moderator temperature coefficient will be derived, and that it meets the applicable acceptence criteria.
(17) 14.2.12.3.9 - Include testing at approximately 50% power.
Commit to performing step and ramp changes of full design -
value, or explain how changes of a lower value can be used to determine the proper response to design load swings.
(18)14.2.12.3.27-Commit to retesting rods, whose scram times fall outside the two-sigma limit, at least three additional times.
i SNUPPS - Page 21 of 27 640.13 We have noted on other plan, startups that the capacities of (14.2.12) pressurizer or main steam power-operated relief valves are sometimes in excess of the values assumed in the accident analyses for inadvertent opening or failure of these valves.
Provide a description Sf the initial plant test or manufac-turer's test that demonstrates that the capacity of these valves is consistent with your accident analysis assumptions.
~
640.14 Consnit to the demonstration of the operability of the (14.2.12.1) temperature sensors downstream of the,.rimary power operated r'elief valves and safety valves (Figure 5.1-1, Sheet 2).
640.15 Failura of pressurizer overpressure protection valves to (14.2.12) reseat, coupled with false position indication has cccurred recently.
One possible failure cause which' has been identified was galling of the valve bndy due to dry stroking the valves when setting release limits.
Explain what procedures will be used to protect valves during limit setting.
640.16 Verify that "unctional testing performed on valves with two (14.2.12.1) actuation trains, such as the Main Steam (Subsection 10.3.2.2) and Main Feedwater (Subsection 10.4.7.2.2) Isolation Valves, i.ncludes verification of the operability of each actuation train.
640.17 Correct the following deficiencies that were noted in your (14.2.12.1)
Cuntainment Isolation Valve test description:
1)
Subsection 14.2.12.1.10 states that Pressurizer Relief Tank Nitrogen Isolation Valves shut upon receiving a CIS, but these valves do not appear in Table 6.2.4-1.
2)
The following valves should close upon receiving a CIS (Table 6.2.4-1) but are not specifically addressed in your test procedure descriptions:
HV-7,8 - Containment Spray Recirculation FV-R9 - Instrument Air to Reactor Building FV-95,96 - Reactor Sump Pump to Floor Drain Tank HV-8843 - Boron Injection Tank to CIS Test Line
,,,y,-
r ~
~
~
.SNUPPS Page 22 of 27 3)
Containment isolation valves should be tested in an integrated manner in as much as practicable.
Note that a consnitment-satisfying 'his intent could be made in Subsection 14.2.12.1.71.4.C.
640.18 Provide test descriptions 1) that will verify that the plant's (14.2.12.1) 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 equipment within their design temperature range with the equipment operating in a manner that will produce the maximum heat load in the compartment.
If it is not practical to produce maximum heat loads in a compartment, describe the methods that will be used to verify design heat removal capability of the emergency ventilation systems.
Note that it is not apparent that post-accident design heat loads will be produced in ESF equipment rooms during the power ascension test phase; therefore, simply assuring that area temperatures remain within design limits during this period will probably not demonstrate the design heat removal capability of these systems.
It will be necessary to -
include measurement of air and cooling water temperature and flows and the extrapolations used to verify that the ventilation systems can remove the postulated post-accident heat loads.
640.19 Modify the appropriate test descriptien of the Engineered-(14.2.12.1)
Safety Features System to ensure that the following items are addressed:
(1) The starting of the ESF pumps should be verified for both emergency and normal power sources.
(2) The SI.and RHR pumps should be run under full flow conditions to verify an adequate margin to electrical trip.
5 o
-,---,,n
---,,,c+--,,
SNUPPS - Page 23 of 27 (3)
ESF pumps should be verified able to start under maximum startup loading conditions.
(4) present or reference the full flow analysis done to satisfy the intent of Regulatory Guide 1.79., C.la(2),
as coamitted to in Appendix 3A.
(5)
Ensure that the recirculation portion of the ECCS Sump Test (Subsection 14.2.12.1.83) verifies a value of NPSH greater than that required under accident temperature conditions.
640.20 Recently, questions have arisen concerning the operability (14.2.12.1) and dependability of certain ESF pumps.
Upon investigation, the staff found that some completed preoperational test procedures did not describe the test conditions in sufficient detail.
Provide assurance that the preoperational test procedures for ECCS and containment spray pumps will require recording the status of the pumped fluid (e.g., pressure, temperature, chemistry, amount of debris) and the duration of testing for each pump.
In addition, provide preoperational test descriptions to verify 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 criteria.
(The bases provided should consider both flow requirements for ESF functions and pump NPSH require-ments).
640.21 Our review of licensee event reports has disclosed that many (14.2.12.1) events have occurred because of dirt, condensed moisture, or other foreign objects inside instruments and electrical com-ponents (e.g., relays, switches, breakers).
Describe administrative controls that will be implemented to prevent component failures such as these at your facility including precautions that will be taken during initial testing program.
SNUPPS - Pag 2 24 of 27 64.. 22 For your DC Power System tests (Subsections 14.2.12.1.67, (14.2.12) 14.2.12.2.-17 and 18), verify that individual cell limits are not exceeded during the design discharge test and demonstrate t5at the DC loads will function as necessary '*
assure plant safety at a battery terminal voltage equal a the acceptance criterion that has been established for minimum battery terminal voltage for the discharge load test.
