ML051990510
| ML051990510 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 06/29/2005 |
| From: | Nuclear Management Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML051990510 (432) | |
Text
{{#Wiki_filter:IMPROVED TECHNICAL SPECIFICATIONS MONTICELLO NUCLEAR GENERATING PLANT VOLUME 11 ITS Section 3.6, Containment Systems Committed to
Attachment 1, Volume 11, Rev. 0, Page 1 of 431 ATTACHMENT 1 VOLUME 11 MONTICELLO IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS SECTION 3.6 CONTAINMENT SYSTEMS Revision 0 Attachment 1, Volume 11, Rev. 0, Page 1 of 431
Attachment 1, Volume I1, Rev. 0, Page 2 of 431 LIST OF ATTACHMENTS
- 1. ITS 3.6.1.1
- 2. ITS 3.6.1.2
- 3. ITS 3.6.1.3
- 4. ITS 3.6.1.4
- 5. ITS 3.6.1.5
- 6. ITS 3.6.1.6
- 7. ITS 3.6.1.7
- 8. ITS 3.6.1.8
- 9. ITS 3.6.2.1
- 10. ITS 3.6.2.2
- 11. ITS 3.6.2.3
- 12. ITS 3.6.3.1
- 13. ITS 3.6.4.1
- 14. ITS 3.6.4.2
- 15. ITS 3.6.4.3
- 16. Improved Standard Technical Specifications (ISTS) not adopted in the Monticello ITS Attachment 1, Volume 11, Rev. 0, Page 2 of 431
, Volume 11, Rev. 0, Page 3 of 431 ATTACHMENT 1 ITS 3.6.1.1, Primary Containment , Volume 11, Rev. 0, Page 3 of 431
Attachment 1, Volume 11, Rev. 0, Page 4 of 431 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 11, Rev. 0, Page 4 of 431
ITS 3.6.1.1 ITS ITS 3.0L UMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS
- 2. Primazy Containment 2. Primary Contalnme enMeg LCO 3.6.1.1 Primary Cortainme nt i a. Perform required visual examinatlons and (0 L !2iJ. shell be a mes SR 3.6.1.1.1 leakage rate testing except for primary w en a raeaor i critica or when the containment air lock testing. In accordance with 3 Applicability reocorwatertemperature hs bov 2 M.1 the Primary Containment Leakage Rate Testing (
CD and fuel Is In the rea exce Program. 3
- l SeelTS3.tO.1 -iedinl3
< t)rfn onanetg ynorqi - -{SeetITS 3.6.1.3} < (0 nlinmentjpoer isno t _ (20 E' 3 a oerIwsnot to xceced CDVt) (0D( l3) Pdimory Contalnment Integrity Is; not required when perforrning reactor vessel hydrostatic u_ See ITS 3.10.1 XC lenkooo tests vvith the reador ndI critcal. X oACTI ACTION A A (4) I requirements of 3.7A.Za.(l) cannot be mt restore Prlmary Cont~ ntalmetC {330A2) o witH~n one hoyf~orb In at beast Hot lan ow wnt the next 12 hours and Cold Di ID ACTION B Shutdown within the following 24 hours. to toDC (0 0 a a 3.714.7 158 01/28105 Amendment No. 30r55, ,5 107,432,141 Page 1 of 6
( ( C ITS 3.6.1.1 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS
> I
- 0) IS a) 0 0 3
0 0 0, 0 5
-A
- U 0) 0 CD Co 0 -9' to a Co 0) 0 0 (A~
3.7/4.7 159 02/04/03 Amendment No. 52, 55-,72, 06, 132 Page 2 of 6
( ( ITS 3.6.1.1 ITS 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS
.1 _____________________
.
- c. When Primary Containment Integrity Is c. (1) Perform required primary containment air .
required, the primary containment air lock shall lock leakage rate testing in accordance be operable with both doors closed except with the Containment Leakage Rate 0) 0) CD I when the air lock Is being used, then at least one air lock door shall be dosed. With the primary containment air lock _ Testing Program.(m") (2) Once per 24 months, verify that only one
+See ITS 3.6.1.2}
door in the primary containment air lock 0 inoperable. maintain at least one air lock door can be __ onpened __ w__ _ _.at _a time.
..... _.
closed and restore the air lock to Operable status within 24 hours or be in at least Hot CD Shutdown within the next 12 hours and in Cold d. 0 Shutdown within Ihe following 24 hours. SR 3.6.1.1.1 CD
-
0i CD 0) Eo CD CD 0 0 INt -0A
- An Inoperable air lock door does not Invalidate the previous successful performance of the overall air lock leakage test. See ITS 3.6.1.21 Results than be evaluated against acceptance criteria .
- appricabla to SR 4.7.A.2.a.
3.7/4.7 NEXT PAGE IS 163 160 02/04/03 Amendment No. 66, 69,05,132 Page 3 of 6
( ( ITS 3.6.1.1 0 0 0 D 3 20 0 2 0 3 0 -A -o 0
;U 0
0 0 Co
- 1. Instrument Channel - An Instrument channel means an arrangement of a sensor and auxiliary equipment required to Co generate and transmit to a trip system, a single trip signal related to the plant parameter monitored by that Instrument channel.
- 2. -ipSe A trip system means an arrangement of Instnument channel trip signals and auxiliafy equipment required to 0 Initiate a protection action. A trip system may require one or more Instrument channel trip signals related to one or more See ITS 1.0I -4 plant parameters to Initiate trip system action. initlation of the protedive function may require tripping of a single trip system (e.g., HPCI system Isolation, off-gas system Isolation, reactor building Isolation and standby gas treatment InitiatIon, and rod block), or the coincident tripping of two trip systems (e.g., Initiation of scram, reactor Isolation, and primary containment Isolation).
- 3. Protective Acton - An action Initiated by the protection system when a limit Is exceeded. A protective action can be at channel or system level.
3 1/23/84 Amendment No. 21 Page 4 of 6
( C ITS 3.6.1.1 ITS 3.0 LIMITING CONDIONS FOR OPERATION 4.0 SURVEILLANCE REOUIREMENTS 3.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS tu Aimlicabirdtv. ApolicablifIv: S 0 2) Applies to the operating status of the primary end secondary Applies to the primary and secondary containment Integrity. 0 containment systems. 3 Oblective Obiecliva: 0
- To essure the integrity of the primary and secondary To verify the Integrity of the primary and secondary See ITS 3.6.2.1 3 containment systems. containment. aendITS 3.6.2.2J 2 rD 0
Soecification: Specificlaton: A. Primery Containment 3 A. Primary Containment D
;U 0 0 1. Suppression Pool Volume and Temperature 1. Suppression Pool Volume and Temperature -A
{
+
.I Whe irradiated fuel is in the reactor vessel and I a. The suppression chamber water temperature 2D See ITS 3.6.2.1 and ITS 3.6.2.2 eIiher e or wtrrmerat r wo is ei done which hs the ntii shall be checked once per day.
rD llo drwo Is ein ewhchhs shaRe 0) 0) to { See ITS 3.5.2 and ITS 3.6.2.1 r 5 I __ a. _ 1s _ _ _ _ _ _ _ ln _ __ _ e _ _ c _ Water temperature during normal operating shall be s 90* F. _ _ __l _ See ITS 3.5.21 b. Whenever there Is Indication of relief valve operation which adds heat to the suppression pool, the pool temperature shall be continually monitored and also observed and logged every See ITS 3.6.2.1 } o CD CD 5 minutes until the heat addition isterminated. 0 ITS 3.6.2.1. and b. Water temperature during test operation which Co 0 I ITS 3.6.2.2 adds heat to the suppression pool chal be SR 3.6.1.1.1 C. A visual inspection of the suppression chamber
< Is00 F and shall not be >90F for more than S .6 interior includin liaterregions and the Gne -4' -4' 24 hours. Interior palntec surfaces above the writer ine A4
- c. If the suppression chamber water temperature shall be made at ee uellng is > 110F. Ihe reactor shall be scrammed immediately. Power operation shall not be In accordance with the Containment resumed until the pool temperature Is ::901F. {i See ITS 3.6.2.1 I ULeakage Rate Testing Program 3.714.7 15e 01 20/05 Amendment No. &3,23,141 Page 5 of 6
( ( ITS 3.6.1.1 ITS
. 3.0 LIMmNI CONDITIONS FOR OPERATION -4.0 SURVEILLANCE REQUIREMENTS
. I . . _ ..
- 4. Pressure Suppression Chamber-Drywell Vacuum 4. Pressure Suppression Chamber-Drywell Vacuum Breakers Breakers pi
- 2) a. When primary containment integrity is required. a. Operability and full closure of the I See ITS 3.6.1.7 } 0 0 all eight drywell-suppression chamber vacuum drywell-suppression chamber vacuum breakers breakers shall be operable and positioned In shall be vertfled by performance of the (D the dosed position as indicated by the position following:
0 Indication system, except during testing and a except as specified in 3.7.A.4.b through (1) Monthly each operable dryweli-suppresslon 2 3.7.A.4.d below. chamber vacuum breaker shall be exercised through an opening-closing A.3 Z3 0
- b. Any drywell-suppression chamber vacuum cycle.
breaker may be nonfuly closed as indicated by the position indication and alarm system (2) Once each[r drywell to (D provided that drywell to suppression chamber suppression chamber leakage shall be Add seconndT - 0 0 SR 3.6. 1.1.2 demonstrated to be less than that uency M.2 differential pressure decay does not exceed equivalent to a one-Inch diameter orifice .:A E) that shown on Figure 3.7.1 and each vacuum breaker shell be visually rD CD c. Up to two drywell-suppression chamber vacuurr inspected. (Containment access required) breakers may be Inoperable provided that: (1) 0 the vacuum breakers are determined to be fully (3) Once each operating cyde, vacuum la -9' breaker position indication and alarm dosed and at least one position alarm circult Is systems shall be calibrated and functionally 0 -A) operable or (2) the vacuum breaker Is secured In the dosed position or replaced by a blank tested. (Containment access required)
-(See ITS 3.6.1.71 to 0
CD flange. (4) Once each operating cyde, the vacuum 0)11 breakers shall be tested to determine that CD1
- d. Drywell-suppression chamber vacuum breakers the force required to open each valve from IN may be cycled, one at a time, during fully closed to fully open does not exceed containment inerting and deinerting operations to assist in purging air or nitrogen from the that equivalent to 0.5 psid acting on the suppression chamber face of the valve I
suppression chamber vent header. dIsc. (Cortainment access required.) 3.714.7 164 01128105 Amendment No. a346, 80,404, 141 Page 6 of 6
Attachment 1, Volume 11, Rev. 0, Page 11 of 431 DISCUSSION OF CHANGES ITS 3.6.1.1, PRIMARY CONTAINMENT ADMINISTRATIVE CHANGES A.1 In the conversion of the Monticello Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1433, Rev. 3, "Standard Technical Specifications General Electric Plants, BWR/4" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 CTS 3.7.A.2.a.(1) references the CTS Section 1.0 Primary Containment Integrity definition. ITS does not use this terminology; it requires the primary containment to be OPERABLE. This changes the CTS by deleting the reference to Primary Containment Integrity and replaces it with a requirement for the primary containment to be OPERABLE. This change is acceptable since all requirements of the definition of Primary Containment Integrity have been incorporated into ITS 3.6.1.1, ITS 3.6.1.2, or ITS 3.6.1.3. In addition, ITS 3.6.1.1 requires the primary containment to be OPERABLE. The definition of OPERABLE and the subsequent ITS 3.6.1.1 LCO, ACTIONS and Surveillances are sufficient to encompass the requirements of the CTS definition. This change is designated administrative because it does not result in any technical change to the CTS. A.3 CTS 4.7.A.4.a.(2) requires the drywell to suppression chamber leakage to be demonstrated "once per operating cycle." ITS SR 3.6.1.1.2 requires performance of a similar test every "24 months." This changes the CTS by changing the Frequency from "Once each operating cycle" to "24 months." This change is acceptable because the current "operating cycle" is "24 months". In letter L-MT-04-036, from Thomas J. Palmisano (NMC) to the USNRC, dated June 30, 2004, NMC has proposed to extend the fuel cycle from 18 to 24 months and at the same time has performed an evaluation in accordance with Generic Letter 91-04 to extend the unit Surveillance Requirements from 18 months to 24 months. CTS 4.4.A.2 was included in this evaluation. This change is designated as administrative because it does not result in any technical changes to the CTS. A.4 This change to CTS 4.7.A.1.c is provided in the Monticello ITS consistent with the Technical Specifications Change Request submitted to the USNRC for approval in NMC letter L-MT-04-036, from Thomas J. Palmisano (NMC) to USNRC,.dated June 30, 2004. As such, this change is administrative. MORE RESTRICTIVE CHANGES M.1 CTS 3.7.A.2.a.(1) is applicable at all times when the reactor is critical or when the reactor water temperature is above 2120F and fuel is in the reactor vessel. ITS 3.6.1.1 is applicable in MODES 1, 2, and 3. This changes the CTS by Monticello Page 1 of 4 Attachment 1, Volume 11, Rev. 0, Page 11 of 431
Attachment 1, Volume 11, Rev. 0, Page 12 of 431 DISCUSSION OF CHANGES . ITS 3.6.1.1, PRIMARY CONTAINMENT requiring the Primary Containment to be OPERABLE in MODE 2 when reactor water temperature is less than or equal to 212 0F. The purpose of-CTS 3.7.A.2.a.(1) is to ensure the primary containment is OPERABLE to mitigate the consequences of a design basis accident. Primary containment is required to be OPERABLE during MODES 1, 2, and 3 when a design basis accident could cause a release of radioactive material to the primary containment. In MODES 1 and 3, the reactor coolant temperature will always be above 2120F. In MODE 2, the reactor coolant temperature may be less than or equal to 2120 F when the reactor is subcritical but control rods are withdrawn. Therefore, it is necessary and acceptable to require the primary containment to be OPERABLE. This change is designated as more restrictive because the LCO will be applicable under more reactor operating conditions than in the CTS. M.2 CTS 4.7.A.4.a.(2) requires the drywell to suppression chamber leakage to be demonstrated once each operating cycle. ITS SR 3.6.1.1.2 requires performance of a similar test at a similar Frequency, but also requires the test every 12 months if two consecutive tests fail, and continues at this 12 month Frequency until two consecutive tests pass. This changes the CTS by requiring an increased Surveillance Frequency upon two consecutive test failures. The purpose of CTS 4.7.A.4.a.(2) is to ensure the pressure suppression function of the primary containment is OPERABLE. This change is acceptable because two consecutive test failures could indicate unexpected primary containment degradation. Therefore, more frequent testing is prudent to ensure any further degradation is detected in an appropriate time. This change is designated as more restrictive because Surveillances will be required more frequently under the ITS than under the CTS. M.3 CTS 3.7.A.2.a.(2) states that the Primary Containment Integrity is not required when performing low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed 5 MW(t). The ITS does not include this allowance. This changes the CTS by deleting the allowance to not require Primary Containment Integrity (changed to Primary Containment OPERABILITY as described in DOC A.2) during certain low power physics tests. The purpose of CTS 3.7.A.2.a.(2) is to allow the primary containment to be open during certain low power physics tests to facilitate the low power physics test requirements. This exception is no longer needed at Monticello since all low power physics tests performed with the reactor critical and requiring primary containment integrity requirements to be suspended have been completed. This change is designated as more restrictive because this LCO allowance has been deleted. RELOCATED SPECIFICATIONS None Monticello Page 2 of 4 Attachment 1, Volume 11, Rev. 0, Page 12 of 431
Attachment 1, Volume 11, Rev. 0, Page 13 of 431 DISCUSSION OF CHANGES ITS 3.6.1.1, PRIMARY CONTAINMENT REMOVED DETAIL CHANGES LA.1 (Type 2- Removing Descriptions of System Operation) CTS 1.0.P definition of Primary Containment Integrity states, in part, that "Primary Containment Integrity means that the drywell and pressure suppression chamber are intact," and that "All... manways are closed." ITS 3.6.1.1 does not include these requirements. This changes the CTS by moving these requirements to the ITS Bases. The removal of these details, which are related to system operation, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement for the containment to be OPERABLE and the relocated material describes aspects of OPERABILITY. The ITS 3.6.1.1 Bases, LCO Section states that compliance with this LCO will ensure a primary containment configuration, including equipment hatches and manways, that is structurally sound and that will limit leakage to those leakage rates assumed in the safety analyses. This change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because information relating to system operation is being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES L.1 (Category 7 - Relaxation of Surveillance Frequency, Non-24 Month Type Change) CTS 4.7.A.2.d requires a visual inspection of the interior surfaces of the drywell once each operating cycle for evidence of deterioration. CTS 4.7.A.1.c requires visual inspection of the accessible portions of the suppression chamber interior each refueling interval. ITS SR 3.6.1.1.1 requires visual examinations in accordance with the Primary Containment Leakage Rate Testing program. This changes the CTS by reducing the Frequency of the visual inspections (examination). The purpose of CTS 4.7.A.2.d and CTS 4.7.A.1.c is to ensure the interior surfaces of the containment are free of structural deterioration that might affect the primary containment OPERABILITY. This change extends the Surveillance Frequency from once each operating cycle or each refueling interval (i.e., 24 months) to in accordance with the Primary Containment Leakage Rate Testing Program (i.e., three visual examinations in each ten year period at approximately equal intervals). The visual examination required by CTS 4.7.A.2.a (ITS SR 3.6.1.1.1) duplicates the visual inspection (examination) required by CTS 4.7.A.2.d and CTS 4.7.A.1 .c except for the Frequency of the required examinations. CTS 4.7.A.2.a (ITS SR 3.6.1.1.1) is required by the Primary Containment Leakage Rate Testing Program (which is based on 10 CFR 50, Appendix J, Option B) to be performed prior to each Type A test and two additional times during each 10 year interval. Thus the CTS 4.7.A.2.a (ITS SR 3.6.1.1.1) required visual examination is performed at least 3 times in each 10 year period while the CTS 4.7.A.2.d and CTS 4.7.A.1.c required visual Monticello Page 3 of 4 Attachment 1, Volume 11, Rev. 0, Page 13 of 431
Attachment 1, Volume 11, Rev. 0, Page 14 of 431 DISCUSSION OF CHANGES ITS 3.6.1.1, PRIMARY CONTAINMENT inspections are performed once per 24 months (or five times in a 10 year period). This change is acceptable because the new Surveillance Frequency has been evaluated to ensure that it provides an acceptable level of reliability. The results of examinations conducted over more than 25 years of plant operation and through 17 refuel outages has shown that no significant deterioration has taken place. This operating experience demonstrates that performing the visual examinations at the Frequency required by the Primary Containment Leakage Rate Testing Program (at least three examinations in a 10 year period) is adequate to detect significant deterioration of the accessible interior surfaces of the drywell. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS. Monticello Page 4 of 4 Attachment 1, Volume 11, Rev. 0, Page 14 of 431
Attachment 1, Volume 11, Rev. 0, Page 15 of 431 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 15 of 431
Attachment 1, Volume 11, Rev. 0, Page 16 of 431 Primary Containment 3.6.1.1 CTS 3.6 CONTAINMENT SYSTEMS 3.7.A.2 3.6.1.1 Primary Containment 3.7A.2.a.(1) LCO 3.6.1.1 Primary containment shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 3.7.A.2.a.(4) A. Primary containment A.1 Restore primary 1 hour inoperable. containment to OPERABLE status. 3.7.A.2.a.(4) B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met. AND B.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.7A2.a, SR 3.6.1.1.1 Perform required visual examinations and leakage In accordance 4.7.A.2.d, rate testing except for primary containment air lock with the Primary 4.7.A.1 .c testing, in accordance with the Primary Containment Containment Leakage Rate Testing Program. Leakage Rate Testing Program BWR/4 STS 3.6.1.1-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 16 of 431
Attachment 1, Volume 11, Rev. 0, Page 17 of 431 Primary Containment 3.6.1.1 CTS SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY 4.7.A.4.a.(2) SR 3.6.1.1.2 Verify drywell to suppression chamber/differential m months pressure goes not decrease at a ra > [0.25] inch water gauge per minute tested ov a [10] minute AND 0D period a~ an initial differential pres ure of [1] psi .
