ML20096H108
| ML20096H108 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 09/07/1984 |
| From: | Stead F CLEVELAND ELECTRIC ILLUMINATING CO. |
| To: | |
| Shared Package | |
| ML20096H048 | List: |
| References | |
| OL, NUDOCS 8409110261 | |
| Download: ML20096H108 (49) | |
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9 .~)$ndISFONDENCE September 7, 1984 UNITED. STATES OF AMERICA
. NUCLEAR REGULATORY COMMISSION
-BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )
)
THE CLEVELAND ELECTRIC ) Docket Nos. 50-440 ILLUMINATING COMPANY ) 50-441
)
(Perry Nuclear Power Plant, )
Units 1 and 2) )
AFFIDAVIT OF FRANK R. STEAD ON THE DESIGN OF THE INITIATION FUNCTION OF THE STANDBY-LIQUID CONTROL SYSTEM STATE OF OHIO )
- ss COUNTY OF LAKE )
Frank R. Stead, being duly sworn, deposes and says as
- follows:
- l. I, Frank R. Stead, am Manager of Nuclear Engineering of.The Cleveland Electric Illuminating Company. My business address is 10 Center Road, Perry, Ohio 44081. In my position, I have responsibility for the system design of all nuclear sys-tems of the Perry. Nuclear Power Plant, including the Standby Liquid Control System. A summary of my professional qualifica-tions and experience is attached hereto as Exhibit "A." I have personal knowledge of the matters set forth herein and believe them to be true and correct.
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- 2. The Standby. Liquid Control System ("SLCS") has been
- included in the Perry design since the construction permit stage. The Perry Preliminary Safety Analysis Report ("PSAR")
discussed the SLCS and stated that it was manually initiated.
.PSAR, 5 4.2.3.4-(Exhibit "B" hereto).
- 3. The Final Safety Analysis Report ("FSAR") from its first'aubmission to NRC included the SLCS in the Perry design.
As tendered to the Staff in June 1980 and docketed in January 1981 the FSAR described the SLCS as having manual initiation.
See, for. example, FSAR 55 7.4.1.21 / and 9.3.5.2,2/ Figure 7.4-2 (Exhibit "C" hereto). Additional information on SLCS initiation was. included in subsequent revisions of the FSAR; in all-cases the information continued to show a manually initi-ated SLCS system. For example, in Amendment 11, dated February 15, 1983,3/ a-detailed discussion ~of modifications to prevent and mitigate the consequences of anticipated transients without scram ("ATWS")'was provided,-including further information on SLCS initiation. See, for example, FSAR 5 15C.5.II.4/
]/ "The SLCS is initiated by the control room operator by turning a keylocked switch for system A, or a different keylocked switch for system B to the 'RUN' position."
FSAR, p. 7.4-6.
2/ "The standby liquid control system (see Figure 9.3-19) is manually initiated in the main control room ...." FSAR,
- p. 9.3-19.-
3,/ The draft version of this amendment was transmitted to the NRC on January 26, 1983.
4/ "The standby liquid control system (SLCS) action is to be initiated manually in a failure to scram condition in ac-
.cordance with Emergency Instructions." FSAR, p. 15C-5.
l (Exhibit "D" hereto.) Already existing references (such as those cited above in FSAR 55 7.4.2 and 9.3.5) remained, and continued to describe the SLCS design as having manual initiation.
- 4. The FSAR in its current status still shows the Perry SLCS design as including only manual initiation See, e.g.,
FSAR 55 7. 4.1. 2,5/ 7. 4. 2. 2,5/ 7. 4. 2. 3,2/ 9. 3. 5. 2,$/ 15C. 5. II,9/
and Tables 15C-3 to -7.lE/ This is consistent with the entire history of the FSAR which always reflected manual SLCS initiation.
- 5. The Electrical Elementary Diagrams prepared by Gener-al Electric ("GE"), the vendor for Perry's nuclear steam supply system, and Gilbert Associates, Inc. ("GAI"), the plant's architect-engineer, for the SLCS originally reflected a manu-ally initiated SLCS. GE Drawing No. 828E234CA Rev. O and GAI Drawing No. B-208-030, Rev. - . (GAI produces Perry-specific drawings for systems within GE's scope of design (i.e., the 5/ "The SLCS is a backup independent method of manually shutting down the reactor ...." FSAR, p. 7.4-5.
6/ "SLCS is initiated by the control room operator." FSAR,
- p. 7.4-19.
7/ "The SLCS is initiated manually ...." FSAR, p. 7.4-26.
8/ The SLCS "is manually initiated." FSAR, p. 9.3-19.
9/ "The standby liquid control system (SLCS) action is to be initiated manually ...." FSAR, p. 15C-5.
10/ Sequences of events showing that " Operator initiates SLCS." FSAR, pp. 15C-13-19.
I ..
nuclear steam supply systems) based on the GE-furnished generic
-or plant specific documentation.)
- 6. CEI and GE were both aware that NRC was considering an ATWS rule.and that the rule, when issued, might require au-
. tomatic SLCS initiation. Automatic initiation was one of the ATWS design modifications considered by the NRC Staff in its ATWS' report. issued in 1978, NUREG-0460, " Anticipated Transients Without Scram For Light Water Reactors," Vol. 1-3 (1978). CEI
. believed then (and still believes) that the operators have the appropriate indications.and training to promptly initiate SLCS if needed. Further, automatic SLCS initiation carries with it a high probability that an inadvertent initiation would occur at some. point during plant operation, causing a costly and un-necessary outage. See, for example, CEI's letter to GE, dated February 22, 1980 (Exhibit "E" hereto).
- 7. As mentioned earlier, the SLCS first appeared in the Perry design when the PSAR was issued. However, GE was carry-ing out generic and plant specific ATWS analyses and design work both before and after the FSAR was submitted. The great.
On bulk of this work was unrelated to SLCS initiation.11/
December'20, 1979, GE presented an unsolicited proposal to CEI to prepare " reports analyzing the BWR during an ATWS event in accordance with the requirements of NUREG-0460, Volumes I-III 11/ The great majority of.the work covered such features as Recirculation Pump Trip, Alternate Rod Insertion, feedwater runback and increased SLCS flow capacity.
~
and for work to support CEI concerning the NRC 'Early Verifica-I tion' Program Reports (May and December 1979)". The proposal, referred to as Quotation 149, was accepted by CEI on January 24, 1980. The analyses initiated by GE in carrying out Quotation 149 were based upon the package of ATWS modifications subsequently referred to as Alternate 3A, which included (con-sistent with NUREG-0460) automatic SLCS initiation.12/
- 8. Following the publication of the NRC Staff's ATWS recommendations in March 1980 (Vol. 4 of NUREG-0460), GE on December 22, 1980 submitted to CEI a proposal, referred to as Quotation 149-A, for " design changes related to the [ATWS] mat-ter currently being considered by the NRC." The proposal was based on the NRC Staff's Alternate 3A, set forth in NUREG-0460, Vol. 4, based on GE's belief that " Alternate 3A ... appears to be the modifications which the NRC will eventually apply to the BWR." One of the ATWS-related modifications described in Al-ternate 3A was automatic SLCS initiation. Thus, the scope of work for Quotation 149-A included an SLCS which "will be initi-ated automatically." Although the quotation referred to both
" design services and associated equipment," the equipment was undefined (and unpriced) since the design work had not been un-dertaken.11/
12/ NEDE-25518, " Design Analysis and SAR Inputs for ATWS Per-formance and Standby Liquid Control System" (December 1981).
13/ The scope of work did include a list of ATWS hardware.
However, the list was a " preliminary estimate" which was (Continued next page)
_J
- 9. CEI was concerned that an ATWS rule requiring the Al-ternate 3A modifications might be adopted by the NRC such that the changes required by the rule might impact Perry's fuel load .
schedule (then estimated at May 1983). With respect to the SLCS initiation portion of Alternate 3A, CEI wanted to retain manual initiation if the status of the ATWS rule and the fuel loading schedule permitted. To anticipate a possible ATWS rule, CEI proceeded with the entire Alternate 3A package, including automatic SLCS initiation. That way, automatic initiation could be installed if necessary. Because CEI had concerns with the schedules, scope and other aspects of Quota-tion 149-A, particularly its compatability with a May 1983 fuel load date, CEI rejected it by letter dated January 13, 1981.
CEI then stated in a letter dated February 9, 1981 that it would accept the Quotation if these matters were resolved. GE resubmitted its proposal on April 13, 1981 (Quotation 149-B).
(This proposal superceded Quotation 149-A.) The revised Quota-tion again included the entire Alternate 3A ATWS package. Quo-tation 149-B called for GE to generate a " standard ATWS design j package", to apply that generic design.to the specific project, l
and to provide equipment. As in Quotation 149-A, only a gener-I al estimate of overall equipment needs was supplied. (Exhibit (Continued) very general (i.e. "20 switches", "8 meters", etc.) and consolidated the equipment needs for all ATWS changes in
! Alternate 3A including SLCS initiation.
