ML20054E639
| ML20054E639 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 06/04/1982 |
| From: | Vincent R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| References | |
| TASK-06-07.A3, TASK-6-7.A3, TASK-RR NUDOCS 8206140017 | |
| Download: ML20054E639 (22) | |
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e Consumers Power Company General Offices: 212 West Michigan Avenue, Jackson, Michigen 40201 e (517) 7984550 June 4, 1982 Dennis M Crutchfield, Chief Operating Reactors Branch No 5 Nuclear Reactor Regulation US Nuclear Regulatory Commission Washington, DC 20555 DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLANT - SEP TOPIC VI-7.A.3, ECCS ACTUATION By letter dated February 22, 1982 the NRC issued a draft evaluation of SEP Topic VI-7.A.3 for the Big Rock Point Plant. Consumers Power Company has reviewed this evaluation and the attached TER, and provides the attached comments for your consideration.
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Robert A Vincent Staff Licensing Engineer CC Administrator, Region III, USNRC hTC Resident Inspector-Big Rock Point 1 pages t
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1 REVIEW 0F NRC DRAFT SAFETY EVALUATION OF SEP TOPIC VI-7.A.3, ECCS ACTUATION SYSTEM Big Rock Point Plant nu0682-0001c142
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1 TONR 13-82 ENCLOSURE REVIEW OF NRC DRAFT SAFETY EVALUATION OF SEP TOPIC VI-7.A.3, "ECCS ACTUATION SYSTEM" A.
SUMMARY
OF THE NRC DRAFT SAFETY EVALUATION According to the NRC's draft safety report (see Reference 1), the NRC based its Safety Evaluation (SE) on a Technical Evaluation (TE) performed by its contractor, EC&G Idaho, Inc (see Enclosure I to Reference 1).
The TE states that the objective of'the review was "to determine if all Emergency Core. Cooling System (ECCS) components,' including pumps and valves, are included in component and system tests, if the scope and frequency of periodic testing are identified, and if the test program meets current licensing criteria. The systems included in the ECCS'are the Core Spray and Core Spray Recirculation System."
The current licensing criteria used in the TE consisted of the fol-lowing:
A.1 General design Criterion 37 (GDC 37), " Testing of Emergency Core Cooling Systems," required that:
The ECCS be designed to permit appropriate periodic pressure and functional testing to assure the operability of the system as a whole and to verify, under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation, including operation of applicable portions of the protection system, the transfer between normal and emergency power sources, and the operation of the associated cooling water system.
A.2 Branch Technical Position ICSB 25, " Guidance for the Interpretation of GDC 37 for Testing the Operability of the Emergency Core Cooling System as a Whole," states that:
All ECCS pumps should be included in the system test.
A.3 Regulatory Guide 1.22, " Periodic Testing of the Protection System' Actuation Functions," states in Section D.1.a. that:
The periodic tests should duplicate, as closely as practicable, the performance that is required of the actuation devices'in the event of an accident.
A.4 Standard Review Plan, Section 7.1, Appendix B, " Guidance for Evaluation of Conformance to IEEE STD 279," states in Section 11 that:
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2 Periodic testing should duplicate, as closely as practical, the overall performance required of the protection system. The test should confirm operability of both the automatic and manual circuitry. The capability should be provided to permit testing during power operation. When this capability can only be achieved by overlapping teste, the test scheme must be such that the tests do, in fact, overlap from one test segment to another.
A.5 Regulatory Guide 1.22 states in Section D.4 that:
Where actuated equipment is not tested during reactor operation, it should be shown that:
1.
There is no practical system design that would permit operation of the actuated equipment without adversely affecting the safety or operability of the plant.
2.
The probability that the protection system will fail to initiate the operation of the actuated equipment is, and can be maintained, acceptably low without testing the actuated equip-ment during reactor operation.
3.
The actuated equipment can be routinely tested when the reactor is shut down.
Section 4.0 of the TE summarizes the NRC's contractor determina-tions regarding the testing and testability of the Core Spray and Core Spray Recirculation Systems. The Section 4.0 determinations are as follows:
"The design of both the Core Spray and the Core Spray Recircu-lation Systems make testing of the systems impractical during reactor operation."
