ML20236R008
| ML20236R008 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 11/09/1987 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20236R009 | List: |
| References | |
| NUDOCS 8711230020 | |
| Download: ML20236R008 (10) | |
Text
{{#Wiki_filter:. ancy'o . UNITED STATES 8 NUCLEAR REGULATORY COMMISSION 4 n 3' .h' WASHINGTON, D C. 20555 ' N*...*/ PUBLIC SERVICE ELECTRIC & GAS COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-354 30PE CREEK GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.11 License No. NPF-57 1. The Nuclear Regulatory Comission (the Comission or the NRC) has found ' that: A. The application.for amendment filed by the Public Service Electric-8 Gas Company (PSE&G) dated July 14, 1987, complies with the standards and requirements of the Atomic E.nergy Act of 1954. 'as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will. operate in conformity with the application, the-provisions of the Act, and the rules and regulations of the Comission; C. There is reasonable' assurance: (1)that.theactivitiesauthorizedby k this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations' set forth in 10 CFR Chapter I; j l D. The issuance of this amendment will not be inimical to the common d defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied. 1 2. Accordingly, the license is amended by changes to the Technical.Specifica-tions as indicated in the attachment to this license amendment, and paragraph 2.C' (2) of Facility Operating License No. NPF-57 is hereby amended to read as follows: (2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No.11, and the Environmental Protection Plan contained in i Appendix B, are.hereby incorporated in the license. PSE8G shall operate the facility in accordance with the Technical Specifications and the-Environmental Protection Plan. 8711230020 e71109 PDR ADOCK~05000354 P
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. 1 3. This license emendment is effective as of its date of issuance and shall be implemented within 60 days of issuance. FOR THE NUCLEAR REGULATORY COMMISSION /s/ ' Walter R. Butler, Director Project Directorate I-2 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications ( Date of Issuance: November 9, 1987 i j I l l L3)h.DRPI/II YDRPI/II PM: OGC <#p b. D:PDI-2:DRPI/II My 3 Men GR venbark:ca WButler jb/f///87 g/D/87 j /g /87 }l/3/87 l w
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3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance. FOR THE NUCLEAR REGULATORY COMMISSION Walter R. Butler. Director l Project Directorate I-2 l l Division of Reactor Projects I/II 1 Office of Nuclear Reactor Regulation ] l
Attachment:
Changes to the Technical Specifications I i Date of Issuance: November 9, 1987 l l i L
ATTACHMENT TO LICENSE AMENDMENT NO.11 FACILITY OPERATING LICENSE NO. NPF-57 DOCKET NO. 50-354 Replace the following pages of the Appendix "A" Technical Specifications with' I the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. Overleaf pages provided to maintain document completeness.* Remove Insert 3/4 1-19* 3/4 1-19* 3/4 1-20 3/4 1-20 3/4 1-21 3/4 1-21 q B 3/4 1-3* B 3/4 1-3* B 3/4 1-4 8 3/4 1-4 B 3/4 1-5 ] a l .i l 1
) R.EACTIVITY CONTROL SYSTEMS ] 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM f LIMITING CONDITION FOR OPERATION 1 3.1.5 The standby liquid control' system consists of two redundant subsystems j and shall be OPERABLE. APPLICABILIH: OPERATIONAL CONDITIONS 1, 2, and 5* l ACTION: a. In OPERATIONAL CONDITION 1 or 2: 1. With one system subsystem inoperable, restore the subsystem to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours. 2. With both system subsystems inoperable, iestore at least one subsystem to OPERABLE status within 8 hours or be in at least HOT SHUTDOWN within the next 12 hours. b. In OPERATIONAL CONDITION 5*: 1. With one system subsystem inoperable, restore subsystem to' ) OPERABLE status within 30 days or. insert all insertable control rods within the next hour.- 2. With both standby liquid control system subsystems inoperable, l insert all insertable control rods within one hour. 1 SURVEILLANCE REQUIREMENTS 4.1.5 The standby liquid control system shall be demonstrated OPERABLE-At least once.per 24 hours by verify'ing that; a. 1. The temperature of the sodium pentaborate solution in the storage tank is greater than or equal to 70*F. 2. The available volume of sodium pentaborate solution is within the limits of Figure 3.1.5-1. 3. The heat tracing circuit is OPERABLE by determining the temperature of the pump suction piping to be greater than or equal to 70*F. l "With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2. HOPE CREEK 3/4 1-19