------NOTE bypass leakage Is less than that equivalent to a one inch Only required diameter orifice after two 11 consecutive tests fail and continues until two consecutive tests pass X months 0
BWR/4 STS 3.6.1.1-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 17 of 431
Attachment 1, Volume 11, Rev. 0, Page 18 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.1.1, PRIMARY CONTAINMENT
- 1. The brackets have been removed and the proper plant specific information/value has been provided.
Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 18 of 431
Attachment 1,Volume 11, Rev. 0, Page 19 of 431 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 19 of 431
I Attachment 1, Volume 11, Rev. 0, Page 20 of 431 Primary Containment B 3.6.1.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.1 Primary Containment BASES B/ \CKGROUND The function of the primary containment is to isolate and contain fission products released from the Reactor Primary System following a design ressure vessel basis loss of coolant accident (LOCA) and to confine the postulated radioactive material. The primary containment consists of a [enclose drywell. which ls:a steele reinforced concrete i sem an pr surrounds the Reactor n essentially leak tight barrier against an 0 nd a suppression uncontrolled release of radioactive material to the environment. chamber, which is a pressure vessel, The isolation devices for the penetrations in the primary containment connected to the boundary are a part of the containment leak tight barrier. To maintain this drywell by vent pipes. The pnimary leak tight barrier: containment
- a. All penetrations required to be closed during accident conditions are either:
- 1. Capable of being closed by an OPERABLE automatic containment isolation system or (which include_
plugs and caps) 2. Closed by manual valves, blind flange!E or de-activated D automatic valves secured in their closed positions, except as provided in LCO 3.6.1.3, "Primary Containment Isolation Valves (PCIVs)" 0
- b. The primary containment air lock Is OPERABLE, except as provided in LCO 3.6.1.2, "Primary Containment Air Lock<"X 0
- c. All equipment hatches are closece
[d. The prdef sealing me cth ssociated with a pentaiol { exceppnf-ded in LCO 6OPER[ABLE, This Specification ensures that the performance of the primary containment, in the event of a Design Basis Accident (DBA), meets the assumptions used in the safety analyses of References I and 2. SR 3.6.1.1.1 leakage rate requirements are in conformance with 10 CFR 50, Appendix J, Option EM3 (Ref. 3), as modified by approved exemptions. 0 BWR/4 STS B 3.6.1.1-1 Rev. 3.0, 03/31104 Attachment 1, Volume 11, Rev. 0, Page 20 of 431
Attachment 1, Volume 11, Rev. 0, Page 21 of 431 Primary Containment B 3.6.1.1 BASES APPLICABLE The safety design basis for the primary containment is that it must SAFETY withstand the pressures and temperatures of the limiting DBA without ANALYSES exceeding the design leakage rate. The DBA that postulates the maximum release of radioactive material within primary containment is a LOCA. In the analysis of this accident, it is assumed that primary containment is OPERABLE such that release of fission products to the environment is controlled by the rate of primary containment leakage. Analytical methods and assumptions involving the primary containment are presented in References 1 and 2. The safety analyses assume a nonmechanistic fission product release following a DBA, which forms the basis for determination of offsite doses. The fission product release is, in turn, based on an assumed leakage rate from the primary containment. OPERABILITY of the primary containment ensures that the leakage rate assumed in the safety analyses is not exceeded. The maximum allowable leakage rate for the primary containment (L.) is
$1.21% by weight of the containment air per 24 hours at the design basis
[l!3 LOCA maximum peak containment pressure (Pa) oTf15 .5l psig ff 2 otirer 24 houeda reg288l igf28.8 (Ref. 1). Primary containment satisfies Criterion 3 of 10.CFR 50.36(c)(2)(ii). LCO Primary containment OPERABILITY is maintained by limiting leakage to s 1.0 La, except prior to the first startup after performing a required Primary Containment Leakage Rate Testing Program leakage test. At this time the applicable leakage limits must be met. Compliance with this LCO will ensure a primary containment configuration, including equipment hatches that is structurally sound and that will limit leakage to those leakage rates assumed in the safety 03 analyses. Individual leakage rates specified for the primary containment air lock are addressed in LCO 3.6.1.2. APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, primary containment is not required to be OPERABLE in MODES 4 and 5 to prevent leakage of radioactive material from primary containment. BWR/4 STS B 3.6.1.1-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 21 of 431
Attachment 1, Volume 11, Rev. 0, Page 22 of 431 Primary Containment B 3.6.1.1 BASES ACTIONS A.1. In the event primary containment is inoperable, primary containment must be restored to OPERABLE status within 1 hour. The 1 hour Completion Time provides a period of time to correct the problem commensurate with the importance of maintaining primary containment OPERABILITY during MODES 1, 2, and 3. This time period also ensures that the probability of an accident (requiring primary containment OPERABILITY) occurring during periods where primary containment is inoperable is minimal. B.1 and B.2 If primary containment cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.6.1.1.1 REQUIREMENTS Maintaining the primary containment OPERABLE requires compliance with the visual examinations and leakage rate test requirements of the [ Primary Containment Leakage Rate Testing Program. Failure to meet air lock leakage testing (SR 3.6.1.2.1);l~rs nmen sndvent lle61e .1 .1 Iresilient seal primary containment purgiaye (i) H)leakag testing (SR 3.6.1.J71,alfsotefon va lel l~ 31.3,2~does no ncsaiyrsuit in a failure of this SR. The impact of the failure to meet these SRs must be evaluated against the Type A, B, and C acceptance criteria of the Primary Containment Leakage Rate Testing Program. As left leakage prior to the first startup after performing a required Primary Containment Leakage Rate Testing Program leakage test is required to be < 0.6 La for combined Type B and C leakage, and _. r 0 t s 0.75 La or on for overall Type A leakage. At all other times between required leakage rate tests, the acceptance criteria is based on an overall Type A leakage limit of s 1.0 L.. At s 1.0 La the offsite dose consequences are bounded by the assumptions of the safety analysis. The Frequency is required by the Primary Containment Leakage Rate Testing Program. BWR/4 STS B 3.6.1.1-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 22 of 431
Attachment 1, Volume 11, Rev. 0, Page 23 of 431 Primary Containment B 3.6.1.1 BASES ., SURVEILLANCE REQUIREMENTS (continued)
=---------REVIEWER' OTE ------------
Regulatory de 1.163 and NEI 9 1include acceptance criteri oras-left and ound Type A leaka rates and combined Type nd C leake rates, which may eflected in the Bases. A_ _ _ _ _ o_ f _ __ __ _ _ _ _ _ _ _ _ _ _ _ __ ____ ____ 0 SR 3.6.1.1.2 Maintaining the pressure suppression function of primary containment requires limiting the leakage from the drywell to the suppression chamber. Thus, if an event were to occur that pressurized the drywell, the steam would be directed through the downcomers into the suppression pool. masures drywell to suppression chamber differential pressure durring a[ minute period to ensure that the leakage paths that would bypass the suppression pool are within allowable limits. 0 Satisfactory performance of this SR can be achieved by establishing a ass leakagel isless thn known differential pressure between the drywell and the suppression
- equivalent tcva one inc chamber and verifying that thprsuei either the suppression neter orifice K/hmber/r the drywell does o hneb oeta [0.2,51 inch ofwtr 0 24m er te over a 0Imi e . The leakage test is performed every monthsa The8 monthM Frequency was developed considering it is prudent that this Surveillance be performed during a unit outage and also 0 in view of the fact that component failures that might have affected this test are identified by other primary containment SRs. Two consecutive test failures, however, would indicate unexpected primary containment (
degradation; in this event, as the Note indicates, increasing the Frequency to once every-O monthsM is required until the situation is remedied as evidenced by passing two consecutive tsts-.- moh REFERENCES r_,13SAR, Section ff 1ED (I 2.- SAR, Section on 9 *14 2 ( 0
- 3. 10 CFR 50, Appendix J, Option0DEt .
BWR/4 STS B 3.6.1.1-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 23 of 431
Attachment 1, Volume 11, Rev. 0, Page 24 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.1.1 BASES, PRIMARY CONTAINMENT
- 1. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the plant specific nomenclature, number, reference, system description, analysis description, or licensing basis description.
- 2. The brackets have been removed and the proper plant specific information/value has been provided.
- 3. The Reviewers Note is deleted as it is not part of the plant-specific ITS.
- 4. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, NEI 01-03, Section 5.1.3.
- 5. Changes are made to reflect those changes made to the Specification.
- 6. Editorial change made for enhanced clarity.
Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 24 of 431
Attachment 1, Volume 11, Rev.O0, Page 25 of 431 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 11, Rev. 0, Page 25 of 431
Attachment 1, Volume 11, Rev. 0, Page 26 of 431 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.6.1.1, PRIMARY CONTAINMENT There are no specific NSHC discussions for this Specification. Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 26 of 431
, Volume 11, Rev. 0, Page 27 of 431 ATTACHMENT 2 ITS 3.6.1.2, Primary Containment Air Lock Attachment 1, Volume 11, Rev. 0, Page 27 of 431
Attachment 1, Volume 11, Rev. 0, Page 28 of 431 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1,Volume 11, Rev. 0, Page 28 of 431
C C ITS 3.6.1.2 C 0 ITS ITS 3.0 LIMITING CONDIONS FOR OPERATION [ 4.0 SURVEILLANCE REQUIREMENTS
- c. When Primar Containment Inte riv Is I . . . . . . .
Applicability primary containment air lock shail SR 3.6.1.2.1 c. (1) Perlorm required primary containment air lock leakage rate testing In accordance be operable with bth drsdosed caept 0 with the Containment Leakage Rate LCO 3.6.1.2 irlockis tnateast Sten Testing Program.(e(d) C) aneir lock doofhall be dosse SR 3.6.1.2.2 (2) Once per 24 months, verify that only one 0 With the primary oa nment air lock door In the primary containment air kick n maintai at least one air I can be opened at a time. 3 ACTION C closed and restore the air lock to Operable status within 24 hoursfrbe in at least Hot The interior surfaces of the dry-CD uown hin tenext12 hours and in Cold well shag ACTION D visually Inspected each operating cycle for See ITS 3.15.1.1I Shutdown within the following 24 hours. evidence of deterioration.
- a a proposed 0 equi-ActionC. 0 MO,
-o Co -o CD Add proposed ACTIONS Note I and ACTIONS A and B and Condition C<) :A CO -I' i""\ Add proposed ACTIONS Note 2 ;
. * , A02 CD)
CD 0
-9, SR 3.6.1.2.1 Note 1 An Inoperable air lock door does not Invaridato the previous successful performance of the overall air lock leakage test.
SR 3.6.1.2.1 Note 2 Results shall be evaluated against acceptance criteria applicable torSRYA.25a 3.7/4.7 NEXT PAGE IS; 163 160 02/04/03 Amendment No. 66, 60, 06, 132 Page 1 of 2
C C C 0 ITS 3.6.1.2 When a system. subsystem, train, component or device Is determined to be Inoperable soley because Rs emergency power
- >
tu source Is Inoperable, or sodey because its normal power source Is Inoperable It may be considered operable for the purpose of satisfying the requirements of Hs appitcable Umhlng Condition or'Operatlon provided: (1) Rs corresponding normal or emergency power source Is operable; and 2) alloof ns redundant system(s) subsystef(s)n trins(s) component(s) and devIce.s) _ See ITS 3.8.1 } 0 are Operable. or likevise satisfy the requirements of this paragraph. 3 V-4 0 CD
- i. Operating - Operating means that a system or component Is performing Rs specified functions.l c
Operaing Cycle - Intervai between the end of one refueling outage and the end of the next subsequent refueling outage. See ITS 1.01 C) 3 Power Operaln+/- - Power Operation is any operation vith the mode switch in the Start-Up or Run position with the reactor 0 critical and above 1% rated thermal Dower. ED 0 0 2 0 CD 0 CD ED (o i. rI.,.ctIvu *I~*5VIWs~nmLen Lou=v* mlu CD) CD ID.
- 1. Instrument Channol - An Instrument channel means an arrangement of a sensor and auxiliary equipment required to 0
-0 to 0 generate and transmit to a trip system, a single trip signal related to the plant parameter monitored by that Instrument C~) channel.
0 2. Tip System - A trip system means an arrangement of Instrument channel trip signals and auxiliary equipment required to (A) 0 Initiate a protection action. A trip system may require one or more Instrument channel trip signals related to one or more See ITS 1.0} plant parameters to Initiate trip system action. Iniiation of the protective function may require tripping of a single trip system
-9' (e.g.. HPCI system Isolation, off-gas system isolation, reactor buliding Isolation and standby gas treatment Inilation, and rod block), or the coincident tripping of two trip systems (e.g., Inflation of scram, reactor Isolation, and primary containment Isolation).
- 3. Protective Action - An action Initiated by the protection system when a limit Is exceeded. A protective action can be at channel or system level.
3 1/23/84 Amendment No. 21 Page 2 of 2
Attachment 1, Volume 11, Rev. 0, Page 31 of 431
.DISCUSSION OF CHANGES ITS 3.6.1.2, PRIMARY CONTAINMENT AIR LOCK ADMINISTRATIVE CHANGES A.1 In the conversion of the Monticello Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1433, Rev. 3, "Standard Technical Specifications General Electric Plants, BWR/4" (ISTS).
These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 ITS 3.6.1.2 ACTIONS Note 2 states "Enter applicable Conditions and Required Actions of LCO 3.6.1.1, "Primary Containment," when air lock leakage results in exceeding overall primary containment leakage rate acceptance criteria." This requirement is not specifically stated in the CTS. This changes the CTS by explicitly requiring the Primary Containment Actions be entered when the Primary Containment LCO is not met as a result of air lock leakage exceeding limits. This change is acceptable because it reinforces the requirement in CTS 3.7.A.2.a to meet overall containment leakage limits. This change is designated as administrative because it does not result in technical changes to the CTS. A.3 CTS 3.7.A.3.c requires the primary containment air lock to be OPERABLE whenever the Primary Containment Integrity Is required. ITS LCO 3.6.1.2 requires the primary containment air lock to be OPERABLE during MODES 1, 2,
- 3. This changes the CTS by deleting a cross reference to the Primary Containment Integrity Applicability and replacing Itwith the specific Applicability for the primary containment air lock.