Assure that each battery charger is capable of floating the battery on the bus or recharging the completely i
discharged battery within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> while supplying the l
largest combined demands of the various steady-state loads under all plant operating conditions.
640.23 Your test descriptions are not sufficiently detailed to (14.2.12) ascertain if the voltage levels at the safety-related buses are optimized for the full load and minimum load conditions that are expected throughout the anticipated range of voltage I
variations of the offsite power source by appropriate adjustment of the voltage tap settings of the intervening transformers.
We require that the adequacy of the design in this regard be verified by actual measurement and by
)
correlation of measured values with analysis results.
Provide a description of the method for making this verification.
640.24 Make a' comitment in your test procedure descriptions to (14.2.12.1) perform the pre-and post-hot functional examination for integrity as described in Subsection 3.9(N).2.4.
640.25 There are a number of discrepancies between Tables 14.2-1 and (14.2)
Table 14.2-4.
Make the appropriate corrections to address the following problems:
(1) S-03BBil Reactor Coolant Syrtem Hydrostatic Test is included in Table 14.2-1 (Sheet 1) but missing from Table 14.2-4.
(2) S-X3NG01 480-V Class IE System Preoperational Test is included in Table 14.2-1 (Sheet 4) but missing from Table 14.2-4.
SNUPPS - Page 25 of 27 640.26 Table 14.2-5 (Sheet 3) lists S-090007 Plant Performance (14.2)
Test as one of the startup tests. This test is not included in Table 14.2-3.
Provide a footnote indicating that the test is a continuation of a nonsafety-related preoperational test.
640.27 Table 14.2-5 does not in many cases clearly indicate the II4*2I power levels specified by the test method portion of the individual startup test descriptions.
Modify Table 14.2-5 to indicate the power level or plateau at which each of the iridividual startup tests will be conducted.
i 5
4 i
i
,m,
SNUPPS - Pag] 26.of 27 Errata--SNUPPS Page Section 14.0 - i 14.2.12.2 "14.2-118" should be "14.2-120" 14.2.12.3 "14.2-150" should be "14.2-152" 14.2 - 23 14.2.12.1.8
" Turbine Driven" should be
" Turbine-Oriven" 14.2 - 30 14.2.12.1.15
" Tests" should be " Test" 14.2 - 65 P.evision number is misleading.
14.2 - 73
- 14. 2.12'.1. 51. 4. b
" response" should be " respond" 14.2 - 90 14.2.12.1.64
" Sequences" should be " Sequencer" 14.2 - 102
- 14. 2.12.1. 71. 4. c
" components," should be " components" 14.2 - 105 14.2.12'.1.72.4.a
" Low of offsite power" should be
" Loss of offsite power" 14.2 - 108 14.2.12.1.74.1
" alarms" should be ' alarms,"
14.2 - 110 14.2.12.1.76.1
" alarm," should be " alarms,"
14.2 - 111 14.2.12.1.77.2.d "of outside" should be "or outside" 14.2 - 113 14.2.12.1.79.3
" pressure and deflection measurements,"
should be " pressure, and deflection measurements" 14.2 - 120 14.2.12.2
" abstract" should be " abstracts" 14.2 - 122 14.2.12.2.2.3.c
" times" should be " times are" 14.2 - 155 All references to 14.2.12.13 should be to 14.2.12.3.
i4.2.12.3.3.3.b
" critical" should be " criticality" 14.2 - 156 14.2.12.3.4.3.a "date" should be " data" 14.2 - 159 14.2.12.3.7.3
" moderator temperature" sshould be
" moderator temperature coefficient" 14.2 - 165 14.2.12.3.13.1-
" transits" should be " transients" T,
4 4
A p
s.
j
SNUPPS - Paga 27 of 27 Pace Section 14.2 - 173 "14.2.12.3.2.1.3" should be "14.2.12.3.21.3" 14.2.12.3.21.4
" positive" should 'be " positive /
negative" 14.2 - 180 14.2.12.3.28.2.a "on-load" should be "no-load" 14.2 - 181
'14.2.12.3.29.3
" transit" should be " transient" 14.2 - 182 14.2.12.3.30.3.a All references to " percent" should be to " percent power".
14.2 - 183 "14.2.12.3.3.14" should be "14.2.'12.3.31.4" 14.2-184 14.2.12.3.32.3
" work" should be "wurth" "an/or" should be "and/or" 14.2 - 186 14.2.12.3.34
" Loading" should be " Load" Table 14.2-1 S-03BB05
" System Hot" should be Hot" (Sheet 1)
" System Table 14.2-1 S-03GG01 "HVAC" should be "HVAC System" (Sheet 2)
S-03KE05
" Manipulator Crane" shculd be
" Refueling Machine" S-03XE06
" Manipulator Crane" should be
" Refueling Machine" Table 14.2 4 S-03EC01 Title should be " Fuel Pool (Sheet 2)
Cooling and Cleanup System 4
Preoperational Test" S-03EM04
" Tests" should be " Test" Table 14.2-4 S-04PB01 "4.160" should be "4160" (Sheet 4)
Table 14.2-4 S-04PK01 "dc" should be "DC" (Sheet 5)
Table,14.2-5 S-070015 "Assension" should be i
(Sheet 1)
" Ascension" i
--