ML20207D264
| ML20207D264 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 12/19/1986 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20207D106 | List: |
| References | |
| NUDOCS 8612300362 | |
| Download: ML20207D264 (14) | |
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JAFWPP LIST OF TABLES (CONT'D)
Page Toble Title Safety Related Shock Suppressors (Snubbers) 156b 3.6-1 157 4.6-1 Comparison of the James A. FitzPatrick Nuclear Power Plant Inservice Inspection Program to ASME Inservice Inspection Code Requirements 198 3.7-1 (DELETED)
Mininum Test and Calibration Frequency for 210 4.7-1 Containment Monitoring Systems 211 4.7-2 Exception to Type C Tests 261 6.10-1 Component Cyclic or Transient Limits Protection Factors for Respirators 262 6.11-1 O
Amendment No. pI p ' p '
vi 8612300362 DR 861219 ADOCK 05000333 PDR
- o JAFNPP 3.7 (cont'd) ,
4.7 (cont'd)
- c. Secondary containment capability to main-tain a 1/4 in, of water vacuum under calm wind conditions with a filter train flow .
rate of not more than 6,000 cfm, shall be demonstrated at each refueling outage prior to refueling.
Primary Containment Isolation Valves D. Primary Containment Isolation Valves D.
- 1. During reactor power operating conditions, all 1. The primary containment isolation valves surveillance shall be performed as follows:
I isolation valves and all instrument line flow check valves shall be operable, except as specified in 3.7.D.2.
- a. At least once per operating cycle, the operable isolation valves that are power operated and automatically initiated shall be tested for simulated automatic initiation and closure times.
- b. At least once per operating cycle, the instrument line excess flow check valves shall be tested for proper operation.
- c. At least once per quarter:
(1.) All normally open power-operated isola-tion valves (except for the main steam line and RBCLCWS power operated isola-tion valves) shall be fully closed and reopened.
Amendment No.
185 L
- O JAFEPP 3.7 (cont'd) ,
4.7 (cont'd)
(2.) With the reactor at reduced power -1 level trip, main steam isolation valves and verify closure time. .
d . At least twice per week, the malii steam line power-operated isolation valves shall be exercised by partial closure and subsequent reopening.
any isolation valve becomes e. The RBCLCWS isolation valves shall be fully
- 2. In the event inoperable, reactor power operation may continue, closed and reopened anytime the reactor is I provided at least one valve in each line having in the cold condition exceeding 48 hrs, if the valves have not been fully closed and an inoperable valve is in the mode corresponding to the isolated condition (except for RBCLCWS reopened during the preceeding 92 days.
isolation valves for which no valves are required to be in the isolated condition.) 2. The position of the valves placed in the isolated condition to satisfy 3.7.D.2 shall be recorded If Specification 3.7.D.1 and 3.7.D.2 cannot be daily, 3.
met, an orderly shutdown shall be initiated and the reactor shall be in the cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Amendment No.
186
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JAFNPP 3.7 BASES (cont'd) .
of the containment. Closure of one of the valves in each line would be sufficient to maintain the .
integrity of the Pressure Suppression System.
Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a loss-of-coolant accident.
Nine remote manually operated isolation valves have been added to the Reactor Closed Loop Cooling Water System (RBCLCWS) in order to comply with 10 CFR 50 Appendix A CDC 57; These valves are air operated (with solenoid pilot valves), normally ope'n, and are designed to fail 'open' on loss of electrical power or ** a s is" upon loss of instrument air. Each AOV is provided with a accumulator tank to allow Seismic Class I operation of the valves upon loss of instrument air up to 2 full valve cycles. The affected portion of the RBCLCWS is an essential system and performs a post-accident safety function. The fail-open design permits continued operation of the system to supply water to thecoolers recirculation during pump-motor coolers and drywell normal operation and as necessary under accident conditions. The RBCLCWS lines are part of a closed system, as defined in GDC 57, and do not communicate with either the primary system or the containment. Since the piping itself comprises an isolation boundary, the single isolation valve need not be closed if it should become inoperable. If there is a postulated accident, the and indications of leakage from RBCLCWS, operator will selectively close the A0V's affected to provide containment isolation.