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"F"' hereto,[ commercial'information deleted].) CEI accepted Quotation'149-B-on. June 3, 1981. On November 9, 1981,.GE sub-mitted'tolCEI-Quotation 149-D (Quotation 149-C did not relate
- to SLCS). ' Quotation 149-D quoted a price for all GE-scope ATWS equipment-to implement' Alternate 3A, including'the few items related to automatic SLCS initiation.14/
~
CEI accepted Quota-tion 149-D on January- 26, 1982.
- 10. During early 1982, GE continued its design and ana-lytical work on the entire Alternate 3A package, including au-tomatic SLCS initiation. CEI continued to monitor the ATWS regulatory situation. Based upon the overall status of plant construction, CEI decided to'present the Alternate 3A package with manual-initiation to the NRC Staff.
- 11. - In June 1982, GE completed its' design work under Quo-
- tation 149'B for automatic SLCS initiation and furnished the electrical elementary drawings to GAI. GE Drawing No.
828E234CA Rev. 3-(dated June 18,.1982). However, consistent
, ~with CEI's determination to retain manual initiation, at the.
i; June 29,- 1982 meeting of the Advisory Committee on Reactor Safeguards subcommittee, CEI discussed manual initiation of i -SLCS. Tr. 281-2.' And, at a July 20, 1982 meeting with the NRC l
l Staff, CEI described the " systems upgrade for ATWS" as ,
[- including ~"a manually operated standby liquid control system."
14/ The equipment listed which applied to automatic SLCS initiation were the 2 "Three Position Eletroswitch[s][ sic]" and the 6 " Relay [s] (Agastat or equivalent)".
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, See NRC memorandum.from Stefano to Schwencer, dated July 22, '
1992 (Exhibit ~"G". hereto). Similarly, in an August 6, 1982 let-ter from CEI to GE (Exhibit "H" hereto) commenting on.
NEDE-25518, CEI directed that GE correct the report so that it would reflect manual SLCS initiation. Finally, in CEI's August 13,'1982 letter'to the NRC (Exhibit "I" hereto), CEI's Vice President, System Engineering and Construction stated that while "the design includes both manual and automatic initiation 4
capability, only manual initiation will be functional." Mr.
- Davidson's Affidavit addressee this letter in more detail.
- 12. Notwithstanding the manually initiated design de-scribed in the FSAR and CEI's expressed intent to retain manual initiation (while being prepared to convert to automatic if re-quired by the final NRC ATWS rule), GAI's Electrical Elementary Drawings for the SLCS syste'm were modified to show automatic SLCS initiation based on GE's June 1982 SLCS Electrical Elemen-tary Diagrams. GAI Drawing No. B-208-030 Rev. F, dated August 2, 1982. GAI made similar changes in drawings for re-lated systems, p
- 13. Having heard the NRC Staff's reaction to CEI's ATWS l proposals (including manual initiation) at the July 20, 1982 l
l- meeting, CEI on August 9, 1982, wrote to GE to request that GE's design return SLCS to manual initiation (Exhibit "J" I hereto). GE forwarded preliminary modification diagrams (" mod-
. ification kits") to CEI on November 8, 1982 (Exhibit "K" L hereto). GE also transmitted at the same time a preliminary t
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-draft of its revision to NEDE-25518 (renumbered NEDE-22276),
which (among other things) reflected manual initiation of the SLCS. CEI made frequent requests to GE to expedite the issu-ance of final versions of the modification kits (see Exhibit "L"
hereto). The appropriate GE drawings were effectively changed by Engineering Change Notice NJ 50426, dated Decenber 28, 1983 (Exhibit "M" hereto), and the formal di-wings wers issued on January 13, 1984 (Drawing No. 828E234CA, Rev. 8),
GAI made the corresponding changes in its drawings on February 16, 1984 (Drawing No. B-208-030, Rev. K). Similar changes to drawings of related systems have also been made.
- 14. In summary, the FSAR has always shown a manually ini-tiated SLCS as the Perry design. GE was asked to perform de-sign and analysis work including automatic initiation as part of the total Alternate 3A package as a precaution against the construction impact in the event that a final ATWS rule would require automatic initiation prior to fuel loading. At about the'same time that GE was completing its drawings for SLCS au-tomation, CEI was informing the ACRS and the Staff that its final ATWS package would include manual initiation. Shortly thereafter, on August 2, 1982 the GE drawings were incorporated into GAI's Perry specific drawings. On August 9, 1982, based on CEI's meeting with the NRC Staff, CEI requested GE to return GE's SLCS drawings to a manual configuration. The GE drawings were effectively changed in December 1983. In February 1984, the GAI SLCS drawings were revised to again reflect manual
- initiation. Although GE and CAI drawings for a time showed an automatically initiated SLCS, CEI has always intended that the SLCS be designed for manual initiation if allowed by the final ATWS rule. In addition, CEI intended to be prepared to imple-ment the final ATWS rule based on Alternate 3A (if that were adopted) with a minimum of impact on the construction and fuel loading schedule. The design and analytical work undertaken by GE for automatic initiation was to provide a contingency in case an ATWS rule might compromise CCI's ability to make its fuel load schedules. In conclusion, the Perry SLCS design provides for manual SLCS initiation and complies with the June 26, 1984 ATWS rule.
2 4 .
Frank R. Stead Subscribed and sworn to before me this 7 day of September, 1984.
AN ., L , ,->tb Notary Public My Commission Cxpires PATRICIA G. CECEK, Notary Public STATE OF CHIC (Lake Countyi
_ ___ :- Empires Apes is,1988 Itx C-_
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RESUME OF FRANK R.~ STEAD
' EDUCATION AND TRAINING: - Bachelor of Mechanical Engineering, Cleveland State University,.1970
- Master of Science Degree'in Mechanical Engineering, Ohio State University, 1972
- Three-week In-Core Fuel Management Course, Purdue University, 1972
- Westinghouse Large Steam Turbine Operator Awareness Program, Dutton Mill, 1980
-EXPERIENCE:
~ 1965-Present - The Cleveland Electric Illuminating Company Joined CEI in 1965 and held _various engineering positions at CEI's Perry Project including Senior Design Engineer of Nuclear _ Fuels, gaining seven years experience in the Fuel Management area.
Also held positions in Nuclear Licensing, Balance-of-Plant Equipment, and Civil Engineering.
Most recently, rotated through several General Supervisor positions in Maintenance and Operations at CEI fossil plants.
.In July 1982, named to present position of Manager, Nuclear Engineering Department,.with responsibilities for operational support and modification engineering, reliability and design assurance, and nuclear licensing and fuel management at the Perry Nuclear Power Plant.
Reports to the Vice President, Nuclear Operations Division.
' PROFESSIONAL
~ MEMBERSHIPS: - American Society of-Mechanical Engineers
- American Nuclear Society
- Registered Professional Engineer, State of Ohio Exhibit "A"
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the power of' de gsdninia-urania rods is about 0.3 that of peak rod power. At the and 'of tha ' initial' cycly it is approximately 0.8 that of peak rod power.
Later in'11f $tite dowT 'of the gadolinia-urania fuel rods decreases.
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4.2.3.3.3 Safety Evaluation l
N The description shown the't the gado11nia-urania fuel rods meet the design basis requirements. ,
I g 4.2.3.3.4 Inspection and Testing -
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'!- f Che same rigid quality control requirements observed for standard UO2 fuel are
's.aployed in manufacturing gadclinia-urante.Wa1. Gadolinia-bearing UO2 fuel pellets of a given enrichment N d gadolinia concentration are maintained in separate groups throughout the manufacturing process. The percent anrichment y and gadolinia concentration charat.terizing a pelle't group are identified by a ..
stamp on the pellet.'
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/ Fuel rods are individually numbered pr,ior to loadinJ of fuel pellets into the fuel rods: (1) to identify which pellet group is to be loaded in each fuel red; (2) to identify which position in the fuel assembly each fuel rod is to be loaded; and (3) to , facilitate total material accountability for. a given project. Correct orie_ station of gadolivia-bearing rods within the fuel assembly is further assured by the innger upper and plus shanks for these rods.
The following quality control inspections are made: .
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}f a. gadolinia concentration in the gadolinia-urania powder blend is verified;
- b. sintered pellet V0 -Cd 2 02 3 solid-solution homogeneity across a fuel pellet is verified by examination of'retallographic specimens;
- c. gadolinia-urania pellet identificatiorils verified; and
- d. gadolinia-urania fuel red identification is checked.
All assemblies and rode of a given project are inspected to assure overall '
' accountability of fuel quantity and placement for the project.