REACTIVITY CONTROL SYSTEMS -{ SURVEILLANCE REQUIREMENTS (Continued) J b. At least once'per 31 days by: 1. Verifying the continuity of the explosive charge. 2. Determining that the available weight of sodium pentaborate is greater than or equal to 5,776 lbs and the concentration of l boron in solution is within the limits of Figure 3.1.5-1 by chemical analysis.* 3. Verifying that each valve (manual, power operated or automatic) in the flow path that is not ~ 1ocked, sealed, or otherwise secured in position, is in its correct position, c. ' Demonstrating that, when tested pursuant to Specification 4.0.5, the minimum flow requirement of 41.2 gpm, per pump,'at a pressure of greater than or equal to 1255 psig is met. d. At least once per 18 months during shutdown by: 1. Initiating one of the standby liquid control. system subsystem, including an explosive valve, and verifying that a flow path from the pumps to the reactor pressure vessel.is available.by pumping demineralized water into the reactor vessel and. verifying that the relief valve does not actuate.- The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired'or from another batch which has been certified by having one of that batch successfully fired. Both injection subsystems shall be tested in 36 months. 2.
- Demonstrating that all heat traced piping between the storage tank and the injection' pumps is unblocked and then draining and flushing the piping with demineralized water.
3. Demonstrating that the storage tank heaters.are OPERABLE by verifying the expected temperature rise of the sodium penta-borate solution in the storage tank after the heaters are energized.
- This test shall also be performed anytime water or boron is added to the solution or when the solution temperature' drops below 70*F.
- This test shall also be performed whenever both heat tracing circuits have been found to be inoperable and may be performed by any series 'of sequential,.
overlapping or total flow path steps such that the entire' flow path'is included. HOPE CREEK 3/4 1-20 Amendment No.11 l
i i EE d 8" 1 8 3 i m ad 1 ga -* l- -l l 1 1 i !ls-* g l I-m W s .lg d EW isk-se -I. gW"w v dB a ex gjgM g S !s c- !d m g"E 63 7 g 8h E d e g g W y = ( ~ E e n- - :l g -B. a l @Wa EE l l-+ W5 _3 5 1E B a l l g I I I l E S 3 2 8 IN3383d 1HDI34h011W1N33NG3 31VWO901N3d HnIDOS HOPE CREEX 3/4 1-21 Amendment No.ll
REACTIVITY CONTROL SYSTEMS BASES CONTROL RODS (Continued) Control rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident in the FSAR. The overtravel position feature provides the only positive means of determining that a rod is properly coupled and therefore this check must be performed prior to achieving criticality af ter completing CORE ALTERATIONS that could have affected the control rod coupling integrity. The subsequent check is performed as a backup to the initial demon-stration. In order to ensure that the control rod patterns can be followed and there-fore that other parameters are within their limits, the control rod position indicatien system must be OPERABLE. The control rod housing support restricts the outward movement of a control rod to less than 6 inches in the event of a housing failure. The amount of rod reactivity which could be added by this small amount of rod withdrawal is l 1ess than a normal withdrawal increment and will not contribute to any damage pressure to act as a driving force to rapidly eject a drive housing. to the primary coolant system. The support is not required when there is no The required surveillance intervals are adequat'e to determine that the i rods are OPERABLE and not so frequent as to cause excessive wear on the system components. 3/4.1.4 CONTROL ROD PROGRAM CONTROLS Control rod withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth enough to result in a peak fuel enthalpy greater than 280 cal /gm in the event of a control rod drop accident. The specified sequences are characterized by homogeneous, scattered patterns of. control rod withdrawal *. When THERMAL POWER is greater than 20% of RATED THER'4AL POWER, there is no possible rod w' orth which, if dropped at the design rate of the velocity limiter, could result in a peak enthalpy of 280 cal /gm. Thus requiring the RSCS and/or RWM to be OPERABLE when THERMAL POWER is less than er equal to 20% of RATED THERMAL POWER provides adequate control. The RSCS and RWM provide automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted. The analysis of *he rod drop accident is presented in Section 15.4.9 of the FSAR and the techniques of the analysis are presented in a topical report, Reference 1, and two supplements, References 2 and 3. The RBM is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power operation. Two channels are provided. Tripping one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage. This system backs up the written sequence used by the operator for withdrawal of control rods. HOPE CREEK B 3/4 1-3 i