A list of containment isolation valves, including a brief description of each valve, is included in the updated FitzPatrick FSAR.
Amendment No. 192
/ JAFNPP 4.7 BASES (cont'd) operability results in a more reliable system.
The main steam line isolation valves are func-
- tionally tested on a more frequent interval to establish a high degree of reliability.
The primary containment is penetrated by several small diameter instrument lines connected to the reactor coolant system. Each instrument line contains a 0.25 in. restricting orifice inside the primary containment and an excess flow check valve outside the primary containment.
valves are excluded from the The RBCLCWS quarterly surveillance requirements because closure of these valves will eliminate the coolant flow to the drywell air and recirculation pump-motor coolers.
Without cooling water, the air and equipment temperature will drywell increase and may cause damage to the equipment Therefore, during normal plant operations.
testing of these valves would be conducted only in the cold condition.
A list of containment isolation valves, including a brief description of each valve, is included in the updated FitzPatrick FSAR.
Amendment No. 197
PAGES 198 THROUGH 209 HAVE BEEN DELETED o
198 Amendment No.
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ATTACHMENT II TO JPN-86-60 NEW YORK POWER AUTHORITY James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 DPR-59 Safety Evaluation for Technical Snecification Chanaes Related to the Deletion of Table 3.7-1.
" Process Pinina Penetratina Primary Containment" (JPTS-86-016)
I. Descrintion of Chanaes The' changes proposed by this amendment application delete Table 3.7-1, " Process Piping Penetrating Primary containment" along with the associated notes and any references to this table from the FitzPatrick Technical Specifications. A reference to a similar table in the updated FitzPatrick Final Safety Analysis Report (FSAR) replaces the deleted table. Exemption from certain Limiting Condition for Operations (LCOs) and surveillance requirements have been added for Reactor Building Closed Loop Cooling Water System containment isolation valves.
Specifically, this amendment proposes to make the following changes to the FitzPatrick Technical Specifications:
- 1. Table 3.7-1 and its associated notes include a total of twelve pages extending from page 198 to i
. 209, inclusive. All twelve pages will be replaced with a single page bearing the words "PAGES 196 THROUGH 209-HAVE BEEN DELETED." A note at the bottom of the page will further alert the reader that the "Next page is 210."
These twelve pages previously contained the following:
Pages 198 through 204 - Table 3.7-1,
" Process Piping Penetrating Primary Containment."
Page 205 - was intentionally blank.
Pages 206 and 207 " Notes for Table 3.7-1, Isolation Signal Codes," listed and described the isolation signal codes.
Pages 208 and 209 " Notes for Table 3.7-1" listed seventeen miscellaneous notes associated with Table 3.7-1.
- 2. The List of Tables (p. vi) has been revised to show that Table 3.7-1 has been deleted.
- 3. References to Table 3.7-1 have been deleted from three sections and RBCLCWS isolation valves have been excluded from certain limiting conditions for operation and quarterly surveillance requirements:
- a. In Section 3.7.D.1 on page 185, the phrase " listed in Table 3.7-1" has been deleted from the first sentence in this section. The first sentence has been revised to read: "During reactor power operating conditions, all isolation valves and all instrument line flow check valves shall be operable, except as specified in 3.7.D.2."
- b. In Section 3.7.D.2 on page 186, the phrase "specified in Table 3.7-1" has been deleted and a phrase has been added to exclude the RBCLCWS isolation
. valves from this Limiting Condition for operation. The section has been revised to read: HIg thg gyggt ggy isolation valve becomes inoperable, reactor power operation may continue, I
provided at least one valve in each line having an inoperable valve is in i the mode corresponding to the isolated condition (except for RBCLCWS isolation valves for which no valves are required to be in the isolated condition.)"
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- c. In Section 4.7.D.2 on page 186, the rewritten to delete a reference to Table 3.7-1 and adds a reference to Section 3.7.D.2 to define which valves require this surveillance. The addition of this reference will exclude RBCLCWS isolation valves from possible daily recording requirements. The section has been revised to read: "The position of the valves placed in-the isolated condition to satisfy 3.7.D.2 shall be recorded daily."