"The valves and pumps associated with the Core Spray System are periodically tested as is the instrumentation associated with the system; however, there is no systems integrated test required by the Technical Specifications to determine the operability of the system as a whole. Therefore, the system does not comply to General Design Criteria 37."
"There are no test!ng requirements for the valves in the flow-path of the Core Spray Recirculation System nor is there a requirement for a periodic systems integrated test to determine the operability of the system as a whole. The system does not comply to the current reactor licensing criteria."
The TE summary goes on to say that:
"It is left to the NRC Staff to determine whether enough operating experience at Big Rock Point can adequately establish that the nu0582-0126a-32-42
3 present testing of the Core Spray and Core Spray Recirculation System assures a low probability of system failure."
Based on the review of its contractor's TE, the NRC's SE conclu-sions were documented as Enclosure 2 to the letter described as Reference 1.
These conclusions are quoted below:
" Based upon our review of our contractor's evaluation, the staff concludes that Big Rock Point conforms to current licensing criteria except for the Technical Specifications. The staff recommends the following changes in the Technical Specifications:
1.
"The Core Spray Low Pressure Coolant Injection System valves shall be subjected to a system level, no flow, test annually when shutdown. This test shall demonstrate proper, automatic valve functioning while the pumps are locked out.
Upon comple-tion, the pumps shall be tested to assure that they have been returned to service."
i 2.
"A similar no flow test for the Core Spray Recirculation System shall be conducted."
" Proposed changes 1 and 2 above are consistent with past staff recommendations in this regard."
f "The staff's judgment is that the system test frequencies are sufficient. Our judgment is based on operating experience. The test frequencies can be changed if additional experience indicates that such action is necessary."
1 B.
DETAILED REVIEW OF THE SEP TOPIC TECHNICAL EVALUATION (TE) AND 5
SAFETY EVALUATION (SE)
A detailed review of both the TE and the SE has been conducted.
For the most part, both the TE and the SE are accurate. Exceptions to this, however, are noted in the paragraphs that follow. Also included in the following paragraphs are descriptions of the design and operation of the Core Spray System (CSS) and Core Spray Recirculation System (CSRS).
It is the opinion of Consumers Power j
Company (CP Co) that these descriptions provide the basis for cor-i recting inaccurate statements discovered in the TE and SE and for arriving at final conclusions concerning the testing and test-ability of the ECCS.
B.1 THE NRC'S TECHNICAL EVALUATION (TE)
B.1.1 The Core Spray System (CSS)
B.1.1.4 Corrections B.1.1.a.1 Page 2, Paragraph 3.1, Line 8, states, " Flow to the Core Spray System in the initial stages of a LOCA is provided by two station r
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1
4 fire pumps (one electric-and one diesel-driven) which draw water from Lake Michigan." Although not incorrect, it should be noted that the Fire Water System (FWS), which serves as the supply to the CSS, is normally maintained pressurized by the fire jockey pump.
Therefore, FWS flow to the core would commence as soon as the CSS injection valves automatically open and reactor pressure decreases below the FWS pressure to unseat the CSS check valves. That is, flow to the core would commence even if the fire pumps did not start.
Section B.1.1.b provides a detailed description of CSS and FWS operation.
B.1.1.a.2 Page 3, Paragraph 3.2, atte.; pts to describe Plant Technical Speci-fications reouirements for periodic testing of ECCS components.
The paragraph, however, fails to mention that, in addition to the valve and pump testing requirements, certain other " active" component verifications are required. Specifically, the calibra-tion of CSS actuation and pressure and flow instrumentation is performed at every refueling outage. Also, CSS actuation in-strumentation is checked, tested and calibrated at a frequency given in Technical Specifications Tables 11.3.1.4a and 11.4.1.4a.
B.1.1.b CSS Design and Operation In addition to the description of CSS operation given in the
" Bases" section of Big Rock Point Technical Specification 11.3.1.4, the operation of the CSS is also described in Reference 2.
To sum-marize the aforementioned descriptions, it can be said that the CSS response to a LOCA is relatively simple. The CSS response consists of the automatic, simultaneous opening of the four CSS Injection Valves (M0-7051, MO-7061, M0-7070 and M0-7071; refer to Attach-ment 1) whenever reactor water level and pressure decreases below predetermined set points during primary system depressurization following the onset of a LOCA. As described in Reference 2, core injection will commence as soon as the reactor depressurizes below the operating pressure of the FWS, thereby unseating Check Valves VPI-303 and VPI-304 to allow FWS wa'ter into the core.