I 4 i REACTIVITY CONTROL SYSTEMS ~ l l BASES 1 1 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM .The standby liquid control system provides a backup capability for bringing j the reactor from full power to a cold, Xenon-free shutdown, assuming that the i withdrawn control rods remain' fixed in the rated power pattern. To meet this I objective it is necessary to inject a quantity of boron which produces a concen- ) tration of 660 ppm in the reactor core and other piping systems connected to the i reactor vessel..To allow for potential leakage and imperfect mixing, this con-centration is increased by 25%. The generic design basis of the standby liquid control system provides a specified cold shutdown boron concentration in the reactor _ core. The standby liquid control system was typically designed to in-ject the cold shutdown boron concentration in 90 to 120 minutes. The t:me re-quirement was selected to override the reactivity insertion rate due to cool down following the xenon poison peak. The pumping rate of 41.2 gpm meets the l requirement. The minimum storage volume of the solution is established to include the generic shutdown requirement and to allow for the portion below the pump suction nozzle that cannot be inserted. An additional allowance in the standby liquid control storage volume is provided to account for storage tank instrument inac-curacy and drift. Even with the maximum specified instrument inaccuracy and drift, the required quantity of sodium pentaborate solution is always available forinjection. A normal quantity of 4640 gallons of sodium pentaborate solution having i a 14.0 percent concentration is required to meet the shutdown requirements. The temperature requirement for sodium pentaborate solution and the pump suc-tion piping is necessary to ensure the sodium pentaborate remains in solution. With redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to 5 continue for short periods of time with the system inoperable or for longer ~ periods of time with one of the redundant components inoperable. 1 Surveillance requirements are established on a frequency that assures a high reliability of the system. Once the solution is established, boron con-l centration will not vary unless more boron or water is added, thus a check on i the temperature and volume once each 24 hours assures that the solution is available for use. Replacement of the explosive charges in the valves-at regular intervals will assure that these valves will not fail because of deterioration of the charges. l l The ATWS Rule (10 CFR 50.62) requires the addition of a new design require- - ment to the generic standby liquid control system design basis. Changes to flow i HOPE CREEK B 3/4 1-4 Amendment No.11 .________.__._____________-_____.___m___._____________e____-______
REACTIVITY CONTROL SYSTEMS BASES rate, solution concentration or boron equivalent to meet the ATWS Rule must not invalidate the original system design basis. Paragraph (c)(4) of 10 CFR 50.62 states that: "Each boiling water reactor must have a Standby Liquid Control System (SLCS) with a minimum flow capacity and boron control equivalent in control capacity to 86 gallons per minute of 13 weight percent sodium pentaborate solution (natural boron enrichment)." The described minimum system parameters (82.4 gpm, 13.6 percent concentra-tion and natural baron equivalent) will ensure an equivalent injection capability that exceeds the ATWS Rule requirement. The stated minimum allowable pumping rate of 82.4 gallons per minute is met through the simultaneous operation of both pumps. 1 1. C. J. Paone, R. C. Stirn and J. A. Woolley, " Rod Drop Accident Analysis { for Large BWR's", G. E. Topical Report NEDO-10527, March 1972 2. C. J. Paone, R. C. Stirn and R. M. Young, Supplement 1 to NEDO-10527, July j 1972 3. . J. M. Haun, C. J. Paone and R. C. Stirn, Addendum 2, " Exposed Cores", Supplement 2 to NE00-10527, January 1973 HOPE CREEK B 3/4 1-5 Amendment No.ll 1 __- - -}}