4.2.3.4 Standby Liquid Control System 4.2.3.4.1 Design Bases 4 s. Getseral Design Bases \,
Safety D_esign Bases The standby liquid control system shall meet the following safety design bases: ,
(1) Backup capability for reactivity control shall be provided, y j independent of normal reactivity control provisions in the
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nucisar reactor, to be able to shut down the reactor if the I normal control ever becosas inoperative, l l
- (2)- ne' backup system shall have the capacity for controlling the i reactivity difference between the steady-state rated opera-ting condition of the reactor with voids and the- cold shutdown condition, including shutdown margin, to assure co g lete shutdown from the most reactive condition at any time in
. : core life.
(3) he time required for actuation and affectiveness of the backup control shall be consistent with the nuclear reactivity rate
-of change predicted between rated operatin5 and cold shutdown conditions. A f ast scram of the reactor or. operational control of fast reactivity transients is not specified to be accomplished-by this system.
- (4) Means shall be provided by which the functional performance capa-bility of the backup control system' componente can be verified periodically under conditions approaching actual use requirements.
A substitute solution, rather than the actual neutron absorber solution, can be injected into the reactor to test the operation of all components of the redundant control system.
(5) The neutron absorber shall be dispersed within the reactor core in sufficient quantity to provide a reasonable margin for laak-age or imperfect mixing.
(6) The system shall be reliable to a degree consistent with ita role as a special safety system; the possibility of uninten-tional or accidental shutdown of the reactor by this system shall be minimized. n 4.2.3.4.2 Description The standby liquid control system (Figure 4.2-23) is manually initiated from the main control room to pump a baron neutron absorber solution inco the reactor if the operator believes the reactor cannot be shut down or kept shut down with the control roda. Onea the operator decision for initiation of the SLC system is made, the design intent is to simplify the manual process by providing a " key locked" l switch. This prevents inadvertant injection of neutron absorber by the SLC system. ,
L -However, ' insertion of control rods is expected to assure prompt shutdown of the l
reactor should it be required.
The " key locked" control room switch is provided to assure positive action from i the ma'.n control room should the need arisa. Standard power plant procedural l controls are applied to _ the operation of the " key locked" control room switch.
The SLC system is required only to shut down the reactor and keep the reactor from going critical again as it cools.
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The SLC syscam is needed only in the inprobable event that not enough control
, M reds can 've ' inserted in the reactor core to accomplish shutdown and cooldown in the normal menner. ,
3 The boron solution tank, the test varar tank, the two positive-displace:nent pumps, the two explosive' valves, and associated local valves and controls are mounted
"- in the containment vessel. Thv. . f.pid is piped into the reactor vessel and dis-charged ne'ar the bottom of the core shroud so it mixes with the cooling water *
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- rising through the core (section 5.4 and Subsection 4.2.2).
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_ _ _-- _ _~ __. r,c l- The boron absorbs thermal neutrons and thereby terminates the nuclear fission chain reaction in the uranium fuel.
The speified neutron absorber solution is sodium pentaborate (Na2 3 10016 10H2O).
It is prepared by dissolving stoichiometric quantities of borax and boric acid in domineralized water. A sparger is provided in the tank for mixing, using air. To prevent system plugging, the tank outlet is raised above the bottom of the tank.
At all times when it is possible to make the reactor critical, the SLC system shall be able to deliver at least 3840 gallons of 13.4% sodium pentaborate solution or equivalent into the reactor (Figure 4.2-24). This is accomplished by placing 4860 lb of sodium pentaborate in the SLC tank and filling with domin-eralized water to at least the low level alarm point, and can be diluted with water up to the overflow level volume to allow for evaporation losses or to lower the saturation temperature.
The saturation temperature of the recommended solution is 59'F at the low level alarm volume and approximately 49'T At the tank overflow volume (Figure 4.2-25).
The equipment containing the solution is installed in a room in which the air temperature is 'to be maintained within the range of 60 to 105*F. In addition, a l heater system maintains the solution temperature at 75 to 85'F to prevent precipi-tation of the sodium pentaborate from the solution during storage. High or low temperature, or high or low liquid level, causes an alarm in the control room.
Each positive displacement pump is sized to inject the solution into the reactor in 50 to 125 min, independent of the enount of solution in the tank. The pump and system design pressure between the explosive valves and the pump discharge is 1400 peig. The two relief valves are set slightly under 1400 psig. To pre-vent bypass flow from one pump in case of relief valve failure in the line from the other pump, a check valve is installed downstream of each relief valve line in the pump discharge pipe.
The two explosive-actuated injection valves provide assurance of opening when needed and ensure that boron will not leak into the reactor even when the pumps are being tested.
Each explosive valve is closed by a plug in the inlet chamber. The plug is cir-cumscribed with a deep groove so the and will readily shear off when pushed with the valve plunger. This opens the inlet hole through the plug. The sheared end is pushed out of the way in the chamber; it is shaped so it will not block the ports after release.
The shearing plunger is actuated by an explosive charge with dual ignition primers inserted in the side chamber of the valve. Ignition circuit continuity is monitored by a trickle current, and an alarm occurs in the control room if either circuit opens. Indicator lights show which primary circuit opened.
The SLC system is actuated by a three-position kay-locked switch on the control room consols. This assures that switching from the "off" position is a deliber-ate act. Switching to ed.ther side starts an injection pump, actuates both of the explosive valves, and closes the reactor cleanup system outboard isolation valve to prevent loss or dilution of the boron.
. 4.Z-61 Am. 22 (12-9$74)
A green light is the control room indicates that power is available to the pump motor contactor and that the contactor is open (pump not running). A red light
. indicates that the contactor is closed (pump running).
If a pump liait, or exploetve valve light indicatsa that the liquid' '
l not be flowing, the operator can immediately, tu,o the switch to the otdar side.
which actuates the alternate. pump._ . Cross. piping nd. check valves, assure a. flow path, through either pump and either explosive valve. The chosen pump will start sven though its local switch at the pump is in the "stup" position for test or maintenance. Pump discharge pressure is also indicated in the control room.
Equipment drains and tank overflow are not piped to the radwaste system but to separate containers (such as 55-gal. drums) that can be removed and disposed of independently to prevent any trace of boron from inadvertently reaching the reactor.
Instrumentation consisting of solution temperature indication and control, solution level, and heater system status is provided locally at the storage tank.
4.2.3.4.3 Safety Evaluntion The standby liquid control system is a special safety system and is maintained in a standby operational status whenever the reactor is critical. The system is expected never to be needed for eefety reasons because of the large number of 4 Map-Maat control rods available to shut down the reactor.
However, to assure the availability of the SLC system, two sets of the components required to actuate the system - pumps and explosive valves - are provided in parallel redundancy.
The system is designed to bring the reactor from rated power to a cold shutdown at any tima in core life. The reactivity compensation provided will reduce reaccor power from rated to zero level and allow cooling the nuclear system to room temperature, witb the control rods remaining withdrawn in the rated power pattern. It inclue the reactivity gaina that result from complete decay of
.the re*ed power xenon inventory. It also includes the positive reactivity effect.. from elaminating steam voids, changing water density from hot to cold, reduced Doppler effect in uranium, reducing neutron leakage from boiling to cold, and decreasing control rod worth as the. moderator cools. The specified minimum final concentration of boron in the reactor core provides a margin of
-0.05 Ak for calculational uncertainties and assures a substantial shutdown margin.
The specified minimum average concentration of natural boron in the reactor to provide the specified shutdown margin, af ter operation of the SLC system, is 600 ppm (pm ts per million) (Figure 4.2-26). Calculation of the m4*4="a quantity of rodium pentaborate to be injected into the reactor coolant including recirculation loops, at 70*F and reactor nozmal water level.' The result is increased by 25% to allow for imperfect mixing and leakage. An additional 250 ppa is provided to accommodate dilution by the RER system in
, 4.2-62 Am.22(12-9-7d)
the shutdown cooling mode. This concentration will be achieved if the solu-tion is prepared as defined in Subsection 4.2.3.4.2 and maintained above saturation temperature.
Cooldown of the nuclear system will require a minimum of several hours to remove the thermal energy stored in t ac reactor, cooling water, and asso-ciated equipment. The controlled limit for the reactor vessel cooldown is
-100*F/h, and normal operating temperature is approximately 550'7. Use of the main condenser and various shutdown cooling systems requires 10 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to lower the reactor vessel to room temperature (70*F); this is the condition of maximum reactivity and, therefore, the condition that requires the maximum con-centration of boron.
The specified boron injection rate is limited to the range of 6 to 25 ppe/ min.
The lower rate assures that the boron is injected into the reactor in approxi-
. mately two hours. This resulting reactivity insertion'is considerably quicker than that covered by the cooldown. The upper limit injection rate assures that thera is sufficient mixing so the boron does not recirculate through the core in uneven concentrations that could possibly cause reactor power to rise and fall cyclically.