As described Paragraph 3.1 of the TE, both fire pumps (one electric and one diesel-driven pump) will automatically start in the initial stages of a LOCA. Each pump receives an automatic start signal from steam drum level instrumentation when a low level condition is reached (a bistable trip occurs at a steam drum level of -17" below drum centerline to provide an auto start signal for both fire pumps, see' Attachment la). Therefore, both pumps will have started prior to the automatic opening of the four CSS injection valves since the automatic opening of these valves is, in part, a function of reactor low water level (a condition which should occur sometime after a low steam drum level).
\\
icthfirepumpsalsoreceiveanautomaticstartsignalresulting arom decaying FWS header pressure. This automatic start is chscribed in the " Bases" section of Technical Specification nu0582-0126a-32-42
5 11.3.1.4. shows that the electric fire pump will automatically start when PS-615 senses the decaying FWS pressure at 70 psig (+5, -0 psig). Attachment 2 also shows that the diesel-driven fire pump will automatically start when PS-612 senses the FWS decaying pressure at 60 psig (+5, -0 psig).
B.1.1.c CSS Testing B.1.1.c.1 CSS Testing While the Plant Is Shut Down Calibration and testing of the reactor pressure sensors used for core spray actuation are performed at every refueling per Proce-dure IPIS-1 (see Reference 3).
During IPIS-1 calibration and testing, each pressure switch (ie, PS-IG11A through H) is isolated and pressurized.
Attachments 3 and 4 (Schemes 5601 and 5602 for and Schemes B163 and B152 for Attachment 4) show the control circuits for CSS Injection Valves M0-7051, M0-7061, M0-7070 and M0-7071 and their associated pressure switch contacts.
IPIS-1 requires that, at a set point of 2 200 psig, the pressure switch contact (which is wired into its associated CSS injection valve control circuitry) closes.
Calibration and testing of reactor water level sensors used for core spray initiation are performed at every refueling per Proce-dure IRPS-2 (see Reference 4).
During the conduct of IRPS-2, each level switch (ie, LS/RE09A through H) is cold water calibrated using a leveling bottle. During the test, a cold water set point of 5 13.7 H O is verified by monitoring switch contact operation.
2 This cold water set point corresponds to the Technical Specifica-tions set point of 5 2'9" (-1") above the top of the active fuel in the core during reactor operation.
Procedure T180-15 (see Reference 5) overlaps with both of the aforementioned calibration procedures (ie, IPIS-1 and IRPS-2), as well as with Procedure T90-09 (see Section B.1.1.c.2).
Proce-dure T180-15, which is performed during refueling, requires that each CSS Injection Valve (M0-7051, M0-7061, MO-7070 and M0-7071) be verified to open automatically when each of its associated level switches (eg, LS-RE09A and LS-RE09C for M0-7051; see Attachment 3) has its pointer manually lowered below the level switch set point.
It should be noted that this is performed with the reactor depres-surized. Therefore, the valve's associated pressure switches (eg, PS-IG11A and C) will not have to be operated to effect valve opening upon level switch operation. The pressure switch contacts (ie, PS-IG11A C-11C and PS-IG11C C-11C) will already be closed.
After completion of these automatic actuation tests, each CSS injection valve is manually opened from its manual controller in the control room.
During this manual actuation, the time of each valve to open is measured and verified to be 5 25.0 seconds. This test is performed with the FWS isolated from the CSS.
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6 In addition to requiring response tests for the CSS injection valves as described above, T180-15 also requires that the Fire Water Alternate Supply Valve M0-7072 be manually stroked using the remote manual controller and timed for both opening and closing.
There are no automatic functions for MO-7072. Also, T180-15 requires that CSS Cross-Connect Valve M0-7069 be verified as open.
It should be noted that the circuitry for this valve has been disabled.