The purpose of CTS 3.7.A.3.c is to ensure the primary containment air lock is OPERABLE when necessary to satisfy the safety analyses. This change deletes a cross reference to the Primary Containment Integrity Applicability and replaces with the specific Applicability for the primary containment air lock. The proposed Applicability for the primary containment air lock is the same as the Applicability for the Primary Containment in ITS 3.6.1.1. Changes to the Applicability of the Primary Containment are discussed in the Discussion of Changes for ITS 3.6.1.1. The changes identified in the Discussion of Changes for ITS 3.6.1.1 are applicable for the primary containment air lock. This change is considered to be a format change consistent with the ISTS. This change is designated as administrative because it does not result in technical changes to the CTS. MORE RESTRICTIVE CHANGES M.1 When the primary containment air lock is inoperable, CTS 3.7.A.2.c requires maintaining an air lock door closed, and restoration of the inoperable air lock within 24 hours. Under the same condition, ITS 3.6.1.2 ACTION C not only requires similar actions (as modified by DOC L.1) but also specifies an additional Required Action. Required Action C.1 requires the Immediate initiation of action to evaluate overall containment leakage rate per LCO 3.6.1.1, using current air lock test results. This changes the CTS by adding a new Required Action. Monticello Page 1 of 4 Attachment 1, Volume 11, Rev. 0, Page 31 of 431
Attachment 1, Volume 11, Rev. 0, Page 32 of 431 DISCUSSION OF CHANGES ITS 3.6.1.2, PRIMARY CONTAINMENT AIR LOCK The purpose of ITS 3.6.1.2 Required Action C.1 is to verify that the overall leakage rate aspect of containment OPERABILITY is met in the event an air lock is inoperable for a reason other than one door or an interlock mechanism being inoperable. This change is acceptable because if the inoperability is something that could cause the overall containment leakage rate limits to be exceeded, this should be immediately evaluated. This change is considered more restrictive because it provides a new Required Action. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA.1 (Type I - Removing Details of System Design and System Description, Including Design Limits) CTS 3.7.A.2.c states (in part) what constitutes an OPERABLE containment air lock (i.e., both doors closed except when the air lock is being used, then at least one air lock door shall be closed). ITS LCO 3.6.1.2 does not include this level of detail. This changes the CTS by moving details concerning what constitutes an OPERABLE containment air lock to the ITS Bases. The removal of these details, which are related to system design, from the CTS is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement that the primary containment air locks be OPERABLE. Also, this change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the CTS. LA.2 (Type 2 - Removing Descriptions of System Operation) CTS I .O.P.2 definition of Primary Containment Integrity states that "At least one door in the air lock is closed and sealed." ITS 3.6.1.2 does not include this requirement. This changes the CTS by moving these details to the ITS Bases. The removal of these details, which are related to system operation, from the Technical Specifications, is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement for the air lock to be OPERABLE and the relocated material describes aspects of OPERABILITY. This change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail Monticello Page 2 of 4 Attachment 1, Volume 11, Rev. 0, Page 32 of 431
Attachment 1, Volume 11, Rev. 0, Page 33 of 431 DISCUSSION OF CHANGES ITS 3.6.1.2, PRIMARY CONTAINMENT AIR LOCK change because information relating to system operation is being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES L.1 (Category 3 - Relaxation of Completion Time) CTS 3.7.A.2.c states that with the primary containment air lock inoperable, "maintain" at least one air lock door closed. ITS 3.6.1.2 Required Action C.2 requires a verification that within "1 hour" an air lock door is closed. This changes the CTS by allowing 1 hour to close the air lock door, in lieu of the current immediate time (i.e., maintain). The purpose of the CTS 3.7.A.2 action is to ensure an air lock door is closed to minimize leakage through the air lock, to preclude exceeding leakage rate assumptions in the safety analysis. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing time to repair inoperable features. The level of degradation associated with this CTS action is no worse than that allowed for Primary Containment Integrity (CTS 3.7.A.2.a.(4)) not maintained, and CTS 3.7.A.2.a.(4) (ITS 3.6.1.1 Required Action A.1) allows the primary containment to be inoperable for 1 hour. Also, the primary containment air lock doors are normally closed except for entry and exit. Therefore, the probability that the OPERABLE air lock door is open is low during the 1 hour period. This change is designated as less restrictive because additional time is allowed to close the air lock door than was allowed in the CTS. L.2 (Category 4 - Relaxation of Required Action) CTS 3.7.A.2.c states that with an air lock inoperable (for any reason), maintain at least one air lock door closed and restore the air lock to OPERABLE status within 24 hours or the unit must be shutdown. ITS 3.6.1.2 provides separate ACTIONS for different inoperabilities of the air lock. With an airlock inoperable due to a single inoperable door, ITS 3.6.1.2 ACTION A allows operation for an unlimited amount of time, provided the OPERABLE air lock door is closed in 1 hour and locked closed in 24 hours, and a verification is performed every 31 days that the OPERABLE air lock door remains locked closed. For air lock doors in high radiation areas or areas with limited access due to inerting, this 31 day verification can be performed by administrative means. In addition, the ACTION allows containment entry and exit for up to 7 days under administrative controls. With an air lock interlock mechanism inoperable, ITS 3.6.1.2 ACTION B allows operation for an unlimited amount of time, provided an OPERABLE door in the air lock is closed in 1 hour and locked closed in 24 hours, and a verification is performed every 31 days that an OPERABLE air lock door in the air lock remains locked closed. For air lock doors in high radiation areas or areas with limited access due to inerting, this 31 day verification can be performed by administrative means. In addition, containment entry and exit through the air lock is permissible (i.e., the closed and locked OPERABLE door can be opened) under the control of a dedicated individual. Finally, due to these new ACTIONS, ITS 3.6.1.2 ACTION C, which has similar actions as CTS 3.7.A.2.c (as modified by DOC L.1), only applies to an air lock that is inoperable for reasons other than an inoperable door or an inoperable interlock mechanism. For both of these new ACTIONS as well as Monticello Page 3 of 4 Attachment 1, Volume 11, Rev. 0, Page 33 of 431
Attachment 1, Volume 11, Rev. 0, Page 34 of 431 DISCUSSION OF CHANGES ITS 3.6.1.2, PRIMARY CONTAINMENT AIR LOCK ACTION C, as stated in ITS ACTIONS Note 1, entry and exit (i.e., the closed and locked OPERABLE air lock doors can be opened) is also permissible to perform repairs on the affected air lock components. This changes the CTS by allowing operation for an unlimited amount of time, with certain restrictions, for air locks that are inoperable due to an inoperable door or interlock mechanism. The purpose of the CTS air lock action is to ensure the containment is not allowed to operate indefinitely in a condition such that it cannot perform its safety function. The changes are acceptable because the proposed ACTIONS will still ensure the containment safety function is met. Since there are two redundant doors in each air lock, only one OPERABLE air lock door is needed to be maintained closed to ensure the leak tightness requirements are met. The leak tightness of each door is verified, as required by ITS SR 3.6.1.2.1, in accordance with the Containment Leakage Rate Testing Program.. In addition, the interlock mechanism only ensures that both doors in the air lock are not inadvertently opened at the same time. With either an OPERABLE air lock door locked closed, or a dedicated individual ensuring that only one door at a time is opened, the function of the interlock mechanism is being met. The allowances to open the air lock doors to perform repairs or other reasons is acceptable since the time the door is opened is short and the opening is under administrative controls. Also, for the case where the air lock door is opened per ACTION A Note 2 for reasons other than to effect repairs, the time period (7 days) is short. These changes are designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. Monticello Page 4 of 4 Attachment 1, Volume 11, Rev. 0, Page 34 of 431
Attachment 1, Volume 11, Rev. 0, Page 35 of 431 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 35 of 431
Attachment 1, Volume 11, Rev. 0, Page 36 of 431 Primary Containment Air Lock 3.6.1.2 CTS 3.6 CONTAINMENT SYSTEMS 3.7.A.2.c 3.6.1.2 Primary Containment Air Lock 3.7.A.2.c LCO 3.6.1.2 The primary containment air lock shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTIONS
-NOTES.
DOC L.2 1. Entry and exit is permissible to perform repairs of the air lock components. DOCA.2 2. Enter applicable Conditions and Required Actions of LCO 3.6.1.1, "Primary Containment," when air lock leakage results in exceeding overall containment leakage rate acceptance criteria. I 0 CONDITION REQUIRED ACTION COMPLETION TIME DOC L.2 A. One primary --- ---- NOTES---- containment air lock 1. Required Actions A.1, A.2, door inoperable. and A.3 are not applicable if both doors in the air lock are inoperable and Condition C is entered.
- 2. Entry and exit is permissible for 7 days under administrative controls.
A.1 Verify the OPERABLE door 1 hour is closed. AND A.2 Lock the OPERABLE door 24 hours closed. AND BWR/4 STS 3.6.1.2-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 36 of 431
Attachment 1,Volume 11, Rev. 0, Page 37 of 431 Primary Containment Air Lock 3.6.1.2 CTS ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME DOC L.2 A.3 ---------NOTE----- Air lock doors in high radiation areas or areas with limited access due to inerting may be verified locked closed by
- administrative means.
Verify the OPERABLE door Once per 31 days is locked closed. DOC L.2 B. Primary containment air --------- NOTES------- lock interlock 1. Required Actions B.1, B.2, mechanism inoperable. and B.3 are not applicable if both doors in the air lock are inoperable and Condition C is entered.
- 2. Entry into and exit from containment is permissible under the control of a dedicated individual.
B.1 Verify an OPERABLE door 1 hour is closed. AND B.2 Lock an OPERABLE door 24 hours closed. AND BWR/4 STS 3.6.1.2-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 37 of 431
Attachment 1, Volume 11, Rev. 0, Page 38 of 431 Primary Containment Air Lock 3.6.1.2 CTS ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B.3 ----- NOTE----- DOC L2 Air lock doors in high radiation areas or areas with limited access due to inerting may be verified locked closed by administrative means. Verify an OPERABLE door Once per 31 days is locked closed.
-4 3.7.A.2.C C. Primary containment air C.1 Initiate action to evaluate Immediately lock inoperable for primary containment overall reasons other than leakage rate per Condition A or B. LCO 3.6.1.1, using current air lock test results.
AND C.2 Verify a door is closed. 1 hour AND C.3 Restore air lock to 24 hours OPERABLE status.
-4 4 3.7.A.2.c D. Required Action and D.1 Be in MODE 3. 12 hours associated Completion Time not met. AND D.2 Be in MODE 4. 36 hours BWR/4 STS 3.6.1.2-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 38 of 431
Attachment 1, Volume 11, Rev. 0, Page 39 of 431 Primary Containment Air Lock 3.6.1.2 K) CTS SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
-4 4.7A.2.c.(1) SR 3.6.1.2.1 --- NOTES-- -- a--
- 1. An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.
- 2. Results shall be evaluated against acceptance .
criteria applicable to SR 3.6.1.1.1. Perform required primary containment air lock In accordance leakage rate testing in accordance with the Primary with the Primary Containment Leakage Rate Testing Program. Containment Leakage Rate Testing Program
.4.
4.7-A.2.c.(2) SR 3.6.1.2.2 Verify only one door in the primary containment air 24 months lock can be opened at a time. BWR/4 STS 3.6.1.2-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 39 of 431
Attachment 1, Volume 11, Rev. 0, Page 40 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.1.2, PRIMARY CONTAINMENT AIR LOCK
- 1. Change made to be consistent with the nomenclature used in ITS SR 3.6.1.1.1.
Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 40 of 431
Attachment 1, Volume 11, Rev. 0, Page 41 of 431 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 41 of 431
Attachment 1, Volume 11, Rev. 0, Page 42 of 431 Primary Containment Air Lock B 3.6.1.2 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.2 Primary Containment Air Lock BASES BACKGROUND One double door primary containment air lock has been built into the primary containment to provide personnel access to the drywell and to provide primary containment isolation during the process of personnel entering and exiting the drywell. The air lock is designed to withstand the same loads, temperatures, and peak design internal and external pressures as the primary containment (Ref. 1). As part of the primary containment, the air lock limits the release of radioactive material to the environment during normal unit operation and through a range of transients and accidents up to and including postulated Design Basis Accidents (DBAs). Each air lock door has been designed and tested to certify its ability to withstand a pressure in excess of the maximum expected pressure following a DBA in primary containment. Each of the doors contains gasketed seal double sketed seals an cal leakage rate testing 6ility to ensurel pressure seated doors (i.e., an increase in primary containment internal pressure results in increased sealing force on each door). J air lock is nominally a right circular cyl in diameter, with doors at each end that are interlocked to prevent simultaneous opening. lock is provided with limit switches on both doors that provide aaa m*;canrlropli~on-f-ddor tdoor isopen pr.[A control roof9 IAdwly, 0 indcaionisprovided t1alert the operatorWhnvr an air looin'terlocki t r ~chanism is defeatd.11 During periods when primary containment is not required to be OPERABLE, the air lock interlock mechanism may be disabled, allowing both doors of an air lock to remain open for extended periods when frequent primary containment entry is necessary. Under some conditions as allowed by this LCO, the primary containment may be accessed through the air lock, when the interlock mechanism has failed, by manually performing the interlock function. The primary containment air lock forms part of the primary containment pressure boundary. As such, air lock integrity and leak tightness are essential for maintaining primary containment leakage rate to within limits in the event of a DBA. Not maintaining air lock integrity or leak tightness may result in a leakage rate in excess of that assumed in theiiisafety 0 analysis. BWR/4 STS B 3.6.1.2-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 42 of 431
Attachment 1, Volume I1, Rev. 0, Page 43 of 431 Primary Containment Air Lock B 3.6.1.2 loss of coolant accident ( BASES APPLICABLE SAFETY The DBA that postulates the maximum release of radioactive material within primary containment is LOCA. In the analysis of this accident, it 0D ANALYSES is assumed that primary containment is OPERABLE, such that release of fission products to the environment is controlled by the rate of primary containment leakage. The primary containment is designed with a design basis LoCA maximum allowable leakae rate fLa) of 1.2% by weight of the containment air per 24 hours at the a maximum Peak 0 containment rate forms pressure the basis (Pa) for the o 1psig acceptance (Ref4.g criteria This allowable leakage imposed on the SRs 0D associated with the air lock. Primary containment air lock OPERABILITY is also required to minimize the amount of fission product gases that may escape primary containment through the air lock and contaminate and pressurize the secondary containment. The primary containment air lock satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO As part of the primary containment pressure boundary, the air lock's safety function is related to control of containment leakage rates following a DBA. Thus, the air lock's structural integrity and leak tightness are essential to the successful mitigation of such an event. The primary containment air lock is required to be OPERABLE. For the air lock to be considered OPERABLE, the air lock interlock mechanism must be OPERABLE, the air lock must be in compliance with the Type B air lock leakage test, and both air lock doors must be OPERABLE. The interlock allows only one air lock door to be opened at a time. This provision ensures that a gross breach of primary containment does not exist when primary containment is required to be OPERABLE. Closure of ya single door irtleanJ air lock is sufficient to provide a leak tight barrier following postulated events. Nevertheless, both doors are kept closed 0D when the air lock is not being used for normal entry or exit from primary containment. APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the primary containment air lock is not required to be OPERABLE in MODES 4 and 5 to prevent leakage of radioactive material from primary containment. BWR/4 STS B 3.6.1.2-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 43 of 431
Attachment 1, Volume 11, Rev. 0, Page 44 of 431 Primary Containment Air Lock B 3.6.1.2 BASES ACTIONS The ACTIONS are modified by Note 1, which allows entry and exit to perform repairs of the affected air lock component. If the outer door is inoperable, then it may be easily accessed to repair. If the inner door is the one that is inoperable, however, then a short time exists when the containment boundary is not intact (during access through tho(u allowance door). Th to open the OPERABLE door, even if it means the 3 primary containment boundary is temporarily not intact, is acceptable due to the low probability of an event that could pressurize the primary containment during the short time in which the OPERABLE door is expected to be open. The OPERABLE door must be immediately closed after each entry and exit. The ACTIONS are modified by a second Note, which ensures appropriate remedial measures are taken when necessary. Pursuant to LCO 3.0.6, actions are not required, even if primary containment is exceeding its leakage limit. Therefore, the Note is added to require ACTIONS for LCO 3.6.1.1, "Primary Containment," to be taken in this event. A.1, A.2. and A.3 With one primary containment air lock door inoperable, the OPERABLE door must be verified closed (Required Action A.1) in the air lock. This ensures that a leak tight primary containment barrier is maintained by the use of an OPERABLE air lock door. This action must be completed within 1 hour. The 1 hour Completion Time is consistent with the ACTIONS of LCO 3.6.1.1, which requires that primary containment be restored to OPERABLE status within 1 hour. In addition, the air lock penetration must be isolated by locking closed the OPERABLE air lock door within the 24 hour Completion Time. The 24 hour Completion Time is considered reasonable for locking the OPERABLE air lock door, considering that the OPERABLE door is being maintained closed. Required Action A.3 ensures that the air lock in rabl ohas been isolated by the use of a locked closed OPERABLE air lock door. This ensures that an acceptable primary containment leakage boundary is maintained. The Completion Time of once per 31 days is based on engineering judgment and is considered adequate in view of the low likelihood of a locked door being mispositioned and other administrative controls. Required Action A.3 is modified by a Note that applies to air BWR/4 STS B 3.6.1.2-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 44 of 431
Attachment 1, Volume 11, Rev. 0, Page 45 of 431 Primary Containment Air Lock B 3.6.1.2 BASES ACTIONS (continued) lock doors located in high radiation areas or areas with limited access due to inerting and allows these doors to be verified locked closed by use of administrative controls. Allowing verification by administrative controls is considered acceptable, since access to these areas is typically restricted. Therefore, the probability of misalignment of the door, once it has been verified to be in the proper position, is small. The Required Actions have been modified by two Notes. Note 1 ensures that only the Required Actions and associated Completion Times of Condition C are required if both doors in the air lock are inoperable. With both doors in the air lock inoperable, an OPERABLE door is not available to be closed. Required Actions C.1 and C.2 are the appropriate remedial actions. The exception of Note 1 does not affect tracking the Completion Time from the initial entry into Condition A; only the requirement to comply with the Required Actions. Note 2 allows use of the air lock for entry and exit for 7 days under administrative controls. Primary containment entry may be required to perform Technical Specifications (TS) Surveillances and Required Actions, as well as other activitiesZ inside primary containment that are required by TS or activities e-prr that support TS-required equipment. This Note is not intended to preclude performing other activities (i.e., non-TS-related activities) if the primary containment was entered, using the inoperable air lock, to perform an allowed activity listed above. This allowance is acceptable due to the low probability of an event that could pressurize the primary containment during the short time that the OPERABLE door is expected to be open. B.1. B.2, and B.3 With an air lock interlock mechanism inoperable, the Required Actions and associated Completion Times are consistent with those specified in Condition A. The Required Actions have been modified by two Notes. Note 1 ensures that only the Required Actions and associated Completion Times of Condition C are required if both doors in the air lock are inoperable. With both doors in the air lock inoperable, an OPERABLE door is not available to be closed. Required Actions C.1 and C.2 are the appropriate remedial actions. Note 2 allows entry into and exit from the primary containment under the control of a dedicated individual stationed at the air lock to ensure that only one door is opened at a time (i.e., the individual performs the function of the interlock). BWRI4 STS B 3.6.1.2-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 45 of 431
Attachment 1, Volume 11, Rev. 0, Page 46 of 431 Primary Containment Air Lock B 3.6.1.2 BASES ACTIONS (continued) Required Action B.3 is modified by a Note that applies to air lock doors located in high radiation areas or areas with limited access due to inerting and that allows these doors to be verified locked closed by use of administrative controls. Allowing verification by administrative controls is considered acceptable, since access to these areas is typically restricted. Therefore, the probability of misalignment of the door, once it has been verified to be in the proper position, is small. C.1, C.2, and C.3 If the air lock is inoperable for reasons other than those described in Condition A or B, Required Action C.1 requires action to be immediately initiated to evaluate containment overall leakage rates using current air lock leakage test results. An evaluation is acceptable since it is overly 2 conservative to immediately declare the primary containment inoperable if both doors in an air lock have failed a seal test or if the overall air lock leakage is not within limits. In many instances (e.g., only one seal per door has failed), primary containment remains OPERABLE, yet only 1 hour (according to LCO 3.6.1.1) would be provided to restore the air lock door to OPERABLE status prior to requiring a plant shutdown. In addition, even with both doors failing the seal test, the overall containment leakage rate can still be within limits. Required Action C.2 requires that one door in the primary containment air lock must be verified closed. This action must be completed within the 1 hour Completion Time. This specified time period is consistent with the ACTIONS of LCO 3.6.1.1, which require that primary containment be restored to OPERABLE status within 1 hour. (Ay, the air lock must be restored to OPERABLE status within (Action c.3) 1 24 hours. The 24 hour Completion Time is reasonable for restoring an 0 inoperable air lock to OPERABLE status considering that at least one door is maintained closed in the air lock. D.1 and D.2 If the inoperable primary containment air lock cannot be restored to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. BWR/4 STS B 3.6.1.2-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 46 of 431
Attachment 1,Volume 11, Rev. 0, Page 47 of 431 Primary Containment Air Lock B 3.6.1.2 BASES SURVEILLANCE SR 3.6.1.2.1 REQUIREMENTS Maintaining primary containment air locke OPERABLE requires compliance with the leakage rate test requirements of the Primary 0D Containment Leakage Rate Testing Program. This SR reflects the leakage rate testing requirements with respect to air lock leakage (Type B leakage tests). The acceptance criteria were establishedgduring initial air lock and primary containment OPERABILITY testing. The periodic 0 testing requirements verify that the air lock leakage does not exceed the allowed fraction of the overall primary containment leakage rate. The Frequency is required by the Primary Containment Leakage Rate Testing Program. The SR has been modified by two Notes. Note 1 states that an inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test. This is considered reasonable since either air lock door is capable of providing a fission product barrier in the event of a DBA. Note 2 has been added to this SR requiring the results to be evaluated against the acceptance criteria which is applicable to SR 3.6.1.1.1. This ensures that air lock leakage is properly accounted for in determining the combined Type B and C primary containment leakage. t 0 0 SR 3.6.1.2.2 The air lock interlock mechanism is designed to prevent simultaneous opening of both doors in the air lock. Since both the inner and outer doors of an air lock are designed to withstand the maximum expected post accident primary containment pressure, closure of either door will support primary containment OPERABILITY. Thus, the interlock feature supports primary containment OPERABILITY while the air lock is being used for personnel transit in and out of the containment. Periodic testing of this interlock demonstrates that the Interlock will function as designed and that simultaneous inner and outer door opening will not inadvertently occur. Due to the purely mechanical nature of this interlock, and given that the interlock mechanism is not normally challenged when the primary containment airlock door is used for entry and exit (procedures require strict adherence to single door opening), this test is only required to be performed every 24 months. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage, and the potential for loss of primary containment OPERABILITY if the Surveillance were erformed with the reactor at power. lThe,241month Freq~eay for the interlock is iustifiedb~sed on 0 I eneric pGrating experierncel The 24 month Frequency is based on engineering judgment and is considered adequate given that the interlock is not challenged during the use of the airlock. 0 BWR/4 STS B 3.6.1.2-6 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 47 of 431
Attachment 1, Volume 11, Rev. 0, Page 48 of 431 Primary Containment Air Lock B 3.6.1.2 BASES REFERENCES I. SAR, Section c.8.n.8.25.. 0 ( 2. CFR 50, A ndix J, Optiol[BlX 0 2m-
- SAR, Section 5 0D BWR/4 STS B 3.6.1.2-7 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 48 of 431
Attachment 1, Volume 11, Rev. 0, Page 49 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.1.2 BASES, PRIMARY CONTAINMENT AIR LOCK
- 1. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the plant specific nomenclature, number, reference, system description, analysis description, or licensing basis description.