The SLC system equipment essential for injection of neutron absorber solution into the reactor is designed as Category I (seismic) for withstanding the speci-find earthquake loadings (Section 3). The system piping and equipment are designed, installed, and tested in accordance with requirements stated in Chapter 3.
The SLC system is required to be operable in the event of a station power failure; therefore, the pumps, heaters, valves, and controls are powered from the standby a-c power supply or d-c power in the absence of normal power. The pumps and valves are powered and controlled from separate buses and circuits so that a single failure will not prevent system operation.
The SLC system and pumps have sufficient pressure margin, up to the system relief valve setting of approximately 1400 psig, to assure solution injection into the reactor above the normal pressure in the bottom of the reactor. The nuclear system relief and safety valves begin to relieve pressure above approxi-mately 1100 psig. Therefore, the SLC system positive displacement pumps cannot
- overpressurize the nuclear system.
Only one of the two SLC pumps is needed for system operation. If one pump is found to be inoperable, there is no immediate threat to shutdown capability, and reacrer operation can continue during repairs. The time during which one redundant component upstream of the explosive valves may be cut of operation should be consistent with the following: (1) the probability of failure of both the control rod shutdown capability and the alternate component in the SLC system; and (2) the' fact that nuclear system cooldown takes several hours while liquid control solution injection takes approximately two hours. Since this probability is small, considerable time is available for repairing and restoring the SLC system to an operable condition while reactor operation con-
. tinues. Assurance that the system will still fulfill its function during repairs is obtained by denonstrating operation of the operable pump.
4.2-63
g,g l .
4.2.3.4.4 Inspection and Testing Operational testing of the SLC system is performed in at least two parts to avoid inadvertently injecting boron into the reactor.
With the valves to and from the storage tank closed and the three valves to and from the test tank opened, domineralized water in the test tank can be
' recirculated by locally _ starting either pump.
The injection portion of the system can be functionally tested by valving the injection lines to the test tank and actuating the system from the control room. Both injection valves open on actuation. System operation is indicated
, in the control room.
After functional tests, the injection valve ahear plugs and explosive charges must be replaced and all the valves returned to their normal positions.
After closing a local locked-open valve to the reactor, leakage through the injection valves can be detected by opening valves at a test connection in the line-between the containment isolation check valves. Position indicator lights in the control room indicate that the local valve is closed for tests or open and ready for operation. Leakage from the reactor through the first check valve can be detected by opening the same test connection when the reactor is pressurized.
The test tank contains demineralized water for approximately 3 minutes of pump operation. Demineralized water from the makeup system or the condensate storage system is available for refilling or flushing the system.
Should the boron solution ever be injected into the reactor, either intention-ally or inadvertently, then after making certain that the normal reactivity controls will keep the reactor subcritical, the boron is removed from the reac-tor coolant system by flushing for gross dilution followed by operating the reactor cleanup system. There is practically no effect on reactor operations when the boren concentration has been reduced below approximately 50 ppm.
The concentration of the sodium pentaborate in the solution tank is determined periodically by chemical analysis.
4.2-64 . I g - -m - +
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t s-
- a. Reduadsat differential temperature and ambient temperature switches sense acic and RER equipment area ventilation air inlet and outlet high temperature or high ambient temperature. .,
~
- b. Redundant differential temperature and ambient temperature switches sense 3
l RcIC pipe routing area ventilation air talet and outlet high temperatues or .
bish ambient temperature.
- c. Reduadant differential pressure tramanitters sense RCIC or RER/RCIC stesa line high flev er instruneat line break.
- d. Redundant pressure treammitters sense RCIC tarbime enhaust diaphragm high pressure. Both tramanitters in oos of two chaamels anst sense high pressure to cause tsaaltion. .-
,- - - -- ..r..=-,. . . . . . . . . . . . ..
- e. A pressare treammitter senses BCIC loe stamis seyply pressare. . '.. .. ..
. a .
. 1 .
The RCIC sysua may be isolated after initiation by the control room operator' by actuation of a switch which caases the eetboard steaaline isolatica valve to close. ,'*
7.4.1.2 S*--m i Liemid Centrol Systan (ELCE)
- a. ELC3 Functies .
The standby Lignid Centrol System (ses ,Secties. 9.3.5) instrussataties is .
destaaed to init4 ate tsjecties of a 11.pid esatsen absorber into the reactor. Other lastrumentaties is provided to asiatain this liquid chastical soluties well abore saturaties tagratare la readiness for tajecties.
The SICS is a bachap s. *. bod of manually shutting dews the reactor to cold shutdown conditions froh surnal operaties or fois anticipated transiest conditions when contret rod insertion capability is lost.
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e.- The neutron absorber will be dispersed within the reactor core in sufficient quantity to provide a reasonsble margin for leakage or i perfect =4v4as.
- f. The syste:s is reliable" to a degree consistent with its role as a special ,
safety system; the possibility of unintentional or accidental shutdown of the reactor by this system is minimized.
Svstem bescription
~
9.3.5.2 The standby liquid control system (see Figure 9.3-19) is manually initiated in the main control room to pump a boron neutron absorber solution into the reactor if the operator determines the reactor cannot be shut down or kept shut down with the control rods. bucetheoperatordecisionforinitiationof the SLC system is made, the design intent is to si=plify the manus 1 process by i providing a keylocked switch.' This pre; tents inadvertent injection of neutron ab orber by the SLC system. .However, insertion of the control rods is
. e .puted to assure prc=pt shutdown of the reacter should it be required.
i keylock 4 .sisitch is provided in the control room to assure positive action from the mata control room should the need arise. Procedural controls are
.. applied to the operation of the keylocked control room switch. '
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The SLC system is required only to shut down the rescf.or and keep the reactor -
'from going critical as it cools. s The SLC system is needed only in the improbable event that not enough control rods c:n be inserted in the reactor core to accomplish shutdown and cooldown in the nor: sal manner.
The boren solutiot tanh, the test water tank, the two positive displa-nt pcmps, the two expl sive valves, the two motor operated tsni shutoff valves, a=d associated local valves a d controls are located in the centsinaent. The liquid is piped into the reactor vessel and discharged near the bot. tom of the core shroud so it mixes with the cooling water rising through the cora: (see ,
Eectien: 5.3, 3.9.3 and 3.9.5-) .
9.3-19
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The RPT design shall:
l.- . Meet IEEE 323-1974 and 344-1975, or be consistent with existing plant design. requirements;
- 2. Meet IEEE 279,.379 and 384 (except for the Low Frequency Motor /
, Generator breakers);.
.- ~- 3 . . Provide for inservice testability (except for action of final breakers).
- c. Feedwater Runback Upon the receipt of a high pressure signal from the RRCS logic including confirmation of no-scram, feedvater flow is to be limited, thereby reducing power and steam discharge to the suppression pool.
The feedwater runback design shall:
.1. Use control grade equipment, and
- 2. Provide manual operation override to allow an increase in feedwater flow, if needed and available.
III. ' Standby Liquid Control-System The standby liquid control system (SLCS) action is to be initiated manually on a failure to scram condition in accordance with Emergency Instructions. Simultaneous operation of both pumps at full capacity.
(86 gpm total) will control the nuclear fission chain reaction and thereby maintain suppression pool temperatures within specified limits.
The SLCS design shall:
- 1. . Provide a manual sodium pentaborate solution injection function for
~both loops simultaneously operated only from the Control Room; Am. 11 (2-15-83) 15C-5 L
- 2. Provide for replenishment capability of the SLCS tank with mixed sodium pentaborate sc.r; ion from outside the containment;
- 3. Provide capability for periodic functional tests;
- 4. Assure that no single active logic component failure can prevent its function; and
- 5. Meet IEEE 323-1974 and 344-1975 or be consistent with existing plant design requirements.
15C.6 SCRAM DISCHARCE VOLUME MODIFICATIONS Additionally, control rod drive system scram discharge volume shall be modified to minimize the potential for failure of the scram function from unavailability of this volume. The design modification will consist of the addition of redundant instrument vo'lume water level sensors to the control rod drive hydraulic system and instrument line piping modifications. The design change shall:
- a. Provide redundant lE sensors;
- b. Provide redundant vent and drain valves.
15C.7 ATWS EVENT AND RESULTS In order to study the reactor responses with the injection of the boron solution, the Alternate Rod Insertion (ARI) is deliberately ignored in this study, because with'ARI, there is no need for boron injection. Consequently, five anticipated events are selected as initiative transients since they can result in highest responses in comparison with the safety criteria. These initiating transients are:
.a. MSIV Closure Event - This transient when coupled with postulated normal scram system failure produces high RPV pressure, heat flux and suppression pool water temperature.