In addition to the aforementioned procedures, three other proce-dures related to the CSS are performed during refueling. Proce-dure TR-50 (see Reference 6), requires that both primary CSS Instru-mentation FT-2162, FI-2335, FS-2522 and FR-2108A (see Attachment 1) be calibrated by isolating the transmitter (PT-2162) and applying test pressures to the transmitter input. As the input test pres-sure is increased, an alarm set point of 5 500 gpm is verified to occur from FS-2522. Redundant CSS Instrumentation FT-2163, FI-2336, FS-2523 and FR-2108B are calibrated similarly. During this calibration, an alarm set point of 5 500 gpm is verified to occur from FS-2523. TR-50 also requires that Fire System Strainer - Alternate Path Differential Pressure Switch PDIS-7814 (see Attachment 1) be calibrated and its alarms, "High dP" (at 5 2.0 psid) and " Strainer Plugged" (5 5.0 psid), be verified.
Refueling Procedure IRDS-2 (see P.eference 6a) requires that the steam drum level transmitters that provide the automatic start for the electric-and diesel-driven fire pumps be calibrated. During the conduct of IRDS-2, the low steam drum level setpoint (at which both fire pumps start) is verified to be 2 -17" below the steam drum centerline for each of the four level transmitters (ie, LT-3184 through LT-3187 for Channels A through D, respectively). This setpoint is verified at the output of the fire pump start bistable-for each channel (see Attachment la). This procedure overlaps with Procedure T30-26 as described in Section B.1.1.c.2 below.
TR-70 (see Reference 7) is also performed during refueling. This procedure requires that both fire pumps be verified as auto-matica11y starting as a function of decaying FWS header pressure, and that each pump has capacity for a flow of at least 1,000 gpm at a head of 110 psi. During the test, the FWS jockey pump is turned off and an auto start of each pump is verified by opening FWS drain valve FP-118.
The FWS pressure at which each pump starts is recorded and verified to be as indicated in Attachment 2 (refer to Section B.1.1.b).
B.1.1.c.2 Testing Whil,e the Plant Is Operating Several tests relating to the CSS are performed while the Plant is operating.
Procedure T30-22 (see Reference 8), which is performed monthly, requires that the operability be verified for CSS Injec-tion Vc1ves M0-7051, MO-7061, M0-7070 and M0-7071 as well as CSS shell side inlet Valve MO-7066. Each of these valves is verified nu0582-0126a-32-42
7 to oper
_3 close as initiated from the remote manual controller.
Durin each valve's test, the opening and closing strohe is timed.
a In addition, substation deluge Isolation Valve CV-4101 is verified to automatically close any time one of the CSS injection valves is open. CV-4101 closes under such conditions to conserve FWS flow during the LOCA. CV-4101 is then verified to reopen when the injection valves reclose.
T30-22 also requires that the check valve in each core spray line (ie, VPI-303 for the secondary spray line and VPI-304 for the primary spray line) be leak checked. During the verification, the ability of the check valve to isolate the reactor system high pres-sure from the CSS operating pressure is conducted with the electric fire pump running. The ability of each check valve to isolate is evaluated by alternately opening each motor-operated valve on either side of the check valve and then observing the FWS pressure for an increasing indication indicative of check valve leakage.
This check valve testing, along with the check valves' maintenance history, was discussed in a previous CP Co submittal as identified in Reference 9.
Procedure T30-24 (see Reference 10), also performed monthly, requires that a response test of the CSS flow indicators and re-corders be conducted. During the test, a test push button is depressed which results in a test input signal being applied to Instruments FI-2335, FI-2336 and FR-2108 (see Attachment 1).
Upon depressing the push button, each instrument is verified to be indicating an expected value.
Procedure T30-26 (see Reference 10a) is performed monthly to verify the automatic starting of both the electric and diesel fire pump from the circuits that are supplied by the steam drum level transmitters. T30-26 requires that the fire pump start bistables be tripped using the bistable test pushbutton (see Attachment.lb).
During the test, the electric fire pump is verified to automatically start as a result of one of the possible two-out-of-four combinations of fire pump bistable trips (note, Attachment la shows Channel A as typical of Channels A, B, C and D).
All of the remaining two-out-of-four combinations are then verified to generate an automatic electric fire pump start signal as indicated by fire pump status lights. This test is run with the automatic start capability of the diesel fire pump inhibited. The test is then repeated verifying the automatic start of the diesel fire pump with the automatic start capability of the electric fire pump inhibited. This test overlaps with Procedure IRDS-2 as described in Section B.1.1.c.1.