- 4. In Section 4.7.D.l.a on page 186, a new surveillance requirement for the RBCLCWS isolation valves has been added. This new section reads: "The Reactor Building Clcaed Loop Cooling Water System (RBCLCWS) isolation valves shall be fully closed and reopened anytime the reactor is in the cold condition exceeding 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> if the valves hava not been fully closed and reopened during the proceeding 92 days." The wording of this surveillance requirement is consistent with a similar surveillance requirement for recirculation pump discharge and bypass valves in Technical Specification Section 4.A.5.5, page ll5a.
- 5. A reference to the table of containment isolation valves in the updated FitzPatrick FSAR has been added to the bases associated with sections 3.7 l (p. 192) and 4.7 (p. 197). The sentence "A list of containment isolation valves, including a
, brief description of each valve, is included in
! the updated FitzPatrick FSAR." has been added on j . both pages 192 and 197.
- 6. In Section 4.7.D.l.c(1) on page 185, the phrase "and RBCLCWS" has been added. The section has
- been revised to read "All normally open i
power-operated isolation valves (except for the main steam line power operated isolation valves and RBCLCWS power operated isolation valves) shall be fully closed and reopened."
- 7. In bases Section 3.7 on page 192, a paragraph has i
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been added which describes the RBCLCWS containment isolation valves and their design bases..
- 8. In bases Section 4.7 on page 197, a paragraph has
- been added which describes why the RBCLCWS containment isolation valves are tested during cold shutdown.
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_II. Purnose of the Proposed Chances l
- Table of Containment Isolation Valves This proposed change will reduce the amount of administrative resources required by both the NRC and the Authority to maintain an accurate and up-to-date table of
- containment isolation valves. The elimination of lists of i components has been identified as a generic improvement by both industry and NRC programs.
I Rather than preparing and submitting a Technical Specification amendment request for each Table 3.7-1
- alteration, the Authority will maintain the table of -
i containment isolation valves in the FitzPatrick FSAR and j issue it each year as part of the updated FSAR. This will
- assure that the table is updated at least once a year f without the expenditure of resources needed to obtain an
) operating license amendment.
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! FSAR Table 7.3-1, also entitled " Process Piping i Penetrating Primary Containment," was included in the FSAR i
update issued in July 1986 (Reference 1.) This new table arranges entries by containment penetration, better j
describes the penetrations' function and clearly identifies I
2 the associated containment isolation valve (s). Isolation
. signals, valve closure time, and normal status are included
- on the FSAR version of this table. The new table also has j note section similar to the one in the Technical Specifications, but the FSAR notes have been improved and clarified.
4 Several other improvements have been made to make the i
table more useful and easier to use. A different format
! has been used to eliminate the need to photographically reduce the table. Unnecessary columns in the table have l
been deleted.
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The FSAR table is also more complete than its Technical Specification counterpart. Containment isolation
- valves not appearing in Table 3.7-1 in the Technical Specifications have been added to the FSAR table. These additions correct errors or reflect plant modifications.
The Authority endorsed technical specification reform activites in November of 1985 (Reference 2). Over one year ago, both the NRC's Technical Specifications Improvement Project Report (Reference 3) and the Atomic Industrial Forum's report (Reference 4) endorsed that idea of using the FSAR as an appropriate place for this type of information. (At least one other power reactor licensee has proposed an operating license amendment to delete the table of containment isolation valves from their technical specifications.)
RBCLCWS Isolation Valves Nine remote manual containment isolation valves were added to the RBCLCWS to comply with General Design Criteria 57 of Appendix A to 10 CFR 50 and to conform with the plant as described in the Final Safety Analysis Report (FSAR).
The installation of these containment isolation valves were discussed in References 5 and 6.
These are normally open, air-operated valves (AoVs) with solenoid pilot valves. They are designed to fail open on loss of electrical power or as-is on loss of instrument air. Each air-operated valve has a Seismic Class I accumulator to assure that they will remain operable even after a loss of instrument air. The accumulators have been sized for two full valve cycles.
Portions of the RBCLCWS are classified as essential !
because they perform a safety functions in the event of an
- accident - supplying cooling water to equipment (recirculation pump motor and drywell atmosphere coolers) inside the drywell. A continous supply of water is assured during both normal and accident conditions by the fail open design of these new containmenu isolation valves.