Am. 11 (2-15-83) 15C-6
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& .h . ru.uwsAtmo stoc. e ruauC souAntSewg PE.RRY NUCLEAR P0".ER P!. ANT PROJECT
. CLrvrLAmo.
The Best l.ccation m :?
omo moi
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FBILD C CNST AUCTION MA wact utNT Cac Aht2A780N CEI NuCL!AR ENctNitalNG OtPARTMENT ANO LA tSER ENGINti AS. INC.
February Po,1930 #
- v_ cn /qN.hho RECEIVED - --
FEB 2 21980 ReS?onSe t*
<fY-GIN /CEI-1277a Mr. R. C. Mitchell (MC 392) PNPP.SOjDC Generalz lectric Cc pany 175 curtner Avenue Ferry Nuclear Pcwer Planc Units 1 L San Jose, California 05125 ATJS Cost and. Schedule I:;act Dee.: 3cb:
We he.ve estin.ted the cost and schedule 1 pact for Ferry of the ATJS =cd.ificacic:
reqc.irements ii !~JRIG 0460 alternate 3 and 4 Tnese estimates are based on the
=odification require =ents applicable to Ferry, given in attached Table I which was abstracted frc=.ycur GIN /CII-1277 letter.
Ter each of the =ciificatica categories within Alternate 3 and h, we have de-g) v ter
- ei whether Ferry already cc= plies or regaires =cdification, and wheth'er additic:a1 space is needed.
This info =ation is shown in attached Table II.
additic: Table II presents cur best est'-=te of the d.irect cost cf thesa -d'- In ficatic:s, and the expected delay in c;erating license >='a % single largest cost ite: is the SLC capacity increase of Alternate 4 Tne direct costs of Table II are given as 1930 dollars, and include culy the ecs:s of engineering, equip =ent and installation.
Tne additienal costs that shculd be added to this include the indirect costs and the outage / delay costs.
The indirect costs include Contingency /Iscalation/A?'JDC which we estinate as 0 7y the direct' costs, and Licensing /Operatien/Kiintenance which we estimate as 0.6X the direct cests.
In addition to these costs another =ajor impact is the ces cf icst revenues iue to const:acticn delays, and due to outages frc: inadvertent autc-beren-in'ec-icn.
As shown in Table II, we estimate significant constructica delays =cstly due to the Alternate 4 SLC hardware changes. Tnese result in an estinated 12 week delay at a cost of $275,000 per day. We also believe tlere is a high prcbability fcr at least c:e inadvertent auto-injection during plant life, with either Alternate 3 or 4, resulting in a 10 day outage for boron cleanup at a cost of $500,C00 per day.
Tne total of all AZ,*S =cdification costs for Ferry are s"--=rized in Table III. We believe these are realistic estimates, that are perhaps even on the censervative side. A simple ec ; arisen of the direct-plus-indirect cests of $4.37 =illion for Alternate 3 with $11.27 =illion for Alternate 4 shews the great additional h expense cf incorporating Alternate 4 We do not believe that the incremental safet/ benefits of Alternate 4 justify this great additienal cost.
b
ruary 22, l go
. hga -2 i
), . -
- de would be glaf. to supply additional' details for this assess:ent. If you have. any questions please call us.
Very t: .:ly. yours , #'
b Henry A. ?utre Senior Design Ingineer EA?/ fab
- Attachnent 3 cc: G. W. Grcscup
- 3. L. Parkley .
L. O. 3eck P. 3. Gudikunst .
J. A. Kline -
SO/DC.
g ros D.-P.. Green h
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7 M' '. .
G E N E R A L h El.E CTRIC EtEcraic uritiry SATES DIVISION
' GENERAL ELECTRIC COMPANY,1000 LAKESIDE AVENUE, CLEVELAND, ohio 44114 Phone (216) 523-6000 RE: PERRY NUCLEAR POWER PLANT n -
..n '
UNITS 1 & 2 ANTICIPATED ANSIEN S WITHOUT SCRAM g555
$ P~N I s# . { "Yf QUOTATION . 149-B J _ i ECEi 59
= , ,
Cleveland Electric Illuminating Company Perry Nuclear Power Plant P. O. Box 97 Perry, Ohio 44081 FILE PERRY PROJECT Mr. D. J. Zupan ' i Attention: PRF-
Dear Don:
General Electric is pleased to offer this revised quotation, relating to the Alternate 3A AT'#S design, for:
- 1. ATWS Design York *
, 2. Project Design S,ervices- _.
- 3. ATWS Equipment -
This quotation supersedes Quotation No. -_149-A and in accordance with your February 9, 1981 letter requotes this on the basis of a generic and unique breakdown.
GE's scope of work for ATWS Design Work is described in Section 3.1 of the attached scooe of work do_cument. Section 3.2 describes GE's scope of work for Project Design Services, and Section 4 describes the estimated scope of Equipment supply,. ,
PRICE _ _
The price for performance of IG'eneri~c ATWS~ Des 1gn' fork l is,
~
for the Perry Project, if at D a#st' "e"Eth'eY ufili'Ef* e s"acrept this quotation by submitting signed copies of an Amendment bkfore June 15, 1981. If at least ~ other utilities accept this quotation by submitting signed copies of an Amendment before June 15, 1981, the pric'e per project will be J. The price will be subject to adjustment in accordance with tne attached Amendment.
i
\
l e m "e" ,
/
/ -
GENER At h ELECTRIC lir . ' D . J. Zupan - 2- April 13, 1981
'fal$d$e6 e s'ign. Servi 6Es7 will be oarformed on a Time and -
hiaterial (T&M) oasis,~using GE's standard T&3i rates and charges in eIIect when the work is performed 3 The quantity of Project Design survices required wlli cepend upon whether the prescribed equipment i
is in_ stalled M Ure or after fuel nrH gg. If installation is eforefu.elloading}weestimate -
manhours}will be reauired resu ,Jng n an estimated total bi intui aYpYoximately .
Erhis' oR1_on 1rTvail'EH1'e for TJn~1ts hTaiy_ing_a_fuMEaling_af.ren s -
CJuT1n98373 wever, con _sidering J ourJ!ay3 4 98,3,fau,eLAcading, date, we are presently maItiWgMhe"a'rFaTgements necessary ~to supports ,
that date. If installation follows fuel loading, we_ estimate (
anhours will be required with an 'ep.t-imated-T6ElT51111ngi6t och-b1 ling estimates are based upb'n January 1981 billing ratem .
Ve will advise you of the firm price t,o be paid for Equipment when that Equipment is defin_ed. The price will be established in accordance with our then standard pricing practices. The current estimate of the Equipment price is $. '
per unit,. The price will be subject to adjustment in accordance with the attached Amendment.
PAY 11ENT TER3!S The table.
terms of payment for ATWS Design Work will be those in the following CUlfULATIVE DEPOSIT _ PERCENT D.U E " --- -- -- - PERCENT Payment for Tal! Project Design Services will_b_e_due within 30 days _of invoice f,or work.perf6rmed. - * ~
PROPRIETARY INFOR5fATION '
GE anticipates that design work performed under this quotation will rtsult in information that must remain proprietary to GE. The pro-visions of Article XVII (Information) of our NSSS Contract will apply.
TERi!S AND CONDITIONS .
The terms and conditions for ATWS Design Work and ATWS Equipment will be those of the attached Amendment. Project Design Services will be parf.ormed in accordance with Article XXVII (Additional ~ Services) of the NSSS Contract. ;
l
GEtJ ER A L $ ELECTRIC ) ,
Mr. D. J. Zupan April _13n 1981 QUOTATION VALIDITY This quotation will remain valid until June 15, 1981, and our obligation to perform the work as outlined in this quotation is conditioned upon acceptance of this quotation by CEI and five additional utilities by June 15, 1981.
We encourage your prompt review of this important matter, and we will be pleased to discuss any questions you may have regarding this quotation for the ATWS work defined in the Technical Description.'
In addition, we have a team available to visit you personally to discuss all aspects of this quotation, should you so desire.
Very truly yours, ytw s- (
FRANK MIOTTI - GENERATION SALES /SR. APPLICATION ENGINEER WFM/11g cc: Mr. H. A. Putre, CEI Mr. L. O. Beck, CEI e
i k
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g c. % e-o-ogQ 9
4 N~
,/
SCOPE OF WORK FOR ANTICIPATED TRANSIENTS WITHOUT SCRAM ALTERNATE 3A l-O 1
DESIGN AND EQUIPMENT 9
4/3/81 e
sup e
S l
. . . . . ii. _ _ _ _ _ _ _ . _ _ _
L- .y -? ., .