Finally, Procedure T90-04 (see Reference 11) is performed every 90 days to verify the operability of the Emergency Core Cooling System instrumentation via instrument trip testing. During the test, each CSS initiating pressure switch (ie, PS-IG11A through PS-IG11H; see, Schemes 5601 and 5602, and Attachment 4r Schemes B163 nu0582-0126a-32-42
8 and B152) and level switch (ie, LS-RE09A through LS-RE09M; see aforementioned attachments and schemes) is operated and its contact, which is wired into its associated CSS injection valve control circuitry, is. verified to operate. For each pressure switch, the switch's input pressure-is decreased to result in contact closure which is verified. The pressure:is then increased to its normal operating level and the contact is verified to reopen. For each level switch, the switch's level pointer is moved downward and contact closure is verified. The pointer is then returned to its operating pcsition and the contact is verified to reopen. Throughout the testing, as described above, adequate pre-cautions are taken to ensure that neither core spray line is opened which would allow FWS flow into the core. This test overlaps with Procedure T180-15 and Procedures IPIS-1 and IRPS-2 as described in Section B.1.1.c.1.
B.1.1.c.3 CSS Conclusions Since the CSS design requires that the FWS be isolated from the CSS prior to performing system level surveillance testing so as not to periodically flush the core with Lake Michigan water and the FWS serves as the supply to the CSS during an accident, CSS injection valve system level testing is performed only during shutdown under no-flow conditions.
Page 4 of the TE (see keference 1) acknowl-edges this, in part, by s6ating, "The design of both the Core Spray and the Core Spray Recirculation Systems make testing of the systems impractical during reactor operation."
Although the system level testing is performed only during shutdown under no-flow conditions, it is the opinion of CP Co that the nature and frequency of the existing CSS testing are adequate to ensure that the CSS will operate as intended during an accident condition for the following reasons:
1.
The CSS's response to the accident is relatively simple by design (ie, the CSS injection valves automatically open when reactor pressure and water level sensors signal the parameter accident values to align a prepressurized FWS to the reactor core).
2.
System level testing (ie, the automatic actuation from sensor input to verification of proper actuated component response; see description of IPIS-1, IRPS-2 and T180-15 testing in Section B.1.1.c.1), which includes proper test overlap and timing verifications, is performed at every refueling.
3.
Other testing, as described in Sections B.1.1.c.1 and B.1.1.c.2 (which includes verification of automatic start and capacity of the FWS pumps; see description of IRDS-2, TR-70 and T30-26), is also performed with proper overlap.
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9 4.
TVo redundant CSS lines exist (the primary line from a 6" con-nection to the FWS and the secondary line from a 4" connection to the FWS) to provide core spray along with an alternate 4" line which routes from the inlet of the core spray heat exchanger via MO-7072.
B.1.2 The Core Spray Recirculation System (CSRS)
B.1.2.a Corrections
_There are no corrections to be noted for Section 4.0, " Core Spray Recirculation System," of the TE (see Reference 1):
B.1.2.b CSRS Design and Operation The purpose of the CSRS and its design is described in the "Fases" section of Technical Specifications 11.3.1.4 and 11.4.1.4.
As described, the CSRS is provided to prevent excessive water buildup in the containment sphere and to provide for long-term, post-accident cooling. The system consists of two pumps (400 gpm each) and a heat exchanger. The pumps take suction from the lower levels of containment and discharge to the core spray headers. The system is actuated manually when the water level in the containment rises to Elevation 587'.
The 587' elevation will be achieved between 6 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> operation of one Core Spray and one Containment Spray System. The Technical Specifications also state that a test tank and appropriate valving is provided in the CSRS so the pump suction conditions and the flow characteristics of the system can be periodically tested.
In addition, the specifications state that one core spray pump has adequate capacity to provide fuel cooling at any time following a LOCA. Continuous containment spray operation is not required during the post-accident recirculation phase if only one recirculation pump is available.
A detailed description of the use of the CSRS can be obtained by referring to References 14 and 15, Reference 14 details the proce-dure in which the CSRS is used. SOP 8 states that the CSRS is used to cool and recirculate the post-incident water after it accumulates to a level exceeding Elevation 587' in the sphere.