- 2. The brackets have been removed and the proper plant specific information/value has been provided.
- 3. Editorial change made for enhanced clarity or to be consistent with similar statements in other places in the Bases.
- 4. Typographical error corrected.
- 5. These words have been deleted since the primary containment may need to be entered for reasons related to Technical Specifications activities that are not specifically on "equipment." This could include sampling and inspections. The intent has not been changed in that it must still be related to Technical Specifications.
Monticello Page 1of I Attachment 1, Volume 11, Rev. 0, Page 49 of 431
Attachment 1, Volume 11, Rev. 0, Page 50 of 431 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 11, Rev. 0, Page 50 of 431
Attachment 1, Volume 11, Rev. 0, Page 51 of 431 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.6.1.2, PRIMARY CONTAINMENT AIR LOCK There are no specific NSHC discussions for this Specification. Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 51 of 431
Attachment 1, Volume 11, Rev. 0, Page 52 of 431 ATTACHMENT 3 ITS 3.6.1.3, Primary Containment isolation Valves (PCIVs) Attachment 1, Volume 11, Rev. 0, Page 52 of 431
Attachment 1, Volume 11, Rev. 0, Page 53 of 431 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 11, Rev. 0, Page 53 of 431
C C C ITS 3.6.1.3 ITS ITS 3.0 UMmNG CONDIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS ITS 3.6.4.2 J lSeo ITS 3.6.4.1 and QfJ
.... .. _.1;....
reactor core, operations with a potential for reducing the shutdown margin below that specified In C) specification 3.3.A and handling of irradiated fuel or Su the fuel cask In the secondary containment are to be Immediately suspended if secondary //When associated Ihstrumentation Is CD containment integrity is not maintained. // required to beOPERABLE per LCO3.3.6.1 lt (D Applicability D. Primary Containment Isolto /0D. Primay Containmnont Isolation Valves (PCIVa) 24 months ZI, LCO 3.6.1.3 1. Du r 1. The primary containment automati solaIlon valve 0 Primary Containment a omari tsofationvalves and A2 surveillance shall be performed as follows: 0 all primary system Instrument line flow check valves rAdd proposed CD shal be operable eXcept as specfied in 3.7.D.2 and SR 3.6.1.3.7 a. At least once po a operable times CD Isolation valves that are power operated and
; excluding reactor building-to-suppression SR 3.6.1.3.5 automatically initiated shall be tested for CD Ichamber vacuum breakers SR 3 .6 .1.3 .6 J irimulated automatic lnitiatlon=-d closure times, a lo (D SR 3.6.1.3.8 b. At least once per i a prima
-A (7A.34) systern instniment line flow check valv esesA -o CD (D
- c. Al oml pnioe-prtd islton CD evaTestnge ln time lh the reator pwer less tha 75% of rated 0 C71
- d. least once p r week the ma steam-fine ower-opeate Isolatlon va shall be exercised by Iil closure aid subsequent Ireopenlng.II i Add 1proposed SR 3.6.1.3.4 and SR 3.6.1.3.10 3.714.7 170 01128/05 Amendment No. 3, 71. n, 122. 130,141 Page 1 of 5
( C ITS 3.6.1.3 ITS ITS A.6 Iproposed Note 2 to Required Actions A.2 and
- 2. 4Whenever a containment penetration flow path is ACTION A. Isolated by a valve deactivated In the Isolated C)
- 0) ACTION C position to meet the requirements of TS 3.7.D.2 the position of the deactivated and Isolated valves or -3 ed the isolation device outside primary containment A.1O only sus ratlieastone valve]IeachI I A.2 shal be record am.nhlFor a containment 0 0
line having an Inoperable valve Is deactivated 2 penetration flow path Isolated by a valve 3 the Isolated condition. This requirement may deactivated Inthe Isolated position to meet the 0 be satisfied by use of at least one dosed and atlonsrequirements of TS 3.7.D.2, the position of the deactivated automatic valve, dosed manual cle d deactivated and Isolated valves or Isolation devices 0 valve, blind flange, or check valve with flow Inside primary conainment which have not had their aI posftionlrecordedlIn the revious 92 da s 3 V have their positionrecor pnor nteeng Startup -A only I means or Hot Shutdown from Cold Shutdown, If the primary L4 containment was de-inerted while In Cold (0 ' . Shutdown.* 0' af ACTION B CD CD
- 0
-9, Joierable sjusjor at least one valve In each -0 line having inoperable valves Is deactivated in the isolated condition. This requirement may be satisfied by use of at least one dosed and U1 deactivated automatic valve. dosed manual ACTIONS Note I 0 valve or blind flange. (Deactivated me s electri py or pneumatically disarm o otherwis -4'
- Isolated valves closed to satisfy these requirements may be ACTIONS I* isolated valves closed to satisfy these requirements may be reopened on an Intermittent basis under approved Note 1 reopened on an intermittent basis under approved _ administrative controls.
I ,rtnlstrnttvaconfrnls ai our8 l. lill
.* rk Required Isolation devices In high radiation areas may be verified by use Actions -
_ of administrative means. A2and 3.7/4.7 C.2 Note I 1714
- 092302 Amendment No. 71 77, 06, 130 Page 2 of 5
( C CI ITS 0 ITS ITS 3.6.1.3 ao0 UMITNG CONDITONS FOR OPERA'TION~/ Q,
,
.
I 4.0 SURVELLANCE REQUIREM ANote 2 to Required Acto _ _ I
-t-SR 3.6.1.3.1 Note 3. Q Thelinerting and de operatins 3. Whenever containment purge and vent valves ar perantted by TS 3.7.A5.b shall be via the Required Isolated to meet the requrements of TS 3.7.D.3.b,
.18Inchpurge and vent valve Action D.2 the position of the deactivated and isolated valves 0) 0 SR 3.6.1.3.9 D outside primary containment shall beecoded
- Rulidh~lanu II other pu gng1 -\ monthly."
)
W
.wen primy containment Integrity Is requied, shall bevia the 2-Inch ptrge and/
vent valve bypass Dine and the Standby Gas / Treatment System.
- 0) CD 0
- b. In the event one or more penetration flow paths I Add proposed ACTIONS Note 1 (j3 tD 0 0 ACTION D with one or more containment purge and vent
- - valves not within purge and vent valve leakage limits, reactor operation in the continue provided that within t ubseuent 24 hours. frestors'the valvets) to withir leakage _
ASR 3.6.1.3.1 and CD iir or at least one valve In each rne having a purge and vent valve not within leakage Umits Is D deactivated In the Isolated posiohn. This tD requirement may be satisfied by use of one See ITS 5.5 } 0) 0 dosed and deactivated automatic valve, dosed manuafve or bnfld /(E! vai -o ears e Jl or disarmor _ I - loherf4de secune the vle \J 4. The seat seats of the drywell and suppression chamnber 18.inch purge and vent valves shalf be 0) CD a) replaced at least once eve y six operatin cvcdes. If ACTION E 4. If Spedfeation 3.7.D.t, 3.7.D.2 and 3.7.D.3 cannot U2 0 CsR 3.6.1.3.11- periodic Type0 leakage testlng of the valves 0) be met, Initate normal orderly shutdown and have identifies a comnmon mode test failure attributable to
- 0) I 0 reactorjpn the Cold Shutdown condition within seat seal degradation, then the seat seals of anl Ca
-4k- -4, drywell and suppression chamber 1B-lnch purge _16 and vent valves shall be reolacedI (M.7 )I lated valves in high radiation areas may be vodfied by use I of admhinstration means.
3.74.7Add proposed ACTIONS A. B.\ 3.7/47ndFforPCIVsreuiredtobe M.1 Requid 171a 01/28/05 OPERABLE durng MODE 4 a 5 Aion D.2 Amendment No. 434r 141
Note 1 Page 3 of 5
C C C ITS 3.6.1.3 ITS When a system, subsystem, train, component or device Is determined to be Inoperable soley because Rs emergency power ID source Is Inoperable, or soley because its normal power source Is Inoperable, It may be considered operable for the purpose of S satisfying the requirements of Its applicable Umiting Condition for'Operatlon provided: (1)Rs corresponding normal or _ See ITS 3.8.1 2) 0 emergency power source Is operable; and (2) all of Its redundant system(s), subsystem(s), trains(s), component(s) and device(s) 0 are Operable, or likewise satisfy the requirements of this paragraph. M. Operating Operating means that a system or component Is performing Its specified funcons. N. C Qydl - Interval between the end of one refueling outage and the end of the next subsequent refueling outage. 9rating See ITS 1.0} 0 0. PowerOeratio - Power Operation Is any operation wihthe mode switch in the Start-Up or Run position with the reactor tc a and above % rated thermal pwer. 2 Coniainment InnEEr* marv Containment Intearitv means that the drwell and pressure suppression arre[ IP. See ITS 3.6.1.1 0of t80e folowin condnlons are salsin
- 1. Ail manual containment Isolation valves on lines connecting to the reactor coolant system or containment which are not 0 SR 3.6.1.3.2. quired to be open during accident conditions are closed.
SR 3.6.1.333_ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _
- 2. At least one door in the airlock Is closed and sealed. ITS 3.6.1.2 -U 0 LCO 3.6.1.3 3 Allautomatcontinmen soa areoper are deactived In the osed position or aYleas one vase SR 3.6.1.3.2, .A M blind flange and F mnWjaysFare dosed. (D OD .v la 0
-,. -.
s~v
-0 Q. Protedive Instrumentation Lodc Definitons
- 1. 1r_ - An instrument channel means an arrangement of a sensor and auxiliary equipment required to 0 generate and transmit to a trip system, a single trip signal related to the plant parameter monitored by that Instrument to1 channel.
-4
- 2. TipSystem - A trip system means an arrangement of Instrument channel trip signals and auxillaty equipment required to Initiate a protection action. A trip system may require one or more Instrument channel trip signals related to one or more {See ITS 1.01 0~
plant parameters to Initiate trip system action. Initiation of the protective function may require tripping of a single trip system (e.g.. HPCI system isolation, off-gas system Isolation, reactor building Isolation and standby gas treatment Initiation, and rod block), or the coincident tripping of two trip systems (e.g., Initiation of scram, reactor Isolation, and primary containment Isolation).
- 3. Protective Action - An action Initiated by the protection system when a limit Is exceeded. A protective action can be at channel or system level.
3 1/23/84 Amendment No. 21 Page 4 of 5
(. C. ITS 3.6.1.3 C ITS 0 1 3.0 LIMITING CONDIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS A.2 2. Primary Containment Integrity LCO 3.6.1.3 a. Perform required visual examinations and leakage rate testing except for primary when the reactor Is critical or when I containment air lock testing, In aecordance with Applicability reedor water temperature Is above the Primary Containment Leakage Rate Testing 0 I L fuel IsIn the reactortesselAM Proram.
.. vamp a.... .