The RBCLCWS lines which penetrate primary containment are part of a closed system as defined by General Design Criteria 57 of Appendix A to 10 CFR 50. They do not communicate with either the primary system or the drywell atmosphere. Since the piping itself is an isolation l
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boundary, the single isolation valve need not be closed if it becomes inoperable. These valves normally remain open through all accident scanarios. They are closed only if RBCLCWS leakage is evident.
The RBCLCWS isolation valves are excluded from quarterly surveillance requirements during power operation because closing these valves cuts-off the cooling water supply to essential equipment inside the drywell. Without cooling water, the drywell coolers can not cool the drywell atmosphere. The resulting increase in equipment and atmosphere temperatures could overheat equipment inside the drywell.
It is not economically practical to shutdown the plant to conduct quarterly surveillance tests of the RBCLCWS isolation valves. Tests conducted during plant shutdown, but not more frequent than once every three months, will permit these tests to be schedaled during outage conditions when the availability of the RBCLCWS is not required while maintaining a high degree of confidence in the operability of the valves.
A technical specification amendment to add the changes associated with these RBCLCWS valves was previously
, submitted by References 7 and 8. However, the Authority subsequently withdrew this application (Reference 9 and 10.)
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III. Imoact of the Procomed Chances
! Table of containment Isolation valves The deletion of this table is purely administrative in
. nature and will not' degrade safety. This amendment does not alter any operability or surveillance requirements currently in the FitzPatrick Technical Specifications.
10 CFR 50 contains adequate requirements applicable to containment isolation valves to assure the safe operation of FitzPatrick whether or not the specific list of valves is included in the Technical Specifications.
i Both 10 CFR 50.59, " Changes, test and experiments" and 10 CFR 50.71(a) , " Maintenance of records, making of reports" already contain provisions which require the
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Authority to inform the NRC of changes to the plant. 10 CFR 50.59 permits the Authority to:
i "(i) make changes in the facility as described in the safety analysis report, (ii) make changes in the procedures as described in the safety analysis report,...without prior Commission approval unless the proposed change, test or experiment involves a change in the technical i
specifications incorporated in the license or an unreviewed safety question." .
10 CFR 50.59 requires that the Authority annually submit to the NRC "a report containing a brief description of such changes, test or experiments, including a summary of the safety evaluation of each." This part also requires that the Authority complete a safety evaluation for each ,
change to assure that the change does not involve an i "unreviewed safety question." 10 CFR 50.59 (a) (2) defines three criterion to be applied to determine if a change
" involves and unreviewed safety question."
10 CFR 50.71(a) requires that the Authority to revise the FitzPatrick FSAR each year "to assure that the i information included in the FSAR contains the latest material developed."
10 CFR 50.59 and 10 CFR 50.71 provide adequate assurance that changes to the plant that result in changes to this table will be evaluated by the Authority and sufficient opportunity is afforded for NRC review.
Table 3.7-2 " Exception to Type C Tests," (pages 211, 212 and 213) lists for Type C tests exceptions for certain containment penetrations and containment isolation valves.
- This table is not deleted as part of this amendment application and will be retained as part of the Technical Specifications.
RBCLCWS Isolation Valves l The incorporation of these changes will reflect the addition of these new isolation valves in the FitzPatrick
. Technical Specifications.
The proposed surveillance test interval provides
adequate assurance of the valve's ability to operate because: the ability to mitigate the effects of- nn accident are not affected by an inoperable valve; the' lines do not communicate with either the containment atmosphere or primary coolant; the valve is not required to operate (i.
- e. close) in the event of an accident; and, less frequent testing will reduce the possibility of drywell equipment failure or degradation through overheating.
IV. Evaluation of Sionificant 7Jazards Considerations Operation of the FitzPetrick plant in accordance with the proposed amendment does not involve a significant hazards consideration, as defined by 10 CFR 50.92, since it does not:
- a. involve a significant increase in the probability or consequences of an accident previously evaluated. The relocation of this information from the Technical Specifications to the FSAR is purely an administrative change. It will have no effect on how the plant is maintained or operated nor does it alter the plant's design. Federal regulations 10 CFR 50.59 and 10 CFR 50.71 already contain provisions that require the Authority to complete a safety evaluation of any changes to the plant, report these changes annually, and to update the FSAR.