SOP 8 provides the following instructions to manually place the CSRS in service (refer to Attachment 1):
Step 1:
" Check the containment sphere water level, after operation of the Post-Incident System, above Elevation 587' but less than 590'."*
Step 2:
"Open MO-7066 using RMC-5521 in control room.
(If there is a malfunction of this valve, dose rates will be lower at this time for an operater to open the bypass valve.) A dedicated hose for backup cooling water is stored in the screenhouse if M0-7066 cannot be used."
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10 Step 3:
" Check remote manual control for M0-7072 is in the ' Pull-to-Stop' position.
(Head pressure on the core spray pumps is higher than the Fire System pressure and a backfeed to the Fire System'would be possible.)"
Step 4:
" Start one core spray pump. The ' core spray pumps low-discharge pressure' alarm is initiated when a pump is started and stays in alarm condition until the discharge pressure exceeds 50 psig."
Step 5:
"Close the Core Spray-and Enclosure Spray Fire Water Header Supply Valves FP-29 and FP-30 in the machine shop."
Finally, it should be noted that the CSRS recirculation flow path consists of only nonelectrically operated valves which are locked open or nonelectrically operated check valves. Therefore, during power operation, the CSRS is normally prealigned for service such that recirculation flow will commence whenever one of the core spray pumps is started.
B.1.2.c CSRS Testing B.1.2.c.1 CSRS Testing While the Plant Is Shut Down Regarding the instrumentation required by the operator to determine when to commence the manual switchover, Procedure IPIS-6 (see Ref-erence 16) requires that calibration be performed on Instru-ments LT-3171, LT-3175, LR-3110 and LR-3111. Control room light response to their associated level switches (ie, LS-3562, LS-3563, LS-3564 or LS-3565) is also checked during each refueling outage.
Along with certain at power test procedures as identified in Sec-tion B.1.2.c.2, two refueling outage test procedures are performed to verify the capability of the CSRS to serve in the recirculation t
mode.
Procedure TR-05 (see Reference 17) requires that the core l
spray pumps be lined up for recirculation flow testing through the core spray heat exchanger and core spray test tank (See Attach-ment 1).
During the test, one of the pumps is started and certain pump operating parameters are monitored such as discharge pressure, l
suction pressure, pump speed, pump vibration and pump lubrication.
l Pump differential pressure is then calculated and verified as 1
- Sphere water level can be monitored by using either Control Room Recorder LR-3110 I
which is supplied by In-Containment Transmitter LT-3171, Control Room Re-l corder LR 3111 which is supplied by In-Containment Transmitter LT-3175, or by In-Contai. ment Level Switches LS-3562, LS-3563, LS-3564 or LS-3565, each of which supplies i level signal to its own control room light. Attachment I and Refer-l ence 15 p. ovide a detailed account of LT-3171 and the level switches.
LR-3110, l
LR-3111 aid LT-3175 were installed during the recent refueling outage as part of the CP Cc post-TMI response and, as a result, are not yet available in the l
drawing ' Attachment 1).
LR-3110, LR-3111 and the control room level lights are specifically called out in Procedure S0P 8 (see Reference 14).
l t
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11 acceptable. During the pump run, CSRS piping, pump and valves located in the core spray heat exchanger room are checked for leakage. The procedure then requires that the other core spray pump be tested in a similar fashion. During every third refueling outage, the procedure requires that the pump and test loop be ISI tested per ASME Pressure and Vessel Code,Section XI.
Regarding the shell side of the core spray heat exchanger, Proce-dure TR-09 (see Reference 18) requires that the FWS be lined up through the heat exchanger shell to verify proper system flow requirements. During the flow test (with the electric fire pump passing flow through Valve M0-7066, see Attachment 1), the dis-charge pressure of the electric fire pump, as well as the pressure at the discharge of the shell, is recorded. Upon closing Valve M0-7066 and using a 2-1/2" fire hose installed between a Fire System manifold and the core spray heat exchanger tie-in at Valve VPI-10, the test is repeated.
B.I.2.c.2 CSRS Testing While Plant Is Operating Procedure T30-14 (see Reference 19) is performed monthly to verify the integrity of the core spray heat exchanger tube bundles by filling the shell and measuring tube leakage.
Regarding the operability of the core spray heat exchanger Shell Inlet Valve MO-7066 (see Attachment 1), Procedure T30-22 (see Ref-erence 8) requires that the valve be stroked open and closed from the remote manual controller located in the control room.