0 specified InP.7.A.2.a.(2If 3.7A2.a. 3.7.0. {See ITS 3.1 0.1} CD CD (2) Primary Containment Integnty Is not required CD when performing low power physics tests at II See ITS 3.6.1.1} atmospheric pressure during or after refueling at power levels not to exceed S MW(t). See ITS 3.6.1.1 } 0 (3) Primary Containment Integrity is not required 0 -o when performing reactor vessel hydrostatic or C,' CD 03 0
- '1t
-9' iACTION E -A) 0 (D) Add proposed SR 3.6.1.3.2 and SR 3.6.1.3.3 l.6 3.7/4.7 150 01/28105 Amendment No. 30. 5. 60, g5, io7,132.141 Page 5 of 5
Attachment 1, Volume 11, Rev. 0, Page 59 of 431 DISCUSSION OF CHANGES ITS 3.6.1.3, PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs) ADMINISTRATIVE CHANGES A.1 In the conversion of the Monticello Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1433, Rev. 3, "Standard Technical Specifications General Electric Plants, BWR/4" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 CTS 3.7.D.1 includes the requirements for the "automatic" PCIVs. CTS 3.7.A.2.a.(1) Includes the requirements for all "manual" PCIVs since the CTS definition of Primary Containment Integrity includes these valves. ITS LCO 3.6.1.3 includes the requirements for both types of PCIVs. This changes the CTS by combining the requirements for all PCIVs in one LCO statement. This change is acceptable because this change simply combines the requirements for all PCIVs in one LCO statement. This change is designated as administrative since it does not result in a technical change to the CTS. A.3 CTS 3.7.D.1 includes all requirements for "automatic" PCIVs, except for reactor building-to-suppression chamber vacuum breakers, which are covered under CTS 3.7.A.3. ITS 3.6.1.3 also includes requirements for automatic PCIVs, but the specific exclusion statement of, "except reactor building-to-suppression chamber vacuum breakers," is included in the ITS LCO 3.6.1.3 statement. This changes the CTS by adding a specific exclusion statement concerning the reactor building-to-suppression chamber vacuum breakers. This change is acceptable because the addition of the statement excluding the OPERABILITY requirements of the reactor building-to-suppression chamber vacuum breakers is consistent with the current requirements. This change is designated as administrative since it does not result in a technical change to the CTS. A.4 CTS 3.7.D.2.a provides requirements to be taken for one or more penetration flow paths with one PCIV inoperable while CTS 3.7.D.2.b provides requirements to be taken for one or more penetration flow paths with two PCIVs inoperable. ITS 3.6.1.3 includes an explicit Note (ACTIONS Note 2) that provides instructions for the proper application of the ACTIONS for ITS compliance (i.e., Separate Condition entry is allowed for each penetration flow path). This changes the CTS by providing explicit direction as to how to utilize the ACTIONS when a PCIV is inoperable. This change is acceptable because the addition of the Note reflects the CTS allowance to take the appropriate Actions on a per valve basis. This change is designated as administrative since it does not result in a technical change to the CTS. A.5 CTS 3.7.D does not specifically require Conditions to be entered for systems supported by inoperable containment isolation valves. OPERABILITY of Monticello Page 1 of 14 Attachment 1, Volume 11, Rev. 0, Page 59 of 431
Attachment 1, Volume 11, Rev. 0, Page 60 of 431 DISCUSSION OF CHANGES ITS 3.6.1.3, PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs) supported systems is addressed through the definition of OPERABILITY for each system, and appropriate LCO Actions are taken. ITS 3.6.1.3 ACTIONS Note 3 states "Enter applicable Conditions and Required Actions for systems made inoperable by PCIVs." ITS LCO 3.0.6 provides an exception to ITS LCO 3.0.2, stating "When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered." This changes the CTS by adding a specific statement to require supported system Conditions and Required Actions be entered, whereas in the CTS this would be done without the Note. This change is acceptable because the addition of the ITS Note reflects the CTS requirement to take applicable Actions for inoperable systems. The ITS Note is required because of the addition of ITS LCO 3.0.6, and because the requirement to declare supported systems Inoperable is being retained. This change is designated as administrative because it does not result in any technical changes to the CTS. A.6 CTS 3.7.D does not include a reference to entering applicable Conditions and Actions of the Primary Containment Integrity LCO (CTS 3.7.A.2) (changed to Primary Containment OPERABILITY in the ITS). ITS 3.6.1.3 ACTIONS Note 4 states "Enter applicable Conditions and Required Actions of LCO 3.6.1.1, "Primary Containment," when PCIV leakage results in exceeding overall containment leakage rate acceptance criteria." This changes the CTS by explicitly stating an existing requirement that the Primary Containment Specification ACTIONS be taken when the Primary Containment LCO is not met as a result of PCIV leakage exceeding limits. This change is acceptable because it reinforces the existing CTS requirement to meet overall primary containment leakage limits. This change is designated as administrative because it does not result in any technical changes to the CTS. A.7 CTS 4.7.D.1.a requires the OPERABLE automatic PCIVs to be tested once per operating cycle. CTS 4.7.D.1.b requires the primary system instrument line flow check valves to be tested once per operating cycle. ITS SR 3.6.1.3.5 requires verification of automatic PCIV isolation time, except for main steam isolation valves (MSIVs), every "24 months," while ITS SR 3.6.1.3.6 requires the verification of MSIV isolation time every "24 months." ITS SR 3.6.1.3.7 requires verification every "24 months" that each automatic PCIV actuates to the isolation position on an isolation signal. ITS SR 3.6.1.3.8 requires verification every "24 months" that each excess flow check valve actuates on a simulated instrument line break to restrict flow to < 2 gpm. This changes the CTS by changing the Frequency from "operating cycle" to "24 months." This change is acceptable because the current "operating cycle" is "24 months". In letter L-MT-04-036, from Thomas J. Palmisano (NMC) to the USNRC, dated June 30, 2004, NMC has proposed to extend the fuel cycle from 18 to 24 months and the same time has performed an evaluation in accordance with Generic Letter 91-04 to extend the unit Surveillance Requirements from 18 months to 24 months. CTS 4.7.D.1.a and CTS 4.7.D.1.b were included in this evaluation. Monticello Page 2 of 14 Attachment 1, Volume 11, Rev. 0, Page 60 of 431
Attachment 1, Volume 11, Rev. 0, Page 61 of 431 DISCUSSION OF CHANGES ITS 3.6.1.3, PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs) This change is designated as administrative because it does not result in any technical changes to the CTS. A.8 CTS 4.7.D.1.b requires each primary system instrument line excess flow check valve (EFCVs) to be tested for proper operation. Per the CTS Bases, the OPERABILITY requirements are specified in a letter from L. 0. Mayer (Northern States Power) to J. F.O'Leary (NRC) dated July 27, 1973. This letter requires the valves to limit leakage to a maximum of 2 gpm. ITS SR 3.6.1.3.8 requires the verification that the reactor instrumentation line EFCV actuates on a simulated instrument line break to restrict flow to < 2 gpm. This changes the CTS by specifying the leakage limit for the individual EFCVs. The purpose of CTS 4.7.D.1.b is to ensure the EFCVs limit leakage on an instrument line break. This change is acceptable since it is consistent with the current requirements, as specified in the CTS Bases. This change is designated as administrative since it does not result in a technical change to the CTS. A.9 CTS 3.7.D.2.a requires restoring the inoperable valve to OPERABLE status within 4 hours, 8 hours, or 72 hours (based on the kind of valve) or requires at least one valve in each line having an inoperable valve to be deactivated in the isolated condition. CTS 3.7.D.2.b requires restoring the inoperable valves to OPERABLE status within 1 hour or requires at least one valve in each line having inoperable valves to be deactivated in the isolated condition. CTS 3.7.D.3.b requires restoring the inoperable valve(s) to within leakage limits within 24 hours or requires at least one valve in each line having a purge and vent valve not within leakage limits to be deactivated in the isolated position. The ITS 3.6.1.3 ACTIONS do not include the specific option to restore the valve(s) to OPERABLE status or restore leakage to within leakage limits, but includes other compensatory Required Actions to take within 1 hour, 4 hours, 8 hours, or 72 hours, as applicable. This changes the CTS by not explicitly stating the requirement to restore an inoperable valve to OPERABLE status or to within leakage limits. The purpose of CTS 3.7.D.2.a, CTS 3.7.D.2.b, and CTS 3.7.D.3.b is to provide appropriate compensatory actions for inoperable PCIVs. This change is acceptable because the technical requirements have not changed. Restoration of compliance with the LCO is always an available Required Action and it is the convention in the ITS to not state such "restore" options explicitly unless it is the only action or is required for clarity. This change is designated as administrative because it does not result in any technical changes to the CTS. A.10 CTS 4.7.D.2 and CTS 4.7.D.3 require the position of the deactivated and isolated valves or the isolation device(s) to be "recorded." ITS 3.6.1.3 Required Actions A.2, C.2, and D.2 only include the requirement to "verify" the applicable valve is "closed." This changes the CTS by deleting the specific requirement to "record" the valve position. The purpose of CTS 4.7.D.2 and CTS 4.7.D.3 is to ensure the affected penetration flow path is isolated. This change Is acceptable because this requirement duplicates the requirements of 10 CFR 50 Appendix B, Section XVII (Quality Assurance Records) to maintain records of activities affecting quality, Monticello Page 3 of 14 Attachment 1, Volume 11, Rev. 0, Page 61 of 431
Attachment 1, Volume 11, Rev. 0, Page 62 of 431 DISCUSSION OF CHANGES ITS 3.6.1.3, PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs) including the results of tests (i.e., Technical Specification Surveillances). Compliance with 10 CFR 50 Appendix B is required by the Monticello Operating License, which is adequate to ensure appropriate data is taken and maintained. The details of the regulations within the Technical Specifications are repetitious and unnecessary. Therefore, retaining the requirement to perform the associated Surveillance and eliminating the details from Technical Specifications that are found in 10 CFR 50 Appendix B is considered a presentation preference. As such, this change is considered an administrative change. A.1 1 CTS 4.7.D.4 discusses the periodic Type C leakage testing of the 18 inch primary containment purge and vent valves (which is required by CTS 4.7.A.2.a). ITS SR 3.6.1.3.11 requires the performance of leakage rate testing for each 18 inch primary containment purge and vent valve with resilient seals in accordance with the Primary Containment Leakage Testing Program. This changes the CTS by stating to perform leakage rate testing for each 18 inch primary containment purge and vent valve in accordance with the Primary Containment Leakage Testing Program. The purpose of CTS 4.7.D.4 is to ensure the 18 inch primary containment purge and vent valve seals remain within the leakage limits. The Primary Containment Leakage Rate Testing Program requires type Ctesting of these valves. ITS SR 3.6.1.3.11 simply clarifies the requirement, and as such, is acceptable. The 18 inch primary containment vent and purge valves will continue to be tested in accordance with the Primary Containment Leakage Rate Testing Program. This change is designated as administrative because it does not result in any technical changes to the CTS. A.12 CTS 1.0.P definition of Primary Containment Integrity states, in part, that all automatic containment isolation valves are OPERABLE "or are deactivated in the closed position or at least one valve in each line having an inoperable valve is closed." CTS 3.7.D.1 requires all primary containment automatic isolation valves to be OPERABLE and CTS 3.7.D.2 and CTS 3.7.D.3 provide the actions that must be taken when the valves are not OPERABLE, and include similar requirements as are in the CTS 1.0.P definition. ITS LCO 3.6.1.3 requires all PCIVs to be OPERABLE and the appropriate compensatory actions for PCIVs are included in the ITS 3.6.1.3 ACTIONS. This changes the CTS by deleting the explicit CTS Primary Containment Integrity definition for when an automatic containment isolation valve is not OPERABLE. The purpose of CTS 3.7.0 is to provide the appropriate requirements for automatic PCIVs. The requirements in CTS 3.7.D.1 are duplicative of the first part of CTS 1.0.P.3 that the automatic containment isolation valves must be OPERABLE. The requirements in CTS 3.7.D.2 and CTS 3.7.D.3 include similar requirements as CTS 1.0.P.3 to isolate the penetration when an automatic containment isolation valve is inoperable. The requirements of CTS 3.7.0 have been incorporated in ITS 3.6.1.3. Since the requirements prescribed in CTS 3.7.0 are retained in ITS 3.6.1.3, deletion of the CTS definition for these requirements In the Primary Containment Integrity definition is acceptable. These changes are designated as administrative changes because they do not result in technical changes to the CTS. Monticello Page 4 of 14 Attachment 1, Volume 11, Rev. 0, Page 62 of 431
Attachment 1, Volume 11, Rev. 0, Page 63 of 431 DISCUSSION OF CHANGES ITS 3.6.1.3, PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs) MORE RESTRICTIVE CHANGES M.1 CTS 3.7.D.1 requires the automatic PCIVs and excess flow check valves to be OPERABLE during reactor power operating conditions (i.e., > 1% RATED THERMAL POWER (RTP)). ITS LCO 3.6.1.3 requires the PCIVs to be OPERABLE in MODES 1,2, and 3, and when associated instrumentation is required to be OPERABLE per LCO 3.3.6.1, "Primary Containment Isolation Instrumentation." ITS 3.6.1.3 also includes ACTIONS (ACTIONS A, B, and F) to cover the new Applicability of "when associated instrumentation is required to be OPERABLE per LCO 3.3.6.1" (i.e., during MODES 4 and 5). This changes the CTS by requiring the PCIVs to be OPERABLE in MODE 2 when < 1%RTP, in MODE 3, and in MODES 4 and 5 when associated instrumentation is required to be OPERABLE per LCO 3.3.6.1, and by addition new ACTIONS for the third new Applicability. The purpose of CTS 3.7.D.1 is to ensure the PCIVs are OPERABLE when required to mitigate the consequences of an accident. The change adding the MODES 2 and 3 Applicabilities is considered acceptable because in MODES 2 and 3, the pressure and temperature in the reactor coolant is elevated or the reactor may be critical or is approaching criticality, and therefore a design basis accident could cause a release of radioactive material to primary containment. In addition, these new Applicabilities are consistent with the Applicability of the Primary Containment Specifications (ITS 3.6.1.1), which is supported by this PCIV Specification. This change also adds a new Applicability for certain PCIVs. The new Applicability of "when associated instrumentation is required to be OPERABLE per LCO 3.3.6.1" effectively adds a MODES 4 and 5 requirement to the Residual Heat Removal Shutdown Cooling System isolation valves, since these are the only valves with instrumentation requirements in ITS 3.3.6.1 in MODES other than MODES 1,2, and 3. OPERABILITY of these valves is necessary to preclude an inadvertent draindown of the reactor vessel through the shutdown cooling isolation valves that could lower the reactor vessel water level to the top of the fuel. Appropriate ACTIONS have been added (ITS 3.6.1.3 ACTIONS A, B, and F). ACTIONS A and B are consistent with the ACTIONS for other PCIVs in MODES 1, 2, and 3, while ACTION F applies when ACTIONS A and B are not met. These changes are designated as more restrictive because the PCIVs will be required to be OPERABLE under more conditions under the ITS than under the CTS. M.2 .CTS 3.7.A.2.a.(1) requires the Primary Containment Integrity as defined in Section 1 to be maintained and is applicable at all times when the reactor is critical or when the reactor water temperature is above 2120F and fuel is in the reactor vessel. The Primary Containment Integrity definition requires all manual primary containment isolation valves that are not required to be open during accident conditions to be closed. ITS 3.6.1.3 is applicable in MODES 1, 2, and 3 for these valves. This changes the CTS by requiring the manual PCIVs to be OPERABLE in MODE 2 when the reactor water temperature is less than or equal to 212 0F. The purpose of CTS 3.7.A.2.a.(1) is to ensure the primary containment is OPERABLE (in this case the manual PCIVs required to be closed during Monticello Page 5 of 14 Attachment 1, Volume 11, Rev. 0, Page 63 of 431
Attachment 1, Volume 11, Rev. 0, Page 64 of 431 DISCUSSION OF CHANGES ITS 3.6.1.3, PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs) accident conditions) to mitigate the consequences of a design basis accident. The manual PCIVs are required to be OPERABLE during MODES 1,2, and 3 when a design basis accident could cause a release of radioactive material to the primary containment. In MODES 1 and 3, the reactor coolant temperature will always be above 212 0F. In MODE 2, the reactor coolant temperature may be less than or equal to 2120F when the reactor is subcritical but control rods are withdrawn. Therefore, it is necessary and acceptable to require the manual PCIVs to be OPERABLE. This change is designated as more restrictive because the LCO will be applicable under more reactor operating conditions than in the CTS. M.3 CTS 4.7.D.l.a requires the measurement of the closure times of all power operated and automatically initiated PCIVs. ITS SR 3.6.1.3.6 requires verification that the isolation time of each MSIV is > 3 seconds and
< 9.9 seconds. This changes the CTS by specifying the explicit acceptance criteria for the MSIV isolation time.
The purpose of CTS 4.7.D.1 .a is to ensure the PCIVs will close in order to isolate the containment consistent with the transient and design basis events. This changes adds the acceptance criteria for the MSIV isolation time. The specified lower limit will ensure the MSIVs will close slow enough to minimize the pressurization event of a full MSIV closure event and the upper limit ensures the consequences of a main steam line break will be within the calculated radiological consequences. This change is acceptable because the acceptance criteria are consistent with the times assumed in the safety analyses. This change is designated as more restrictive because it adds specific isolation time acceptance criteria for the MSIVs that are not currently specified in the CTS. M.4 CTS 4.7.D does not provide any specific testing requirements for the traversing incore probe (TIP) shear isolation valve explosive squib. ITS SR 3.6.1.3.4 requires a verification of continuity of the TIP shear isolation valve explosive charge every 31 days and ITS SR 3.6.1.3.10 requires the removal and testing of the explosive squib from each shear isolation valve of the TIP System every 24 months on a STAGGERED TEST BASIS. This changes the CTS by requiring two new Surveillance Requirements for verifying TIP shear isolation valve explosive squib OPERABILITY. The purpose of ITS SR 3.6.1.3.4 and SR 3.6.1.3.10 is to ensure that the TIP shear isolation valve explosive squibs are OPERABLE. This change is acceptable because it provides additional assurance that the TIP shear isolation valve explosive squibs are OPERABLE. This change is designated as more restrictive because it adds two new Surveillance Requirements to the CTS. M.5 CTS 3.7.D.3 requires the 18 inch primary containment purge and vent valves to be closed except during certain allowed conditions and to be equipped with 40 degree limit stops. However, no Surveillance Requirements are provided to periodically verify these requirements. ITS SR 3.6.1.3.1 requires a 31 day verification that the 18 inch primary containment purge and vent valves are closed (except under certain allowed conditions) and ITS SR 3.6.1.3.9 requires a 24 month verification that each 18 inch primary containment purge and vent Monticello Page 6 of 14 Attachment 1, Volume 11, Rev. 0, Page 64 of 431
Attachment 1, Volume 11, Rev. 0, Page 65 of 431 DISCUSSION OF CHANGES ITS 3.6.1.3, PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs) valve is blocked to restrict the valve from opening > 40 degrees. This changes the CTS by adding two new Surveillance Requirements for verifying the OPERABILITY of the 18 inch primary containment purge and vent valves. The purpose of ITS SR 3.6.1.3.1 and SR 3.6.1.3.9 is to ensure that the 18 inch primary containment purge and vent valves are OPERABLE. This change is acceptable because it provides additional assurance that the 18 inch primary containment purge and vent valves are OPERABLE. This change is designated as more restrictive because it adds two new Surveillance Requirements to the CTS. M.6 While CTS 1.0 provides requirements for manual and non-automatic valves, the CTS does not provide any specific testing requirements for the manual and non-automatic PCIVs. ITS SR 3.6.1.3.2 requires a verification that each primary containment manual valve and blind flange that is located outside primary containment and not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed every 31 days. ITS SR 3.6.1.3.3 requires a verification that each primary containment manual valve and blind flange that is located inside primary containment and not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed prior to entering MODE 2 or 3 from MODE 4 if primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days. In addition, both these Surveillances are modified by two Notes that allow valves and blind flanges in high radiation areas to be verified by use of administrative means and allows PCIVs to be open under administrative controls. This changes the CTS by requiring two new Surveillance Requirements for verifying the OPERABILITY of manual and non-automatic PCIVs. The purpose of ITS SR 3.6.1.3.2 and SR 3.6.1.3.3 is to ensure that each primary containment manual isolation valve and blind flange is closed. This change is acceptable because it provides additional assurance that the primary containment manual isolation valves and blind flanges are closed. This change is designated as more restrictive because it adds two new Surveillance Requirements to the CTS. M.7 CTS 3.7.D.4 requires the unit to be placed in the cold shutdown condition within 24 hours if Specifications 3.7.D.1, 3.7.D.2, and 3.7.D.3 cannot be met. However, CTS 3.7.D.1, 3.7.D.2, and 3.7.D.3 are only applicable in the reactor power operating conditions (i.e.,.> 1%RTP). Thus the unit is only required to be
< 1% RTP in 24 hours. ITS 3.6.1.3 ACTION E requires the unit to be in MODE 3 in 12 hours and MODE 4 in 36 hours. This changes the CTS by requiring the unit to be in MODE 3 in 12 hours and in cold shutdown (i.e., MODE 4) in 36 hours, in lieu of being < 1%RTP in 24 hours.