SECTION 1
- INTRODUCTION This document describes General Electric's proposed scope of work for ATWS Design Work, Project Design Services and supply of ATWS Equipment O
e G
8 e ao ap 4
1-1 I
'172-P1 i'i m . .m , - i . _ . _ . _ . _ _ . . _ _ _ _ _ _ . _ . _ _ . . _. _ _ _ . _ _ _ _ . _ _ _ _ _ _ __
~
,' a. allow BWR performance to meet acceptance criteria described in Appendix B including consideration of an A WS event without ARI; '
- b. have a reliability objective of 0.96 per demand;
- c. use' control g'rade equipment; and
- d. provide manual operation override to allow an increase in feedwater flow, if needed and available.
2.6 Standby Liquid Control System The standby liquid control system (SLCS) action is to be initiated automatically on receipt of signals indicating an ATWS event is in progress. Simultaneous' operation of both pumps. at full capacity (86 gpm total) is required to maintain suppression pool temperatures within specified limits.
The SLCS design will: '
- a. provide an automatic AWS-injection function for both loops simultaneously;
- b. provide a manual ATWS. injection f nction fo @ h loop D simultaneously operated only from the Control Room;
- c. require replenishment capability of the SLCS tank with mixed C
sodium pentaborate solution from outside the containment;
- d. consider the hydrodynamic pulsation effects of two pump operation;
~
i A-8 172-P27 S
y . _ . _ - . . . . - - , . - . , . . - . _ **[-._
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- e. r'equire automatic SLCS injection through the HPCS. lines (BWR/5 and BWR/6) or through the jet pump instrument lines (BWR/4).
- f. provide automatic Reactor Water Cleanup System isolation;
- g. provide for periodic functional tests of the syste'm in both the manual and the automatic ATVS injection modes;
- h. minimize undesired injection of the borated liquid into the vessel by design of the logic and sensors to have a low failure rate;
- i. accommodate the containment pressure and temperature environment during the period of needed ATVS operation, and for BWR/6 during any predicted operation of containment spray;
- j. seet requirements 1.1 through 1.9 of Append,ix A;
- 1. include actuation circuits which have a reliability objective c h er demand and which are separate from RPS and containment Jsolation circuits;
- m. have a reliabili.ty objective for delivering boron at full capacity of j.95' per demand;
- n. assure that no single active logic component failure can prevent its funcu on; anu 58 A-9 172-P28
. . - me--_ e- mm- -- - m . - - - - - - - -
- p. -*
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- o. u'se equipment, qualified either by analysis or test, to assure meeting the predicted conditions associated with the ATWS event.-
'2.7 Containment Ventilation Isolation Valves To prevent radiological releases to the environs if there are significant fuel failures, the containment ventilation isolation valves are to be closed upon high containment radiation. _This
~ isolation logic is to be separate from the RPS.
The containment isolation interface specifications will:
y a. require conformance with Appendix B for the ATVS events given f in Appendix C;
- b. specify that an electrical failure in.the RPS,.that could prevent a scram, will not prevent the containment isolation function; and
- c. specify that use of circuits and valves,. qualified by either analysis or test, to assure meeting the predicted condi,tions associated with the ATVS event.
2.8 Scram Discharge Volume Modifications Control rod drive system scram discharge volume modifications will be designed to minimize the potential for common mode failure of the scram function due to drain line failure and thus unavailability of this volume. The design modification will consist of the addition of instrument volume water level sensors to the control rod drive hydraulic system and instrument line piping modifications. The design change will:
A - 10 172-P29
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- .. . . .% iUL 2 91982 of UNITED STATES
{ r ir , g_ , NUCLEAR REGULATORY COMMISSION . crriet us
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LDocketNo. 50-440/441 '
' MEMORANDUM FOR: A. Schwencer, Chief Licensing Branch No. 2 Division of Licensing FROM: J. Stefano, Project Manager Licensing Branch No. 2
. Division of Licensing
SUBJECT:
SUMMARY
REPORT ON STAFF MEETING WITH CEI -- PERRY CONTINGENCY PLANS FOR ATWS .
The subject meeting was held on July 20, 1982 in conference room MNBS 6110 at the-request of CEI. The list of attendees is enclosed.
' CEI advised the staff of-its intent to install a systems upgrade for ATWS-in Units 1 and 2 different from the design docketed in the FSAR and which the staff found acceptable in the Perry SER. Generally, system upgrade for' ATWS will . include: .a manually operated standby liquid control system
'which doubles the flow rate capability of the current design (from A6J8' 49dh gpm); safety grade feedwater/ recirculation pump trips, level sensors and -
pressure sensors; an alternate control rod insertion capability; a modified scram ' discharge volume. A final design, currently being completed by GE, is expected by November 1982. . Within the next two weeks CEI plans to document to the staff a summary overview of the design changes for ATWS to be installed in the plant, and a specific date when the final design will be submitted ' for staff approval, m.
Joh J. S fano
-h
' Project nager Li nsin Bran No. 2 -
ision f censing
Enclosure:
. Attendance List
,cc: H. Abelson, LB#2 N. Floravanta, ASB G. Thomas, RSS .
J. Mauck, ICSB R. Stevens, ICSB -
J. Clifford, PRTB TR. Tedesco, AD/L s k "$ "
s .-
. Perry
~
Mr. Dalwyn R. Davidson ,-
Vice President, Engineering 6 The Cleveland Electric Illuminating Ccmpany P. 0.- Box 5000 Cleveland, Ohio 44101 cc: Gerald Charnof f, Esq. _
. Shaw, Pittman, Potts & Trowbridge 1800 M Street, N. W.
Washington, D. C. 20006 Donald H. Hauser, Esq. _
The Cleveland Electric Illuminating Company P. O. Box 5000
. Cleveland, Ohio 44101 Resident Inspector's Off. ice U.S. Nuclear Regulatory Commission Parmly at Center Road JPerry Ohio 44081 Donald T. E: zone, Esq.
' Assistant Prosecuting Attorney 105 Main Street.
- Lake County Administration Center Painesville, Ohio 44077 Daniel D. Wilt Wegman, Hesiler & Vanderberg 7301 Chippewa Road, Suite 102 Brecksville, Ohio 44141 Ms. Sue Hiatt OCRE Interim Representative 8275 Munson Mentor, Ohio 44060 -
Terry lodge, Esq.
915 Spitzer Building Toledo, Ohio 43604 -
John G. Cardinal, Esq.
Prosecuting Attorney Ashtabula County Courthouse
- Jefferson, Ohio 44047 D
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Enclosure 1 MEETING TO DISCUSS CEI ATWS CONTINGENCY PLAN July 20,1982 '
A Name Firm. .
R.=T. George PECo.
T. Shannon PEco.
J. Nelson E. Shel ton - Quadrex R. W./Skuarek Bechtel Power PSEGG S. W. Vail H. L. Hrenda Bechtel Power CEI W. E. Coleman CEI R.'C. Mitchell GE
. E. M. Buzzelli .CEI T. C. Houghton KMC C. W..Veprek
' H. C. Pfefferler GE D. T. Shen GE R. Stevens NRC/ICSB J. Clifford NRC/DHFS/PTRB
- N. Fioravante G. Tnomas NRC/NRR/DSI/ASB NRC/DSI/RSB a
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M P.O. sox 97 5 P ERRY. CHIO 44c81 m TEL EPHON E (216) 259 3737 m ADDRESS-to CENTER RO AD
, Serving The Best Location in the Nation August 6, 1982 PY-CEI/ GEN-597 Mr. R.'C.Mitchell Project Manager
-General Electric company
'175 Curtner Avenue ~~
San Jose, CA'95125 Re: Perry Nuclear Power Plant Units #1 and #2 Review of NEDE 25518
Dear Mr.~Mitchell:
The attachment to'this letter provides comments to the subject report. The report was' reviewed for consistency with the Perry plant design and current GE RRCS design documents, as well as for reference to the automatic initiation of SLCS flow and RWCU isolation.
'.Because of uncertainty in the eventual m_e h d.__of SLCS initiation, it-is suggested that references to the method be deleted from thT body of the report.
The method of actuation should' be described in one section only, to minimize
. future changes.
The remaining coc=ents are apparently due to system design changes made since the report was written. If a revision to this report is in progress, the comments should be' checked against this revision.
It is requested that a corrected final report be provided by October of this year, to support the planned submittal of the ATWS mitigation design information to the NRC..
Very truly yours,
~
. . z. J H. L. Hrenda Responsible Engineer G. L H. A. Putre Senior Engineer HLH/iw .
u . Attachment ec: 'E. M. Buzzelli - R230 D.'-R. Green - W225 yLJ n H. A..Putre - W240 {M{.84 Ii
Attachment to
, PY-CEI/ GEN-597 Page 1 of 2 August 6, 1982 The purpose of this attachment is to provide CEI comments to NEDE-25518,
" Design Analysis and SAR Inputs for ATWS Performance and Standby Liquid Control System, Perry Plant," dated December, 1981.