The replacement of manual RBCLCWS valves with remotely operated valves does not significantly increase the probability or consequences of an accident because their replacement has improved overall plant safety by increasing containment isolation dependability. The fail open/as-is design of these new containment isolation valves assures a continous supply of cooling water during both normal and accident conditions. Surveillance testing cannnt cause an accident because testing will be periormed during plant shutdowns when the RBCLCWS is not required to cool the drywell atmosphere. Surveillance tests for the new remote manual valves will be conducted more frequently than the tests previously conducted on the manual valves. This will further assure increased valve operablility.
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- b. create the possibility of a new or different kind of accident from any accident previously evaluated. The relocation of the table of containment isolation valves does not involve either a modification to the plant or a change in the procedures used for plant operation.
The change from manual to remote manual RBCLCWS containment isolation valves does not create the possibility of a new or different accident because a continous supply of water is assured during both normal and accident conditions by the fail open/as-is design of these new containment isolation valves. RBCLCWS containment isolation valve operability tests cannot create the possibility of a new or different accident because testing will be performed during reactor shutdowns. This will preclude the possibility that their failure to reopen following a test conducted during power operation could precipitate drywell equipment degradation due to loss of drywell atmosphere cooling and high drywell temperatures.
- c. involve a significant reduction in a margin of safety. A similar table has been provided in the updated FitzPatrick FSAR. The FSAR will be revised annually, as a minimum, under the provisions of 10 CFR 50.71(e). This amendment does not alter any operability or surveillance requirments currently in the FitzPatrick Technical Specifications.
The addition of RBCLCWS containment isolation valves has increased the margin of safety by increasing the extent to which the FitzPatrick plant complies with General Design criteria 57 of Appendix A to 10 CFR 50. Compliance with this criteria improves containment isolation dependability. The ability of the plant to mitigate the effects of an accident are not affected by the failure of these valves.
V. Imolenentation of the Chances 1
Implementation of these changes, as proposed, will not impact the ALARA or Fire Protection Programs at FitzPatrick, nor will the changes impact the environment.
r VI. Conclusion i I
The incorporation of these changes:
- a. will not change the probability nor the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety Analysis Report;
- b. will not increase the possibility of an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report;
- c. will not reduce the margin of safety as defined in the basis for any technical specification;
- d. does not constitute an unreviewed safety question as defined in 10 CFR 50.59; and
- e. involves no significant hazards consideration, as 4
defined in 10 CFR 50.92.
VII. References and Notes
- 1. NYPA letter, J. C. Brons to H..R. Denton dated July 17, 1986 (JPN-86-032) regarding 1986 annual FSAR update.
- 2. NYPA letter, J. C. Brons to H. R. Denton, dated November 15, 1985 (JPN-85-083, IPN-85-060) regarding NYPA endorsement of technical specification reform activities.
l 3. NRC Technical Specifications Improvement Project Final Report, " Recommendations for Improving Technical i
- - Specifications," dated September 30, 1985.
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- 4. Atomic Industrial Forum, " Technical Specifications Improvement," Subcommittee on Technical Specification Improvements, dated October 1, 1985.
- 5. PASNY letter, J. P. Bayne to D. B. Vassallo, dated February
- 22, 1982 (JPN-82-021).
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- 6. PASNY letter, J. P. Bayne to T. A. Ippolito, dated January l 7, 1982 (JPN-82-005) l l _lo_
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- 7. NYPA letter, J. P. Bayne to D. B. Vassallo, dated November 10, 1983 (JPN-83-094) submits proposed changes to the Technical Specifications related to Reactor Building Closed Loop Cooling Water System Containment Isolation Valves.
- 8. NYPA letter, J. C. Brons to D. B. Vassallo, dated July 10, 1985 (JPN-85-056) included bases for the Technical Specification change transmitted November 10, 1983.
- 9. NYPA letter, J. C. Brons to D. B. Vassallo, dated November 1, 1985 (JPN-85-079) regarding the withdrawl of proposed change to the Technical Specifications.
- 10. NRC letter, D. B. Vassallo to J. C. Brons, dated November 20, 1985. Include Notice of Withdrawl for Authority November 10, 1983 application for amendment.