The purpose of CTS 3.6.D.4 is to place the unit outside of the Applicability of the Specification within a reasonable amount of time. CTS 3.7.D.1, 2, and 3 require the PCIVs to be OPERABLE during reactor power operating conditions (i.e., > 1% RTP). Thus, while the CTS Action requires a shutdown to MODE 4, in actuality, only a shutdown to < 1%RTP is required.. Once < 1% RTP is achieved, continuation to MODE 4 is not required since the PCIVs are not required OPERABLE when < 1% RTP. However, since the requirement that the Monticello Page 7 of 14 Attachment 1, Volume 11, Rev. 0, Page 65 of 431
Attachment 1, Volume 11, Rev. 0, Page 66 of 431 DISCUSSION OF CHANGES ITS 3.6.1.3, PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs) PCIVs be OPERABLE in MODE 2 < 1% RTP and in MODE 3 has been added (DOC M.1), ITS 3.6.1.3 ACTION E includes a shutdown to MODE 3 and to MODE 4. The allowed Completion Times are reasonable, based on operating experience, to reach required unit conditions from full power conditions in an orderly manner and without challenging unit systems. This change is acceptable because it requires the unit to be in an intermediate condition (MODE 3) sooner than is currently required (12 hours versus 24 hours). This portion of the change reduces the time the unit would be allowed to continue to operate > 1% RTP once the condition is identified. The consequences of a loss of coolant event are significantly reduced when the reactor is shutdown and a controlled cooldown is already in progress. This change is designated as more restrictive because less time is allowed to shut down the unit. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA.1 (Type 6 - Removal of LCO, SR, or other TS requirement to the TRM, USAR, ODCM, OQAP, IST Program, or lIP) CTS 4.7.D.l .c states that "All normally open power-operated isolation valves shall be tested in accordance with the Inservice Testing Program. Main Steam isolation valves shall be tested (one at a time) with the reactor power less than 75% of rated." CTS 4.7.D.1.d states, "At least once per week the main steam line power-operated isolation valves shall be exercised by partial closure and subsequent reopening." ITS 3.6.1.3 does not include these requirements. This changes the CTS by relocating these Surveillances Requirement to the Technical Requirements Manual (TRM). The removal of this Surveillance Requirement from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. ITS LCO 3.6.1.3 still requires the PCIVs to be OPERABLE, includes a Surveillance Requirement that ensures the automatic valves will close on an automatic isolation signal and Surveillance Requirements to ensure the isolation times of automatic valves are within limits. Also, this change is acceptable because this type of Surveillance Requirement will be adequately controlled in the TRM. The TRM is incorporated by reference into the USAR and any changes to the TRM are made under 10 CFR 50.59, which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because a requirement is being removed from the Technical Specifications. LA.2 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) A parenthetical phrase to CTS 3.7.D.2.a, CTS 3.7.D.2.b, and CTS 3.7.D.3.b states that "Deactivated means electrically or pneumatically disarm or otherwise secure the valve." ITS 3.6.1.3 does not define the term "de-activated." This changes CTS by moving the intent of the word "de-activated" to the ITS Bases. Monticello Page 8 of 14 Attachment 1, Volume 11, Rev. 0, Page 66 of 431
Attachment 1, Volume 11, Rev. 0, Page 67 of 431 DISCUSSION OF CHANGES ITS 3.6.1.3, PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs) The removal of these details for performing Required Actions from the Technical Specifications is acceptable because this type of Information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement to close and de-activate the valves. Also, this change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5 of the ITS. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications. LA.3 (Type 2 - Removing Descriptions of System Operation) CTS 3.7.D.3.a specifies that inerting and de-inerting operations permitted by TS 3.7.A.5.b shall be via the 18 inch purge and vent valves "aligned to the Reactor Building plenum and vent." ITS SR 3.6.1.3.1 includes a Note that states that the 18 inch primary containment purge and vent valves may be opened for inerting, de-inerting, pressure control, ALARA, or air quality considerations for personnel entry, or Surveillance that require the valves to be open, but does not specify that the 18 inch purge and vent valves must be aligned to the Reactor Building Plenum and vent. This changes CTS by moving the details on how to purge and vent through the 18 inch valves to the ITS Bases. The change adding additional reasons to use the 18 inch valves are discussed in DOC L.6. The removal of these details, which are related to system operation, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement to limit when the 18 inch primary containment purge and vent valves may be opened. Also, this change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5 of the ITS. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because information relating to system operation is being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES L.1 (Category 4 - Relaxation of Required Action) CTS 3.7.D.3.b requires the associated penetration flow path isolated when a primary containment purge or vent valve is not within purge and vent valve leakage limits. However, this action does not include a provision, similar to that allowed for CTS 3.7.D.2.a and b, that isolated valves closed to satisfy the requirements may be reopened on an intermittent basis under administrative controls. ITS 3.6.1.3 ACTIONS Note I allows any primary containment penetration flow path, including the containment purge and vent valve flow paths, isolated due to a leakage limit not being met, to be unisolated intermittently under administrative controls. This changes the CTS by allowing the containment purge and vent valve penetrations to be opened Monticello Page 9 of 14 Attachment 1, Volume 11, Rev. 0, Page 67 of 431
Attachment 1, Volume 11, Rev. 0, Page 68 of 431 DISCUSSION OF CHANGES ITS 3.6.1.3, PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs) under administrative controls when containment purge and vent valve leakage is not within limit. The purpose of ITS 3.6.1.3 ACTIONS Note 1 is to allow the affected primary containment purge and vent valves to be opened on an intermittent basis as required for such evolutions as performing Surveillances, repairs, and routine evolutions. This change is acceptable because the allowance requires administrative controls to be in place when the containment vent and purge valves are opened under administrative controls. The ITS 3.6.1.3 Bases states that these administrative controls consist of stationing a dedicated individual at the controls of the valve, who is in continuous communication with the control room. In this way, the penetration can be rapidly isolated when a need for primary containment isolation is indicated. This allowance is also acceptable due to the low probability of an event that could pressurize the primary containment during the short time in which the PCIV is open and the administrative controls established to ensure the affected penetration can be isolated when a need for primary containment isolation is indicated. In addition, this allowance is currently provided for isolated containment purge and vent valve penetration flow paths if they are isolated due to a purge or vent valve inoperability not related to not meeting leakage limits (CTS 3.7.D.2.a and b, footnote *). This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. L.2 (Category 6 - Relaxation Of Surveillance Requirement Acceptance Criteria) CTS 4.7.D.1.a requires the performance of a simulated automatic initiation test of all power operated and automatic initiated PCIVs. ITS SR 3.6.1.3.7 requires the verification each automatic PCIV actuates to the isolation position on an actual or simulated automatic isolation signal. This changes the CTS by explicitly allowing the use of an actual signal for the test. The purpose of CTS 4.7.D.1 .a is to ensure that the PCIVs operate correctly upon receipt of an actuation signal. This change is acceptable because it has been determined that the relaxed Surveillance Requirement acceptance criteria are not necessary for verification that the equipment used to meet the LCO can perform its required functions. Equipment can not discriminate between an "actual" or "simulated" signal and, therefore, the results of the testing are unaffected by the type of signal used to initiate the test. This change allows taking credit for unplanned actuation if sufficient information is collected to satisfy the Surveillance test requirements. The change also allows a simulated signal to be used, if necessary. This change is designated as less restrictive because less stringent Surveillance Requirements are being applied in the ITS than were applied in the CTS. L.3 (Category 4 - Relaxation of Required Action) CTS 3.7.D.2.a covers the condition of one or more penetration flow paths with one PCIV inoperable and allows reactor operation in the "run mode" to continue for a short period of time. CTS 3.7.D.2.b covers the condition of one or more penetration flow paths with two PCIVs inoperable and allows reactor operation in the "run mode" to continue for a short period of time. CTS 3.7.D.3.b covers the condition of one or more penetration flow paths with one or more containment purge and vent valves not within purge and vent valve leakage limits and allows reactor operation in the Monticello Page 10 of 14 Attachment 1, Volume 11, Rev. 0, Page 68 of 431
Attachment 1, Volume 11, Rev. 0, Page 69 of 431 DISCUSSION OF CHANGES ITS 3.6.1.3, PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs)
"run mode" to continue for a short period of time. The "run mode" is when the reactor mode switch is in the "run" position. CTS 3.7.D.1 is applicable in the reactor "power operating" conditions, which include either the mode switch in the "run" or "start-up" position when power is > 1% RTP (CTS 1.0.0). Therefore, if the unit is not in the "run mode" when a PCIV becomes inoperable, then CTS 3.7.D.4, which requires a unit shutdown to cold shutdown within 24 hours, would apply. The CTS does not allow unit operation to continue when the mode switch is in the "start-up" position using the allowances provided in CTS 3.7.D.2.a, 3.7.D.2.b, or 3.7.D.3.b. ITS 3.6.1.3 ACTIONS A, B, and C cover similar conditions as in CTS 3.7.D.2.a, 3.7.D.2.b, and 3.7.D.3.b, and allows continued operation (similar to that allowed by CTS 3.7.D.2.a, 3.7.D.2.b, and 3.7.D.3.b as modified by applicable ITS 3.6.1.3 DOCs) regardless of the initial position of the reactor mode switch. This changes the CTS by allowing operation to continue with inoperable PCIVs in any MODE, not just MODE 1 (i.e., the "run mode").
The purpose of CTS 3.7.D.2.a, 3.7.D.2.b, and 3.7.D.3.b is to provide appropriate compensatory actions for inoperable PCIVs. CTS 3.7.D.2.a, 3.7.D.2.b, and 3.7.D.3.b include a specific statement that the associated compensatory action is only allowed in the "run mode." CTS 3.7.D.1 requires PCIVs to be OPERABLE during reactor "power operating" conditions. CTS 1.0.0 states the definition of Power Operation as "Power Operation is any operation with the mode switch in the "Start-Up" or "Run" position with the reactor critical and above 1% rated thermal power." The intent of CTS 3.7.D.2.a, 3.7.D.2.b, and 3.7.D.3.b is to provide the appropriate compensatory actions for whenever the specific conditions exist and the PCIVs are required to be OPERABLE. With the unit operating with the reactor mode switch in the "run" position the unit may be operating at 100% RTP. This change is considered acceptable because with the reactor mode switch in the "start-up" position, the unit is operating at low power levels where the intermediate range monitors are on scale. At these low power levels, there is no reason to require the unit to immediately enter a shutdown action, since the consequences of a design basis accident in which primary containment isolation is necessary are no more severe at this low power level than at 100% RTP. The compensatory actions of CTS 3.7.D.2.a, 3.7.D.2.b, and 3.7.D.3.b should be equally applied whenever the inoperability occurs. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. L.4 (Category 4 - Relaxation of Required Action) CTS 3.7.D.2.a states that with one or more penetration flow paths with one PCIV inoperable, operation...may continue provided that within the subsequent 4 hours (8 hours for MSIVs and 72 hours for EFCVs)...at least one valve in each line having an inoperable valve is deactivated in the isolated condition. This action covers the condition for penetrations with either one or two PCIVs. ITS 3.6.1.3 ACTION C covers inoperabilities associated with penetrations with one PCIV and allows a 4 hour Completion Time to isolate the affected penetration except for EFCVs and penetrations with a closed system and a 72 hour Completion Time to isolate the affected penetration for EFCVs and penetrations with a closed system. This changes the CTS by extending the Completion Time from 4 hours to 72 hours for an inoperable PCIV associated with a closed system. Monticello Page 11 of 14 Attachment 1, Volume 11, Rev. 0, Page 69 of 431
Attachment 1, Volume 11, Rev. 0, Page 70 of 431 DISCUSSION OF CHANGES ITS 3.6.1.3, PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs) The purpose of CTS 3.7.D.2.a is to provide a degree of assurance that the penetration flow path with an inoperable PCIV maintains the primary containment penetration isolation boundary. This change is acceptable because the Completion Time is consistent with safe operation under the specified Condition, the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of a DBA occurring during the allowed Completion Time. In the case of a single penetration with an inoperable valve, 72 hours is a reasonable time period considering the relative stability of a closed system to act as a penetration Isolation boundary. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. L.5 (Category 4 - Relaxation Of Required Action) CTS 4.7.D.2 and CTS 4.7.D.3 require verification that specified containment penetrations are closed. ITS 3.6.1.3 Required Actions A.2, C.2, and D.2 include a similar requirement, but contain a Note (Note 2) that allows isolation devices that are locked, sealed, or otherwise secured may be verified by use of administrative means. This changes the CTS by allowing certain isolation devices to not require physical verification. The purpose of CTS 4.7.D.2 and CTS 4.7.D.3 is to provide assurance that primary containment penetrations are closed when necessary. This change is acceptable because it has been determined that the relaxed Required Actions are not necessary for physical verification that the equipment used to meet the ACTIONS are in the correct position. For those isolation devices that are locked, sealed, or otherwise secured, plant procedures control their operation. Therefore, the potential for inadvertent misalignment of these devices after locking, sealing, or securing is low. In addition, all the isolation devices were verified to be in the correct position (as required by ITS 3.6.1.3 Required Actions A.1, C.1, and D.1) prior to locking, sealing, or otherwise securing. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. L.6 (Category 1- Relaxation of LCO Requirements) CTS 3.7.D.3.a specifies that inerting and de-inerting operations permitted by TS 3.7.A.5.b shall be via the 18 inch purge and vent valves aligned to the Reactor Building plenum and vent, and that all other purging and venting, when primary containment integrity is required, shall be via the 2 inch purge and vent valve bypass line and the Standby Gas Treatment Systems. The ITS SR 3.6.1.3.1 Note states that the 18 inch primary containment purge and vent valves may be opened for inerting, de-inerting, pressure control, ALARA, or air quality considerations for personnel entry, or Surveillance that require the valves to be open. This changes the CTS by allowing the 18 inch containment purge and vent valves to be opened under more conditions than in the CTS. The change to the requirement that the 18 inch purge and vent valves, when opened, be aligned to the Reactor Building plenum and vent is discussed in DOC LA.3. The purpose of CTS 3.7.D.3.a is to limit the use of the 18 inch primary containment purge and vent valves. This change is acceptable because the LCO requirements continue to ensure that the structures, systems, and components are maintained consistent with the safety analyses. This change allows the Monticello Page 12 of 14 Attachment 1, Volume 11, Rev. 0, Page 70 of 431
Attachment 1, Volume 11, Rev. 0, Page 71 of 431 DISCUSSION OF CHANGES ITS 3.6.1.3, PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs) 18 inch containment purge and vent valves to be opened under more conditions than in the CTS. The proposed limits on when the purge and vent valves are permitted to be open, provided in the Note to ITS SR 3.6.1.3.1, will ensure appropriate controls. The Note will continue to allow the purge and vent valves to be open for inerting, and de-inerting, and will now allow the purge and vent valves to also be open for pressure control, ALARA, or air quality considerations for personnel entry, as well as for Surveillances that require the purge and vent valves to be open. Thus, use of the purge and vent valves will continue to be minimized and limited to safety related reasons. In addition, these valves are fully qualified to close in the required time under accident conditions to isolate the affected penetrations. This change is designated as less restrictive because less stringent LCO requirements are being applied in the ITS than were applied in the CTS. L.7 (Category 4 - Relaxation of Required Action) CTS 3.7.A.2.a.(1) includes the requirements for all "manual" PCIVs since the CTS definition of Primary Containment Integrity (1.0.P) includes these valves. If a manual valve that is supposed to be closed is open (i.e., inoperable), CTS 3.7.A.2.a.(4) applies. CTS 3.7.A.2.a.(4) states, in part, "If requirements of 3.7.A.2.a.(1) cannot be met, restore Primary Containment Integrity within one hour," or a unit shutdown is required. Thus, if one or more manual PCIVs are inoperable, 1 hour is allowed by the CTS to restore OPERABILITY. ITS 3.6.1.3 ACTIONS A, B, and C do not differentiate between automatic and manual valves and allow 1 hour, 4 hours, or 72 hours to isolate the affected penetration flow path (depending upon the number of valves inoperable in the penetration and the type of penetration), prior to requiring a unit shutdown. In addition, ITS 3.6.1.3 ACTIONS Notes 1, 2, 3, and 4 allow penetration flow paths to be unisolated intermittently under administrative controls, allow separate condition entry for each penetration flow path, require entry into the applicable Conditions and Required Actions for systems made inoperable by PCIVs, and require entry into the applicable Conditions and Required Actions for LCO 3.6.1.1, "Primary Containment," when PCIV leakage results in exceeding the overall containment leakage rate acceptance criteria. This changes the CTS by providing 4 hours (for two valve penetrations with one valve inoperable or for one valve penetrations, that are not excess flow check valve or closed system penetrations, with one valve inoperable) or 72 hours (for one valve penetrations, that are excess flow check valve or closed system penetrations, with one valve inoperable) to isolate a penetration flow path affected by an inoperable non-automatic primary containment isolation valve and continue to operate with the penetration flow path isolated. This also changes the CTS by allowing penetration flow paths to be unisolated intermittently under administrative controls, allows separate condition entry for each penetration flow path with an inoperable non-automatic PCIV, requiring entry into the applicable Conditions and Required Actions for systems made inoperable by inoperable non-automatic PCIVs, and requiring entry into the applicable Conditions and Required Actions for LCO 3.6.1.1, "Primary Containment," when leakage through a penetration flow path due to an inoperable non-automatic PCIV results in exceeding the overall containment leakage rate acceptance criteria. The purpose of the CTS 3.7.A.2.a.(4) is to ensure that overall containment leakage rate does not exceed the accident analysis assumptions. This change is Monticello Page 13 of 14 Attachment 1, Volume 11, Rev. 0, Page 71 of 431
Attachment 1, Volume 11, Rev. 0, Page 72 of 431 DISCUSSION OF CHANGES ITS 3.6.1.3, PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs) acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing time to repair inoperable features. The Required Actions are consistent with safe operation under the specified Condition, considering the operability status of the redundant systems of required features, the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of a DBA occurring during the repair period. This change makes the actions for an inoperable non-automatic PCIV consistent with the actions for all other types of PCIVs and ensures that leakage through a penetration flow path affected by an inoperable non-automatic PCIV is isolated. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. Monticello Page 14 of 14 Attachment 1, Volume 11, Rev. 0, Page 72 of 431
Attachment 1, Volume 11, Rev. 0, Page 73 of 431 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 73 of 431
Attachment 1, Volume 11, Rev. 0, Page 74 of 431 PCIVs 3.6.1.3 CT) 3.6 CONTAINMENT SYSTEMS 3.6.1.3 Primary Containment Isolation Valves (PCIVs) 3.7.D.I. 3.7.A.2.a.(l) LCO 3.6.1.3 Each PCIV, except reactor building-to-suppression chamber vacuum breakers, shall be OPERABLE. 3.7.D.I. APPLICABILITY: MODES 1, 2, and 3, 3.7.A.2.a.Cl) When associated instrumentation is required to be OPERABLE per LCO 3.3.6.1, "Primary Containment Isolation Instrumentation." ACTIONS
-NOTES.