- 1. Page 1 Footnote 1, Item 4 - Delete. Should be " Automatic annunciation of ATWS event."
- 2. Page'l Footnote 1, Item 5 - Delete. This is not provided by -
the RRCS design.
- 3. Section 4.1.3. 2nd Paracraph - The SLCS pumps are not redundant.
Operation of both pumps is required.
4 Criterion 29 - SLCS valves and pumps are not redundant.
- 5. APCS B 3-1 and MEB 3-1
- a. First Paragraph - The SLC system is not located in its own compartment.
- b. Second Paragraph, Second Sentence - The system is required under transient as well as normal full power operation.
- 6. Section~5.1.9 - Modify to describe manual SLCS initiation.
- 7. Section 5.1.12 - Clarify extent of two-phase mixture inside the shrould.
As written, it appears that the entire volume inside the shrould is two-phase mixture.
8 Table 5.2, Note 2, Last Line - Clarify irtail7r .
- 9. Table 5.3.1
- a. Sequence 4 - Add recire runback (LFMG).
- b. Sequence 8 - Should be LFMG tripped, time should be 29 seconds,
- c. Sequence 13 - Should be "SLCS initiated."
- 10. Table 5.4.1
- a. Sequence 4 - Add recirc runback (LFMG).
- b. Sequence 8 - Time should be 26 secons. Add recire pump trip.
c .- Sequence 9 - Delete
- d. Sequsnee 12 - Should be "SLCS initiated."
- 11. Table 5.5-1
- a. Sequence 3 - ARI and SLCS logic is not triggered by a failure to scram.
~'
1 . .,.
Attachment to
.j -
PY-CEI/ GEN-597
+. Y s j Page 2 of 2 August 6, 1982 y
.t
- 12. Table 5.6.1 ,
~
- a. Sequence 7 - Recire pumps are runback (LFMG).
Recire pump trip activates with feedwater limit. -
& b. Sequence'11 - Add recirc pump trip.
- c. Sequence 16 - Should be "SLCS initiated."
q 13. Table 5.7.1 ,.
, 6 j
- a. Sequence 15 - Should ae "SLCS initiated."
1'm h 14 Table 5.7.2 S -
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- a. Sequence 11 - Level 2 is an ATWS setpoint. Will an LniG
~
trip occur? i.
- b. After sequence 12'g LniG trip, if not in sequence 11.
g 1 }
- c. Sequence 16 - Should b,e "SLCS initiated."
,\i q '
.{ -
, l 'i . Table 5.9.1 ' ,
. \
'a . , Sequence'.4 - Recire pumps are runback (LFMG).
b., Se'quence'd - LFMG ir tripped, time is 29 seconds (same as sequence 9),
- c. -Sequence ~13 - Should be "SLCS initiated."
% s
- 16. . Tele 5.1? ._1_
\
- a. Sequence 4 Recire pumps are ranback (LFMG).
- b. Sequences 8 and 9 - LFMG is tripped, FW is runback at 29
( '
c.
seconds. '
Sequence l'h - St.:uld be "SLCS initiated."
? .
17 Table 5.14.1 i .
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, a. Sequence A. . Peeirc punjs ~are runback (LFMG).
, ( N ;
- 18. Table 5.14.2 '% ,',
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' (a a. Sequence 4 - Recire 'p ups 'are unback (LFMG).
y L % %- i
~19. Table 5.15.1 ,
4 -
3
- a. Sequence ( # Reciy C pumps are runback (LFMG).
s q- b. Sequence 8 - Recire pe:qs are tripped (LntG) and FW runback at w4,"
i 29 .ceconds.l' s
\e. Sequence 14 3 Should be "SLCS initiated."
s- ,
- 20. Table 5.15.2 s '
s y ( ,.
- a. Sequence 4 - Recirc pumps are runback (LRIC).
- b. S quence 8 - I RC trip and FW runback sre initiated at 29 seconds.
, , c. Sequence 13 - Should be'"SLCS initiated."
- 4
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E L E C T P;1 C I L !:U kil N ATI N G COMPANY
[p THE CLEVELAND
. p o sex w>o . ettvtuno. osio ncs . 1ettanost aisierr.osoo . ettu'.uuritwo e co
. n p..a. m.;w Serving The Best localicn in the Nation Iwyn R. Davidson August 13, 1982
- e atsotut .
'lld (MClwllRING AND con 5tRUCitoH l
l Mr. A. Schwencer, Chief ,
Lic;nsing Branch No. 2 -
Division of Licensing ,
U. S.' Nuclear Regulatory Commission -
Wtshington, D.~ C. 20535 .
Perry Nuclear Power Plant
- Docket Nos. 50-440; 50-401 ATW5 Mitiration Desien Features _
Darr Mr. Schwencer: ,
~
between CE! and members of NRR, we discussed our In a meeting .on July 20, 1932, plans for changing several systems associated with the mitiga ATWS event.
current ATWS rulemaking schedule have made it necessary for us to anticipate poten-tial future requirements. We believe it to be in our best interest to mod!!y !! the current design and Install these systems during our construction as opposed to wa! ng until .the construction of these systems impacts our construction schedule or The !nclusion of these systems is based on the proposed ru!emaking and is oper ations.
not based'on a be!!ef by CE! that -these systems are needed to mitigate en ATWS event. As such, we maintain'our support of industry comments on the proposed ATWS rule. .
The basic changes to be made to the Perry plant include the following:
An increased flow capaelty for the Standby Liquid Control System (1) .
- from 43 gpm to 86 gpm. Thls will involve increasing the size of both pumps',,,s.yct!cn lines as well as changing the reactor vesse! inie oo;nt to :ne Ji:C 5 5 e.;l,er. A!! hough 15e design incluces both mat (a!
an?TCTomatic inMistion casab!!!!y, only m:nual in!!!atien wi!! Se
~ '
M-mrfisE The existing pumps w!!! be Eed.
(2) upgrade to safety grade of the Recirculation Pump Trip initiation .
circultry.
U) A control grade feedwater runback feature.
(4)
An Alternate Red insertion system which is redundant to the Reactor Protection System scram log!c.
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- A. Schwencer
- a. ,
, ATW5 Mit!gation f ' =-
r August 13,1982
'1'^ . Page 2 f
p, The detai;s of the above described design wi!! be submitted as an am'endment' .-
.to the Petrry FSAR by Oanuary 1933.L
~
We be!! eve that this design along with. appropriate erriergency chrating procedures
.j!-. and training adequately address the ATW5 Issue for PNFP.
A. :. .
^ Very truly yours, '
~ - .
'f .
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Dalwy. R. Davidson Vice ? resident
.' System Engineering and Construction c DRD:WEC:mb ' '. <
a ccr .hy Si! berg, Esq.
d ehn Stef7_no Max G!Idner A
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j p.c. sox s7- a # ERRY. CHIC 44:81 a T EL. EP H o N E ( 216) 259-37 37 5 ADDRESS-1t CENTER Rc AD Serving The Best Loca:ic- on :he Naticn August 9,~1982 i PY-CEI/ GEN-598
?!r. R. C. !!itchell ,
Project > tanager Gcneral Electric Cc=pany 175 Curtner Avenue San Jose, CA 95125 Re: PNPP Units #1 and #2
, Quotation No. 149A Request for >todification DGar lir. > itchell:
As previeusly discussed, to prevent inadvertent injections of boron into the reac:or vessel, i: has been deter =ined that the automatic SLCS initiation and RWCU isolation provided by the subjec: quote should be replaced by a manual initiation syste=.
Because of uncertain:y concerning the final ATWS =itigation system require =ents this change shculd not be incorporated on the panels prior to delivery. The =anual ini:iation system design should detail the changes required to the docu=ents and equip =en so that these changes can be made after equipment delivery in the event
=anual initiation of the SLCS is ~ allowed by the final ATWS rule, or if no rule is issued prior .to startup of the Perry Plant.
The =r.nual initiation design should include annunciators to ensure the opera:or is infor=ed of the even: and is 'able to deter =ine the necessity for SLCS initia: ion and RECU isolation vi:hin the 120-second period available. The operator vill =ake this deter =ination based on the APR11 readings and/or rod position indication af ter the 25-second delay associated with ARI operation. Since the SLCS initiation /RECU isola: ion ti=e has no: been changed no further plant analysis should be required. An assessmen: ~s hould be =ade of the i= pact of this change on syste= reliability.
To supper: licensing s:hedules, it is requested that this design be co=pleted and issued by Sep:e=ber 15, 1982. Additional =anhours should be provided by T&li esti= ate by Augus: 23, 1982 General Electric is authorized to proceed on this design subject to approval of the estimated =anhours.