Both the opening time and closing time are recorded.
B.1.2.c.3 CSRS Conclusions Given the fact that full _ operation of the CSRS requires that the CSS be in operation (ie, the CSS injection valves must be open) and a water supply available in the bottom of the containment sphere, it is not practical to full flow test the CSRS either during Plant operation or shutdown. However, since this sytem is operated manually only and periodic testing as described in Sections B.1.2.c.1 and B.1.2.c.2 is performed, it is the opinion of CP Co that this system will perform as intended during the accident con-dition to terminate the increase in sphere water level and provide long term core recirculation.
B.2 THE NRC'S SAFETY EVALUATION (SE)
B.2.1 The Core Spray System (CSS)
As concluded in the SE (see Reference 1), the NRC states that Big Rock Point conforms to current licensing criteria except for the Technical Specifications. As a result, the staff recommends, "The Core Spray Low-Pressure Coolant Injection System valves shall be subjected to a system level, no flow, test annually when shutdown.
nu0582-0126a-32-42
12 This test shall demonstrate proper, automatic valve functioning while the pumps are locked out.
Upon completion, the pumps shall be tested to assure that they have been returned to service."
Regarding the above recommendation, it should be noted that Big Rock Point performs system level, no-flow testing during every refueling outage. As described in Sections B.1.1.c.1 and B.1.1.c.3, such testing consists of verifying CSS injection valve opening in response to sensor actuation signals. As described, Test Procedures IPIS-1, IRPS-2 and T180-15 are used to verify proper operation from the input of the actuation sensors (the CSS injection valves' pressure and level sensors) to the output of the actuated device (full opening of the injection valves) utilizing proper test overlap. CP Co considers such testing as system level testing since the above description is consistent with IEEE 279-1971, Paragraph 1 (see Reference 12), which states, "the nuclear power generating station protection system encompasses all electric and mechanical devices and circuitry (from sensors to actuation device input terminals) involved in generating those signals associated with the protective function."
The system level testing at Big Rock Point is already called out in Technical Specification 11.4.1.4B which states, "At each major refueling outage, the following shall be performed: calibration of core spray system actuation and pressure and flow instrumentation."
Technical Specification 11.4.1.4B also states, " Verify manual and automatic actuation of the Core Spray System Valves MO-7051, -7061,
-7070 and -7071 with water flow normally blocked." It should also be noted that it is not necessary to lock out the FWS pumps as recommended in the SE since the FWS isolation valves to the CSS are closed and tagged prior to T180-15 testing.
In addition, pressure and level switch calibration per IPIS-1 and IPRS-2 is performed only after Remote Manual Controllers RMC-5501 (MO-7061), RMC-5519 (M0-7051), RMC-5527 (M0-7070) and RMC-5528 (MO-7071) are placed in the " Pull-to-Stop" position to prevent unnecessary valve opening.
This also ensures that the FWS is isolated from the reactor core.
Neither FWS pump is expected to start during the aforementioned testing.
Regarding the FWS, the automatic start of both the electric and diesel fire pumps is verified monthly per Procedure T30-26. As described in Section B.1.1.c.2, T30-26 verifies the automatic start of the pumps as a function of low steam drum level. Technical Specification 11.4.1.4 states that each month the automatic actuation of both fire pumps is to be performed.
In addition, Procedure TR-70 is performed at each refueling to verify the automatic start of both fire pumps as a function of decaying FWS header pressure. TR-70 also verifies the capacity of the pumps during flow conditions. Technical Specification 12.4.7.11.1.1.d.2 requires that once per 18 months a system functional test be performed which includes a simulated automatic actuation of the system throughout its operating sequence.
nu0582-0126a-32-42
13 Finally, it should be noted that the NRC has determined that the CSS test frequencies are sufficient. This determination was based on operating experience.
B.2.2 The Core Spray Recirculation System (CSRS)
As described on Page 2 of Enclosure 2 of the SE (see Reference 1),
the NRC recommends that CP Co change its Technical Specifications to require that the Plant perform a system level, no-flow test similar to that recommended for the CSS. According to the recom-mendation, the test shall demonstrate proper automatic valve func-tioning. Regarding this recommendation, it should be noted that no automatic valve functions exist in the CSRS. As described in Sec-tion D.1.2.b, all of the valves in the recirculation flow path are nonelectrically operated. Although an electrically operated Valve MO-7066 is used at the inlet to core spray heat exchanger shell, there are no automatic initiation circuits for this valve (see Attachment 3, Scheme 5418).