3'7.D2 ao .
- 1. Penetration flow pathsl[except for purmeariv-He intermittently under administrative controls.
efration flow paths]j may be unisolated 0D DOC LI, DOC L7 DOC A.4, 2. Separate Condition entry is allowed for each penetration flow path. DOC L7 DOC A.5, 3. Enter applicable Conditions and Required Actions for systems made inoperable by PCIVs. DOC L7 DOC A.6A KJ DCO L7 4. Enter applicable Conditions and Required Actions of LCO 3.6.1.1, "Primary Containment," when PCIV leakage results in exceeding overall containment leakage rate acceptance criterialin MODIFY 2, and . 0( CONDITION REQUIRED ACTION COMPLETION TIME A. ---- NNOTE----- A.1 Isolate the affected 4 hours except for 3.7.D.2.a, Only applicable to penetration flow path by main steam line 4.7.D.2, penetration flow paths use of at least one closed DOC L.7 with two or ore PCIVs. and de-activated automatic valve, closed manual valve, AND 0 blind flange, or check valve 8 hours for main with flow through the valve steam line One or more penetration secured. flow paths with one PCIV inoperableafor AND reasons other than P 4e Conditiorgfl D fnf. BWR/4 STS 3.6.1.3-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 74 of 431
Attachment 1, Volume 11, Rev. 0, Page 75 of 431 PCIVs 3.6.1.3 CTS ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME A.2 ----- NOTES------- 4.7.D.2 including 1. Isolation devices in high footnote *- radiation areas may be verified by use of administrative means. I
- 2. Isolation devices that are locked, sealed, or otherwise secured may be verified by use of administrative means.
Verify the affected Once per 31 days for penetration flow path is isolation devices isolated. outside primary containment AND Prior to entering MODE 2 or 3 from MODE 4Aif primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days, for isolation devices inside primary containment BWR/4 STS 3.6.1.3-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 75 of 431
Attachment 1, Volume 11, Rev. 0, Page 76 of 431 PCIVs 3.6.1.3 CTS ACTIONS (continued) CONDITION REQUIRED ACTION j COMPLETION TIME B. -- NOTE---- B.1 Isolate the affected 1 hour 3.7.D.2.b. Only applicable to penetration flow path by DOC L.7 penetration flow paths use of at least one closed and de-activated automatic with two or ore PCIVs. valve, closed manual valve, 0 or blind flange. One or more penetration flow pa ths t I + His afor reasons other thanJJ Condition D pn
.1.
3.7.D.2.a, C. ---- NOTE---- Only applicable to C.1 Isolate the affected penetration flow path by R hours except for excess flow check 0 4.7.D.2, DOC L.7 penetration flow paths *use of at least one closed valves (EFCVs) and with only one PCIV. and de-activated automatic penetrations with a valve, closed manual valve, closed system or blind flange. One or more penetration AND flow paths with one PCIV inoperablo/r 72 hours for EFCVs
.resons oflier~an I ondition[s1 ES [and E1W.
and penetrations with a closed system 0 AND BWR/4 STS 3.6.1.3-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 76 of 431
Attachment 1, Volume 11, Rev. 0, Page 77 of 431 PCIVs 3.6.1.3 CTS ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME 4.7.D.2 C.2 ---- --NOTES----- including 1. Isolation devices in high footnote ** radiation areas may be verified by use of administrative means.
- 2. Isolation devices that are locked, sealed, or otherwise secured may be verified by use of administrative means.
Verify the affected Once per 31 days penetration flow path is isolated. I Prior to entering MODE 2 or 3 from MODE 4 If primary containment was de-Inerted while In MODE 4. if not performed within the previous 0 92 days, for Isolation devices Inside primary containment BWR/4 STS 3.6.1.3-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 77 of 431
Attachment 1, Volume 11, Rev. 0, Page 78 of 431 PCIVs 3.6.1.3 CTS ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME D. [ One or more D.1 Restore leakage rate to [4 hours for/ [secondary containment ithin limit. hydrostatic Ily tested bypass leakage rate,] line leakaoe [not on a [MSIV leakage rate,] closed sy tem]] [purge valve leakage rate,] [hydrostatically AND tested line leakage rate,] [or] [EFCV leakage rate] [4 ho rs for not within limit. sec dary co ainment bypass le age] 0 ND [8 hours for MSIV leakage] AND [24 hours for purge valve leakage] AND [72 hours for hydrostatically tested line leakage [on a closed system] [and EFCV leakage] ] _ 4 3.7.D.3.b or more penetration flow paths 1 Isolate the affected penetration flow path by 24 hours (DO with one or more use of at least oneaclosed containment purge and de-activated automatic valves not within purge / valve, closed manual valve, 0 valve leakage limits. or blind flange]. inch prmary 118 nAt AND BWR/4 STS 3.6.1.3-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 78 of 431
Attachment 1, Volume 11, Rev. 0, Page 79 of 431 PCIVs 3.6.1.3 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME
- 4.7.D.3, including l -- -- NOTES ---
- 1. Isolation devices in high 0D footnote
- radiation areas may be verified by use of administrative means.
- 2. Isolation devices that are locked, sealed, or otherwise secured may be verified by use of administrative means.
Verify the affected Once per 31 days for penetration flow path is isolation devices isolated. outside containment AND/ Prior to en ering MODE 2 r 3 from MODE if not 0 perform d within th previous 92 days for isolatioh devices insidepontainment AND E.3 Perform SR 3.6.1.3 or the resilient seal rge Once per [92] day 00 valves closed comply with Requi Action E.1. 3.7.D.4, 3.7A.2.a.(4) Required Action and. associated Completion I<3 Be in MODE 3. 12 hours 0 Time of Condition A, B. AND C,E or&~ e in MODE 1, 2, or 3. 2c iBe in MODE 4. 36 hours I 0 BWR/4 STS 3.6.1.3-6 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 79 of 431
Attachment 1, Volume 11, Rev. 0, Page 80 of 431 PCIVs 3.6.1.3 CTS ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME G. [ Required Action and G.1 -- -NOTE---- associated Completion LCO 3.0.3 is not applicable. Time of Condition A, B, C, D,or E not met for PCIV(s) required to be OPERABLE during Suspend movement of [recently] irradiated fuel mmediately ] 0 movement of [rece assemblies in [secondary irradiated fuel containment. assemblies in [secondary] containme DOC M.A if I Required Action and associated Completiop WiE Initiate action to suspend
>,OPDRVs IiV~
Immediately 00 Time of Condition loperations With a potential for or not met for OR d PCIV(s) required to be OPERABLE dunn t.2 Initiate action to restore ImmediatelymJ (D 0 MODE 4 or 51or ring valve(s) to OPERABLE op ations with status. p ential for dr ining the r actor vesse +/- 4-0
.1.
BWR/4 STS 3.6.1.3-7 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 80 of 431
Attachment 1, Volume 11, Rev. 0, Page 81 of 431 PCIVs 3.6.1.3 CTS SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6;1.3.1 -N----- [[Only requ ed to be met in MODES 1, 2, and 3.] 0) Ve each [18] inch primary containment urge 31 days] ye is sealed closed except for one p 1ge valve in a penetration flow path while in Con ion E of this
//LCO.
+
3.7.D.3.a, SR 3 .6 .1 .3 DOC M.5 EI 1. [Only required tiniMODES 1, 2, and 3.] -0
] Not required to be met when the 1 inch a3d vent primary containment purgekvalves are open for inerting, de-inerting, pressure control, ALARA or 11 air quality considerations for personnel entry, or Surveillances that require the valves to be open.
V erify V4 eachal1 8 inch primary containment purge alve is closed. 31 days] 0 4 DOC M.6, 1.0.P.1' SR 3.6.1.39 " 4 -M 1. I Valves 2.O.P.4
------ NOTES---------
and blind flanges in high radiation areas 0 may be verified by use of administrative means. 2.. Not required to be met for PCIVs that are open under administrative controls. Verify each primary containment isolation manual 31 days valve and blind flange that is located outside primary containment and not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed. BWR/4 STS 3.6.1.3-8 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 81 of 431
Attachment 1, Volume 11, Rev. 0, Page 82 of 431 PCIVs 3.6.1.3 CTS SURVEILLANCE REQUIREMENTS (continued) . SURVEILLANCE FREQUENCY DOC SR 3.6.1.3 M.6,
-NOTES-
- 1. Valves and blind flanges in high radiation areas 0
1.o.P.1. may be verified by use of administrative means. 1.O.P.4
- 2. Not required to be met for PCIVs that are open under administrative controls.
Verify each primary containment manual isolation Prior to entering valve and blind flange that is located inside primary MODE 2 or 3 from containment and not locked, sealed, or otherwise MODE 4 if secured and is required to be closed during accident primary conditions is closed. containment was de-inerted while in MODE 4, if not performed within the previous 92 days
.
Li U DOCM.4 SR 3.6.1.3[6< t Verify continuity of the traversing incore probe (TIP) shear isolation valve explosive charge. 31 days 0 BWR/4 STS 3.6.1.3-9 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 82 of 431
Attachment 1, Volume 11, Rev. 0, Page 83 of 431 PCIVs 3.6.1.3 CTS BWR/4 STS 3.6.1.3-10 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 83 of 431
Attachment 1, Volume 11, Rev. 0, Page 84 of 431 PCIVS 3.6.1.3 CTS 0 ISR 3.6.1.3.14 Z--NOTES met in MODES 1, 2, and 3.] Verify bned leakage rate through hydrostati y In accordance 0 test lines that penetrate the primary contai ent ithin limits. [ [Only required to be Verify eachIV] inch primary containment purge' valve is blocked to restrict the valve from opening 0D
>rg50°/d move to page 3.6.1.3-10 as Indicated
.
I BWR/4 STS 3.6.1.3-11 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 84 of 431
Attachment 1, Volume 11, Rev. 0, Page 85 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.1.3, PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs)
- 1. The restriction in ISTS 3.6.1.3 ACTIONS Note 1 concerning purge valves has been deleted, consistent with the current licensing basis, as modified by DOC L.9.
- 2. The words 'in MODES 1, 2, and 3" have been deleted from ITS 3.6.1.3 ACTIONS Note 4 since there are no PCIV leakage tests required in MODES other than MODES 1, 2, and 3 for Monticello (i.e., there are no PCIVs required to be OPERABLE in MODES other than MODES 1, 2, and 3 that have specific leakage limits).
- 3. The bracketed term "or more" in ISTS 3.6.1.3 Condition A Note, Condition B Note, and Condition B is not adopted. At Monticello, each penetration flow path has no more than two required valves. This is consistent with the current licensing basis.
- 4. ISTS SR 3.6.1.3.1, the primary containment purge valve sealed closed verification, ISTS SR 3.6.1.3.12, the combined leakage rate verification for all secondary containment bypass leakage paths, ISTS SR 3.6.3.13, the MSIV leakage rate verification, and ISTS SR 3.6.1.14, the combined leakage rate verification through hydrostatically tested lines, have all been deleted since they are currently not required by the CTS. The PCIVs are required to be OPERABLE such that they are in the accident condition or can be automatically repositioned to the accident condition, and certain PCIVs have individual leakage limits. These leakage limits are in addition to the type A, B, and C limits required by LCO 3.6.1.1, Primary Containment OPERABILITY. If a type A, B, or C limit were exceeded due to an individual valve exceeding its specific leakage limit, ISTS 3.6.1.3 ACTIONS Note 4 would require the ACTIONS of LCO 3.6.1.1 to be taken (which require primary containment to be restored within 1 hour). Subsequent Surveillances have been renumbered, as applicable. ISTS 3.6.1.3 ACTION D covers the condition for various types of leakage limits not met. This ACTION is also deleted because it is not currently required by the CTS. Subsequent ACTIONS have been renumbered as applicable. The ISTS 3.6.1.3 Conditions A and B words in the brackets have been modified to reflect the deletion of ISTS 3.6.1.3 ACTION D. The bracketed words in ISTS 3.6.1.3 ACTION C have been deleted since ISTS 3.6.1.3 ACTION D has been deleted and the purge valve penetrations (to which ISTS 3.6.1.3 ACTION E applies) all have two valves. In addition, ISTS 3.6.1.3 ACTION F (ISTS 3.6.1.3 ACTION E) has been modified to reflect the deletion of ISTS 3.6.1.3 ACTION D.
- 5. The brackets have been removed and the proper plant specific information/value has been provided.
- 6. A new Completion Time has been added to ITS 3.6.1.3 Required Action C.2 since these are single valve penetrations that can be isolated by a device inside primary containment. This Completion Time is identical as that for ISTS 3.6.1.3 Required Action A.2. The second Completion Time in ISTS 3.6.1.3 Required Action E.2 (ITS 3.6.1.3 Required Action D.2) has been deleted because there are no isolation devices inside the containment associated with the primary containment vent and purge valve penetration flow paths.
- 7. ISTS 3.6.1.3 Required Action E.3, the requirement to perform SR 3.6.1.3.7 for the resilient seal purge valves closed to comply with Required Action E.1 has been deleted. This allowance is consistent with the current requirements.
Monticello Page 1 of 2 Attachment 1, Volume 11, Rev. 0, Page 85 of 431
Attachment 1, Volume 11, Rev. 0, Page 86 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.1.3, PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs)
- 8. This bracketed requirement has been deleted because it is not applicable to Monticello. Monticello does not have any PCIVs whose Applicability is only during movement of irradiated fuel in the secondary containment. The following requirements have been renumbered, where applicable, to reflect this deletion.
- 9. The words in ISTS 3.6.1.3 Condition H (ITS 3.6.1.3 Condition F), "or during operations with a potential for draining the reactor vessel (OPDRVs)," have been deleted. There are no PCIVs required to be OPERABLE in the Monticello ITS whose Applicability is only during OPDRVs. The only PCIVs required when not in MODES 1, 2, and 3 are the RHR shutdown cooling isolation valves, and their Applicability is MODES 1, 2, 3, 4 and 5. This Condition is still applicable in MODES 4 and 5, which are the only MODES that OPDRVs can be performed. Therefore, the "during OPDRVs" Applicability is duplicative of the MODES 4 and 5 Applicability and has been deleted. In addition, due to this deletion the acronym "OPDRVs" in ISTS 3.6.1.3 Required Action H.1 (ITS 3.6.1.3 Required Action F.1) has been defined, consistent with the format of the ITS, since it is now the first use of this term in this Specification.
- 10. ISTS SR 3.6.1.3.2 (ITS SR 3.6.1.3.1) Note 1, ISTS 3.6.1.3.7 (ITS SR 3.6.1.3.11)
Note, and ISTS SR 3.6.1.3.15 (ITS SR 3.6.1.3.9) Note are bracketed Notes that state "Only required to be met in MODES 1, 2, and 3"and have been deleted. These Surveillance Requirements apply to the 18 inch primary containment purge and vent valves. These Notes were included in the ISTS to clarify that the Surveillances are not applicable in MODES other than MODES 1, 2, and 3 since the Applicability of "When associated instrumentation is required to be OPERABLE per LCO 3.3.6.1," could require certain PCIVs to be OPERABLE in MODES other than MODES 1, 2, and 3. The ISTS 3.6.1.3 Applicability Bases clearly states that this Applicability only applies to certain valves. At Monticello however, the only valves (i.e., the certain valves) affected by the above mentioned Applicability are valves required to be OPERABLE to prevent inadvertent reactor vessel draindown. Therefore, since the three Notes affect only purge and vent valve Surveillances, the Notes are not required.
- 11. Changes have been made to reflect the plant specific design.
- 12. The Surveillance Frequency in ISTS SR 3.6.1.3.7 (ITS 3.6.1.3.11) has been replaced with the Monticello current licensing basis Surveillance Frequency.
- 13. The Reviewers Note is deleted as it is not part of the plant specific ITS.
- 14. Typographical error corrected.