~
Very truly yours,
{ - A H..L. Hrenda
/
Responsible Engineer ELH/iv H. A. Putre ec: E. !!. Su: celli - R230' Senior Engineer D. R. Green - W225
'E. C. Will=an - W250 .
NDS File 41.2/C22/SP-li P0/DC - R290
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- . GENERAL ELECTRIC COMPANY,175 CURTNER AVE., SAN JOSE, CALIFORNIA 95125
,J .._MC 392, (408) 925-2755 Ngy 1 2 h~~o?-
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. November 8, 1982 Responds to:
- 'PY-CEI/ GEN-598 PY-GEN /CEI-1752 INFORMATION RECEIVED g .. . ,
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Hoy 161932
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FROM: DOC. CONTROL .vDATE: ///j.p/f:-
.r Mr. P. A. Nichols
- g.g.Clev, eland Electrie Illuminating Co.
COMES Toil f /./ e Wu tU
_. f j fp; g,;>ma,., .jm,u /j
[L . ...;yP.O; Box 84-10 Center ~ Road, u 77. , , y m . p p n c.j w : v L L.-
. . . if 4):Petry, OH 44081 .-
e wmA f m 4 ' Ys"N.~ . , uicum_eurrmi J.:, 1 7='i5 t 5 H.'L. Hrenda N M// *'M 1/- '
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Dear w Mr. Nichols:
- 1. . _
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7 .SUEJECT: MANUAL SLCS INITIATION MODIFICATION DOCURENTS t.
E This letter transmits two copies each of the following documents in 2.2 5 response to CEI's request.for modification of the ATWS 3A design for E. 1..
m manual initiation of SLCS.
_ 5 . . .
. .- ., p 23A1325 - 2 Redundant Reactivity Composite Requirement - Design
- . '. .g
- . . . Specification iP NT 11 L944E671 Modification Diagram -;RRCS (FCD) - '- ~
T.
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--p.- :n :.y -. R y j k ; & -_ y ._ 2:.; = ; %;. & .-i
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d.?.' : . 944E672 .? g . .
- k. $pG.Modif.iiation Diagram'-?RRCS TELEM) E. 3.r K ~ .' Q}- ".3: Modific.-h:.:-l=.
. ~ h- :.i'* E-? i # $. Sh $R $c E
'ef:?.9'. ..? ~
ation -Diagram 7 -SCCS : (EFEM)' J . Q' :-- Q '%. Q' %
_ -- 3 ~944E673 - " . '
, .y _ .
Modification Diagra'in - SLCS - (FCD)
. . .'. ' 944E674
~ '
, NEDE-22276 Design Analysis and FSAR Inputs for Modified ATWS
- .. p. Performance and Standby 1,.iquid Control system i Please note that.the. four modification diagrams have not t sen design L. verified. This status js, indicated by their revision level "'A". Design and verified documents will be transmitted
- verification
.f...'within two weeks. has been These ordered;Ie will identified by Rev. O.
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- Page 2
/. November 8, 1982
- s. .. . .
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The reference letter also requests an assessment of the impact of this change on system reliability. We find that we are unable to make such an assessment'for lack of adequate human factors input. We note in your
letter PY-CEI/ GEN-612 that CEI has retained a human factor consultant.
We would like to suggest they be requested to provide appropriate input information before we complete this task. .
Very truly yours, 1
R. c.. "
R.'C. Mitchell Projectflanager
. Perry Nuclear Power Plant .
R
.,CM:pab/J11089 Attachment .
cc: P. B. Gudikunst w/o att.
J. J. Larsen w/o att.
W. F. Mietti w/o att.
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HE CLEVELAND ELECTRIC ILLUMINATING COMPANY
@;* , f. ,i' M E M O R A N D U.M. SO/7921 O ',,"e*cive this materia To J. J. Larsen Room rROM E. C. Willman DATc July 22, 1983 GE-NEBO PHONC 5238 Room W240 susaccT Manual Initiation of Standby Liquid Conttol System'C41 Please write a FDDR to remove the following cables from FDI WNVB:
C22A-XX-110 C22A-XX-119 C22A-XX-lll C22A-XX-120 C22A-XX-112 -
C22A-XX-121 C22A-XX-ll3 C22A-XX-122 C22A-XX-114 C22A-XX-123 C22A-XX-115 C22A-XX-124
^ ^
C22A-XX-ll6 C22A-XX-125 C22A-XX-ll7 C22A-XX-126 C22A-XX-118 C22A-XX-142
~.This will remove features not required for our manual initiation concept for the Standby Liquid Control System C41.
n O~
D. R. Gr6en Senior Project Engineer
'ECW/iw-
-cc: J. H. Bellack E. C. Christiansen - TW3 R. E. Coleman - R230 P. A. Nicbols - W250 F. R. Stead -> S245 Electrical File PO/DC - R290 ,
S 5
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[ .* m 9%
.. ] THE-' CLEVELAND ELECTRIC ILLUMINATING '
COMPANY
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MEMDRANDUM 0 sestnu.st
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,fo D. R.' Green nooM W225 rnow E. C. Willman cArg September 28, 1983 PHONg 5238 naos W240 musaccT Status of Standby Liquid Control System C41 Design Change I
'At the present. time GI has made very little progress on-removing the automatic initiation of-the Standby Liquid Control System (SLCS) C-41. They have still not '
' issued the modification kit to the elementary diagrams that we' requested.in September,1982, except cnt a preliminary basis. We have repeatedly asked that a
.. design reviewed mod kit be issued. A promise was made by GE that this would be done.at the last quarterly meeting on July _18 and 19, 1983. Repeated requests
.for issuance of the mod kit have brought no results.
On July 22, 1983, we requested GE to- prepare a FDDR to remove the cables that were
' involved -with automatic initiation (see attached) . This effort in conjunction with GAI's incorporation of the mod kit would complete the design change. As of this date no significant effort has taken place on this FDDR and GE-San Jose has informed me:it is on hold.
"GE:is now requesting approval o'f a work authorization of $80,000 to complete the
. design change.and bring all affected documents up-to-da.ra. I discussed this some time ago with Bob Mitchell and at that time he estimated this to be a S2,000 to S4,000 effort. Also, this was just for revising drawings to include the mod kit.
j Nolindic'ation was given that extra effort would be required to have a. complete design change.
LThe work authorization includes the following:
1.--Issue Revised Redundant Re-activity Control ~ System -(RRCS) Design Specification
' and Data Sheet. *
- 2. Issue revised-balance of plant information document,
~3. -Update Operations and Maintenance Manual.
- 4. : Revise Manual Action Procedure.
S '. - Revise RRCS Elementaries and Functional Conttol Diagrams.
-6. Revise SLCS'elementaries and Functional and Control Diagrams. .
L -
.7. . Prepare New RPCS Panel Schematic and Connection Diagram.
l
- 8. ' Change - the -Programmed Read Only Memory (PROM) in the RRCS Panel.
[.
.' 9 . - Prepare FDI's to . implement cable and PROM changes in the cintrol room, t
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~2 o D. R. Grcsn Ssptsmbar 28, 1983 1
In addition, I pointed out to GE that number 8 is sof tware and there must also be a h'ardware change. They agreed but had no cost estimate for the.' hardware. This item (8) also accounts for approximately one-third of the entire effort. GE is estimating 17. Weeks to complete the work authorization.
From the preceding discussion you can see that we have made very little progress since July. GE has not met their promises and virtually ignored our requests. We need to determine if proceeding in the manual direction is still correct. If so, -
we will have to approve the work authorization. Man'agement attention will be required to ensure GE's prompt completion of the effort.
ECN/iw O
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GENERAL ELECTRIC Ecc- ( M-
.w NUCLEAR POWER SYSTEMS DMSION GENERA:. ELECTRIC CovhANY
- 175 CURTNER AVENUE
- SAN JOSE, CAUTORNIA 95125 MC 399, (408) 925-2755 December 29, 1983 Responds to: N/A PY-GEN /GAI-2002 = INFORMATION Mr. P. B. Gudikunst Gilbert Associates, Inc.
P.O. Box 1498 EECEIVED Reading, PA 19603 JAH 5 1954
Dear Mr. Gudikunst:
SUBJECT:
ATWS MANUAL SLC INITIATION Attached is an advanced . issued copy of ECN NJ50426 sh: wing changes to Standby Liquid Control elementary,8.28E23_4CA. The ECN essentially returns the elementary to Rev'i,sion 2 except for the trip of the pumps on low t.enk level and the relocation of pump and valve control from the containment to a local MCC.
Very truly yours, Ke wa0&- -
R. C. Mitchell Project Manager Perry Nuclear Power Plant /
FROM: DOC. CONTROL DATE:/dg./'
RCH:rm/A12299-
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COPIES TO:'
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W. F. Miotti Q6 -fgez- .k_e P. A. Nichols, w/att.
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