Nevertheless, this valve is stroked and timed monthly as described in Section B.1.2.c.2.
Big Rock Point Technical Specification 11.4.1.4A calls for this monthly verification. Regarding the testing of the CSRS, it is the opinion of CP Co that the test program. conforms to the intent of Reg Guide 1.22 (see Reference 20) which states, "The periodic tests should duplicate, as closely as practicable, the performance that is required of the actuation devices in the event of an accident."
As in the case for the CSS, the NRC has determined (based on operating experience) that the CSRS test frequencies are sufficient.
nu0582-0126a-32-42
1 TONR 13-82 ENCLOSURE LIST OF REFERENCES (Page 1 of 2)
Reference No Reference Description 1.
Letter: D M Crutchfield to D J VandeWalle, 2/22/82, "SEP Topic VI-7.A.3, ECCS Actuation System, Draft Safety Evalua-tion Report For "ig Rock Point" 2.
Letter: R A Vincent to D M Crutchfield, 11/12/81, " Big Rock Point Plant - SEP Topic V-11.A, Isolation of High and Low Pressure Systems," Section B.2 3.
Maintenance Procedure IPIS-1, Rev 10 " Calibration and Testing of Reactor Pressure Sensors Used For Reactor Core Spray Actuation" 4.
Maintenance Procedure IRPS-2, Rev 6, " Calibration and Testing of Reactor Water Level Sensors" 5.
Procedure T180-15, Rev 19, " Core Spray and Enclosure Spray Valve Initiation and Operability Test" 6.
Maintenance Procedure TR-50, Rev 6, " Calibration and Testing of the Core Spray Flow, Enclosure Spray Flow and Fire System Strainer Differential Pressure Instrumentation" 6a.
Procedure IRDS-2, Rev 8, " Steam Drum Water Level Calibration Checks."
i 7.
Procedure TR-70, Rev 5, " Fire Suppression Water System Func-tional Test and Pump Capacity Test" 8.
Procedure T30-22, Rev 17, " Emergency Core Cooling System Valve Tests" 9.
Letter: R A Vincent to D M Crutchfield, 11/12/81, " Big Rock Point Plant - SEP Topic V-11.A, Isolation of High and Low Pressure Systems" 10.
Procedure T30-24, Rev 7, " Emergency Core Cooling System Flow Recorder Test" 4
10a.
Procedure T30-26, Rev 7, " Electric and Diesel Fire Pumo L2 Module Test."
11.
Procedure T90-09, Rev 4, " Core Spray Instrument Trip Test" nu0582-0126b-32-42
i Pags 2 of 2 Reference No Reference Description 12.
IEEE 279-1971, "IEEE Standard Criteria For Protection Systems For Nuclear Power Generating Stations," Paragraph 1, " Scope" 13.
Letter: R A Vincent to D M Crutchfield, 3/29/82, " Big Rock Point Plant - SEP Topic VI-10.A, Electrical, Instrumentation and Control Portions of the Testing of RTS and ESF,"
Section 4.1.3.2, Third Paragraph 14.
System Operating Procedure S0P 8, Rev 30, " Post-Incident System" 15.
Letter: R A Vincent to D M Crutchfield, 10/21/81, " Big Rock Point Plant - SEP Topic VI-7.B, Eoe Switchover From Injection To Recirculation Mode" 16.
Maintenance Procedure IPIS-6, Rev 6, " Calibration of the Containment Sump Level Transmitters (LT-3171 and LT-3175) and Recorders (LR-3110 and LR-3111)"
17.
Procedure TR-05, Rev 14, " Core Spray Pump Run and Test Loop Operation" 18.
Procedure TR-09, Rev 9, " Core Spray Heat Exchanger Shell Side Flow" 19.
Procedure T30-14, Rev 10 " Monthly Core Spray Heat Exchanger Leak Test" 20.
Regulatory Guide 1.22, " Periodic Testing of Protection System Actuation Functions," Paragraph D.1.a. 2/17/72 4
i nu0582-0126b-32-42
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