Monticello Page 2 of 2 Attachment 1, Volume 11, Rev. 0, Page 86 of 431
Attachment 1, Volume 11, Rev. 0, Page 87 of 431 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 87 of 431
Attachment 1, Volume 11, Rev. 0, Page 88 of 431 PCIVs B 3.6.1.3 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.3 Primary Containment Isolation Valves (PCIVs) BASES BACKGROUND The function of the PCIVs, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs) to within limits. Primary containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a DBA. The OPERABILITY requirements for PCIVs help ensure that an adequate primary containment boundary is maintained during and after an accident by minimizing potential paths to the environment. Therefore, the he ) OPERABILITY requirements provide assurance thatprimary containment function assumed in the safety analyses will be maintained. These isolation devices are either passive or active (automatic). Manual valves, de-activated automatic valves secured in their closed position (including heck vales with flow through the valve secured), blind flange and nce closed systems are considered passive devices. Check valves, or other asflisted In automatic valves designed to close without operator action following an accident, are considered active devices. Two barriers in series are
;except for '/provided for each penetrationrso that no single credible failure or Penetaincopnt Isolated malfunction of an active component can result in a loss of isolation or excess ow leakage that exceeds limits assumed In the safety analyses. One of check valvesthese barriers may be a closed system.
The reactor building-to-suppression chamber vacuum breakers serve a dual function, one of which is primary containment isolation. However, since the other safety function of the vacuum breakers would not be available if the normal PCIV actions were taken, the PCIV OPERABILITY requirements are not applicable to the reactor building-to-suppression mber vacuum breakers valves. Similar Sirveillance equirements in Eor Ouilding-to-Auppression Ahamber hacuum Oreakers (j) Eassurance that the isolation capability is available without conflicting with the vacuum relief function. Trimary containment purgE arell inchesin diameter; vent (I2 are jai 81] inches in diameter. The 1 8M inch primary containment (2) andve p~ur alves are normally maintained closed in MODES 1, 2, and 3 to (i ensure the primary containment boundary is maintained. The isolation valves on the Ml18E inch vent lines havea22 inch bypass lines around them dJ for use during normal reactor operation. [Two addI erredunan BWR/4 STS B 3.6.1.3-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 88 of 431
Attachment 1, Volume 11, Rev. 0, Page 89 of 431 PCIVs B 3.6.1.3 BASES BACKGROUND (continued) exce toislto aprbre provided on the vent line/lupstream oftI the tn a ramn (G-T) Svstem filter trains. Thebse isolationll dar~es tatewtth ,Cslilprvnt high pressure froml [se ofveth reaching the01S-0`T1Svstem filter trains in the unlikely event of a loss of l c an accident LOCA dunn ventin Closure of the exces flow Standby Gas isolati mpers will not preve t the SGT System from perf rming its Treatmnent (SGT) desi function (that is,to mai tain a negative pressure in tIe secondary cont inment). To ensure that/a vent path is available. a 12 inch bvrass line is rovided around the d mpers.1 N 1 APPLICABLE The PCIVs LCO was derived from the assumptions related to minimizing SAFETY the loss of reactor coolant inventory, and establishing the primary ANALYSES containment boundary during major accidents. As part of the primary containment boundary, PCIV OPERABILITY supports leak tightness of primary containment. Therefore, the safety analysis of any event requiring isolation of primary containment is applicable to this LCO. ri th The DBAs that result in a release of radioactive material lwithi ninmarl are mitigated by I conta ment are a LOCA and a main steam line break (MSLB). In the J PCvS analysis for each of these accidents, it is assumed that PCIvs are either closed or close within the required isolation times following event (Refs. 2 end 3, initiation. This ensures that potential paths to the environment throu h respectively) PCIVs (including primary containment purgervalves) are minimized.O and vent v nts analyzed in Referen e 1, the MSLB is the most Ii iting event duet radiological consequen es. The closure time of the ain steam E3ioa ion valves (MSIVs) is a~gcMnfMca variabe n from a _a _gca M l INSEITs ar reuired to close within 3 toLSeconds Wl adet te5second closiis med in the T safsS Ly The th the purgvalves et event initiation. onot make it is assumed that the primary containment is isolated such that anyexpl'ct o fission products to the environment is controlled. Seease assumptons concernn The DBA analysis assumes thatlwithin 60see e-acci-en, piorto fuel lisolation of the primary containment iscomplete and leakage is damage terminated, except for the maximum allowable leakage rate, Li'. e primary ontainment isolation total response time of sec ns includ Ke signal Rlay, diesel generator startup (for loss of offsitepkradPIo
- The single failure criterion required to be imposed in the conduct of unit safety analyses was considered in the original design of the primary containment purgefvalves. Two valves in series on each purgeviinie nd ye provide assurance that both the supply and exhaust lines could be isolated even if a single failure occurred:lj BWR/4 STS B 3.6.1.3-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 89 of 431
Attachment 1, Volume 11, Rev. 0, Page 90 of 431 B 3.6.1.3 0 INSERT I When the 18 inch purge and vent valves are opened, they must be aligned to the reactor building plenum and vent. The 18 inch purge and vent valves are capable of closing in the environment of a LOCA.
- 0) INSERT 2 The radiological consequences of the LOCA is discussed in Reference 4 while the
.4 radiological consequences of the MSLB is discussed in Reference 5. The whole body offsite dose consequences is most limiting in the LOCA while the thyroid offsite dose consequences is most limiting in the MSLB event. The control room dose consequences is most limiting in the LOCA event.
03 INSERT 3
. The 3 second closure time is assumed in the MSIV closure (the most severe overpressurization transient) analysis (Ref. 6). The 9.9 second closure time is assumed in the MSLB analysis (The analysis assumes a total time of 10.5 seconds of which 0.6 seconds is assumed for instrument response).
0 INSERT 4 However, the purge and vent valves have been designed to close prior to the onset of fuel failure following a LOCA. Insert Page B 3.6.1.3-2 Attachment 1, Volume 11, Rev. 0, Page 90 of 431
Attachment 1, Volume 11, Rev. 0, Page 91 of 431 PCIVs B 3.6.1.3 BASES APPLICABLE SAFETY ANALYSES (continued) [The primary containment puke valves may be unable to close* the environment following a LTA. Therefore, each of the purg alves is
*required to remain seal closed during MODES 1, 2, an . In this case the single failure crit on remains applicable to the pri ry containment purge valve due t ailure in the control circuit asso ted with each valve.
The primary c ainment purge valve design pre des a single failure from compr ising the primary containment b ndary as long as the system i perated in accordance with this 0. PCIVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO PCIVs form a part of the primary containment boundary. The PCIV safety function is related t minimizing the loss of reactor coolant inventory and establishing the primary containment boundary during a The power operated, automatic isolation valves are required to have isolation times within limits and actuate on an automatic isolation signal The 18finch purge valves must be maintainedIsealegd-ised [o blocikedQ to prevent full openings While the reactor building-to-suppression UG 3 chamber vacuum breakers isolate primary containment penetrations, they are excluded from this Specification. Controls on their isolation function m are adequately addressed in LCO 3.6.1.Z,-"Reactor Building-to-Suppression Chamber Vacuum Breakers." The valves covered by this LCO are listed with their associated stroke times in Reference!i The normally close PCIVs are considered OPERABLE whend valves are closed or open in accordance with appropriate administrative controls, automatic valves are de-activated and secured in their closed position, blind flanges are in place, and closed systems are intact. These 5 Z( passive isolation valves and devices are those listed in ReferenceM Purge valves with resilient'seal secondary a v aai es. M61Vs. and lhvdrostatiested valved must meet a ona leakage rate (D requirement. Other PCIV leakage rates are addressed by LCO 3.6.1.1, lttesting requirements I PrFirary Containment," as Type B or C testing. This LCO provides assurance that the PCIVs will perform their designed safety functions to minimize the loss of reactor coolant inventory and establish the primary containment boundary during accidents. BWR/4 STS B 3.6.1.3-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 91 of 431
Attachment 1, Volume 11, Rev. 0, Page 92 of 431 PCIVs B 3.6.1.3 BASES (.e.. resda APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive heatremoval material to primary containment. In MODES 4 and 5, the probability and (Ringshutdp consequences of these events are reduced due to the pressure and isolation vwves) temperature limitations of these MODES. Therefore, most PCIVs are not required to be OPERABLy and the rimary containment purhe valves are Inot required to be sealed closedlin MODES 4 and 5. Certain valve e however, are required to be OPERABLE to prevent inadvertent reactor vessel draindown. These valves are those whose associated instrumentation is required to be OPERABLE per LCO 3.3.6.1, "Primary Containment Isolation Instrumentation." (This does not include the valves that isolate the associated instrumentation.) ACTIONS The ACTIONS are modified by a Note allowing penetration flow path(s) [except for p ye flow path(s)] to be unisolated intermittently under administrative controls. These controls consist of stationing a dedicated at the controls of the valve, who is in continuous communication with the control room. In this way, the penetration can be rapidly isolated when a need for primary containment isolation is indicated. Due to, he size of/he primary containment purge link penetration and the fa that those penetrations exhaust directly from he containment atmos ere to the Environment, the penetration flow p th containing these va es is not 0D allo ed to be opened under administr ive controls. A single urge valve in penetration flow path may be ope ed to effect re airs t an perable valve, as allowed by SR .6.1.3.1. A second Note has been added to provide clarification that, for the purpose of this LCO, separate Condition entry is allowed for each penetration flow path. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable PCIV. Complying with the Required Actions may allow for continued operation, and subsequent Inoperable PCIVs are governed by subsequent Condition entry and application of associated Required Actions. The ACTIONS are modified by Notes 3 and 4. Note 3 ensures that appropriate remedial actions are taken, if necessary, if the affected system(s) are rendered inoperable by an inoperable PCIV (e.g., an Emergency Core Cooling System subsystem is inoperable due to a failed open test return valve). Note 4 ensures appropriate remedial actions are taken when the primary containment leakage limits are exceeded. Pursuant to LCO 3.0.6, these actions are not required even when the associated LCO is not met. Therefore, Notes 3 and 4 are added to require the proper actions be taken. BWR/4 STS B 3.6.1.3-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 92 of 431
Attachment 1, Volume 11, Rev. 0, Page 93 of 431 PCIVs B 3.6.1.3 BASES ACTIONS (continued) A.1 and A.2 With one or more penetration flow paths with one PCIV inoperable, lexcept forfsecondary containment eafassTaae rate, MSIV leakaeel andlvent purgevalve leakage ratd or hydrostati leakace rate 'KXI mALteine lor EFC aae rate not within limiq, the affected penetration flow paths must be isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-valve means the valve Is activated automatic valve, a closed manual valve, a blind flange, and a either electrically or check valve with flow through the .valve secured For a penetration (j) pneumatically disarmed or - otherwise secured In the isolated in accordance with Require Action A. , the device used to closed position, isolate the penetration should be the closest available valve to the primary containment. The Required Action must be completed within the 4 hour Completion Time (8 hours for main steam lines). The Completion Time of 4 hours is reasonable considering the time required to isolate the penetration and the relative importance of supporting primary containment OPERABILITY during MODES 1, 2, and 3. For main steam lines, an 8 hour Completion Time is allowed. The Completion Time of 8 hours for the main steam lines allows a period of time to restore the MSIVs to OPERABLE status given the fact that MSIV closure will result in isolation of the main steam line(s) and a potential for plant shutdown. For affected penetrations that have been isolated in accordance with Required Action A.1, the affected penetration flow path(s) must be verified to be isolated on a periodic basis. This is necessary to ensure that primary containment penetrations required to be isolated following an accident, and no longer capable of being automatically isolated, will be in the isolation position should an event occur. This Required Action does not require any testing or device manipulation. Rather, it involves that those devices pnmryoutidcontainment [ Y1verification oti adcapable pbeof c onametand e 0f potentially being mispositioned are in the correct position. The Completion Time of "once per 31 days for isolation devices outside primary containment" is appropriate because the devices are operated under administrative controls and the probability of their misalignment is low. For the devices inside primary containment, the time period specified "prior to entering MODE 2 or 3 from MODE 40if primary (i) containment was de-inerted while in MODE 4, if not performed within the previous 92 days" is based on engineering judgment and is considered reasonable in view of the inaccessibility of the devices and other administrative controls ensuring that device misalignment is an unlikely possibility. BWR/4 STS B 3.6.1.3-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 93 of 431
Attachment 1, Volume 11, Rev. 0, Page 94 of 431 PCIVs B 3.6.1.3 BASES ACTIONS (continued) Condition A is modified by a Note indicating that this Condition is only applicable to those penetration flow paths with two or ore PCIVs. For 0 penetration flow paths with one PCIV, Condition C provides the appropriate Required Actions. Required Action A.2 is modified by two Notes. Note 1 applies to isolation devices located in high radiation areasS and allows them to be verified by ( use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted. Note 2 applies to isolation devices that are locked, sealed, or otherwise secured in position and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since the function of locking, sealing, or securing components is to ensure that these devices are not inadvertently repositioned. Therefore, the probability of misalignmentf the eviceN, once they have been verified to be in the E5 proper position, is low. B.1 With one or more penetration flow paths with two [or ore] PCIVs inoperable, xcepfor secondary contajomeilbMass leakage rate, y1D purg valve leakage rate! or hvdro05 EaTv tested lin[ Ileakage rat leakage ra e not within limit, either the inoperable PCIVs must be restored to OPERABLE status or the affected penetration flow path must be isolated within 1 hour. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this Ade-activated automatic' criterion are a closed and de-activated automatic valve, a closed manual valve means the valve Is valve, and a blind flange. The 1 hour Completion Time is consistent with Pneumatically disarmedor the ACTIONS of LCO 3.6.1.1. othervise secured In the Condition B is modified by a Note indicating this Condition is only closed position,. applicable to penetration flow paths with two PCIVs. For 0 penetration flow paths with one PCIV, Condition C provides the appropriate Required Actions. BWR/4 STS B 3.6.1.3-6 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 94 of 431
Attachment 1, Volume 11, Rev. 0, Page 95 of 431 PCIVs B 3.6.1.3 BASES ACTIONS (continued) C.1 and C.2 and With one or more penetration flow paths with one PCIV inoperable, Dexcept for secondarv containmen aae rate, eaMae pur alve leakage ratd or hvdrostatica h one leakage rate 1\ jor U-C-eaRage rate! not within Iimitj the inoperable valve must.be restored to OPERABLE status or the affected penetration flow path must be isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active A de-activated auto valve means the va Iva ' failure. Isolation barriers that meet this criterion are a closed and de-either electrically or r pneumatically disar nmedor otherwise secured Inthe) activated
,
automatic valve, . a closed
.
check valve may not be used to isolate the affected manual
..
valve,
...
and a blind
..
flange.,jA penetration. 0 closed position.
-------------- ~W--R ER'S NOTE---------
The [4] hour Completion Ti is left as 4 hours consistent wit e Completion Time of Re *ed Action A.1 for most penetrati 5; or a plant specific evaluation is ovided for NRC review for cases her than for 05 closed system pen rations and EFCVs (which have en reviewed and approved for 72 ours). If all penetrations are acc ted for 72 hours, the Completion ne is simplified to state 72 hours for valves other than EFCVs and penetrations with a closest system I. The Completion Time of 41ihours reasonable considering the time (i (E) required to isolate the penetration and the relative importance of supporting primary containment OPERABILITY during MODES 1, 2, and 3. The Completion Time of 72 hours for penetrations with a closed system is reasonable considering the relative stability of the closed system (hence, reliability) to act as a penetration isolation boundary and the relative importance of supporting primary containment OPERABILITY during MODES 1, 2, and 3. IThe closed svsteriin1ustmeth requ~irementference 5. The Completion Time of 72 hours fork EFCVs is also reasonable considering the instrument and the small pipe diameter of penetration (hence, reliability) to act as a penetration isolation boundary and the small pipe diameter of the affected penetrations. In the event the affected penetration flow path is isolated in accordance with Required Action C.1, the affected penetration must be verified to be isolated on a periodic basis. This is necessary to ensure that primary containment penetrations required to be isolated following an accident are isola1te~d_.>The Completion Time of once per 31 daysl for veriw ac I affected peantffo-n is isoae iapropriate because the~ r operated under administrative controls and the probability of their dc misalignment is low. This Required Action does not require any testing or valve manipulation. Rather, It Involves verificatIon that these devices outside containment capable of potentially being mispositioned are In the correct position. BWR/4 STS B 3.6.1.3-7 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 95 of 431
Attachment 1, Volume 11, Rev. 0, Page 96 of 431 B 3.6.1.3 Q INSERT 5 A closed system penetrates the primary containment, but does not communicate with the reactor vessel or with the containment free space. Insert Page B 3.6.1.3-7 Attachment 1, Volume 11, Rev. 0, Page 96 of 431
Attachment 1, Volume 11, Rev. 0, Page 97 of 431 PCIVs B 3.6.1.3 BASES ACTIONS (continued) Condition C is modified by a Note indicating that this Condition is only applicable to penetration flow paths with only one PCIV. For penetration flow paths with two [or ore] PCIVs, Conditions A and B provide the appropriate Required Actions. This Note Is 0 Lspecifically necessary since this Conditon Iswritten to address those penetrations with a single PCIV. 0 [isolaton Reuired Action C.2 is modified by two Notes. Note 1 applies tolva~es] and b Ian e located in high radiation areas and allows them to be 0D verified by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted. Note 2 applies to isolation devices that are locked, sealed, or otherwise secured in position and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since the function of locking, sealing, or securing components is to ensure that these devices are not inadvertently repositioned. Therefore, the probability of misalignment of thes valvea, once they have been verified to be in the 0 proper position, is low. [D.1 With the [secondary containm nt bypass leakage rate (SR 3.6.1 .12),] [MSIV leakage rate (SR 3.6.1 3.13),] [purge valve leakage rate (SR 3.6.1.3.7),] [or] [hydrost tically tested line leakage rate (SR 3.6.1.3.14),] [or] [EFCV leakage rate (SR 3.6.1.3.1.0)] not ithin limit, the assumptions of the saf ty analysis may not be met. The fore, the leakage must be restored/o within limit. Restoration can b accomplished by isolatin the penetration that caused the Imit to be exceeded by use of on closed and de-activated automati valve, closed manual valve, or blind ange. When a penetration is isol ted, the leakage rate for the is lated penetration is assumed to the actual pathway leakage thr ugh the isolation device. If two is lation devices are 0 used to isolate the netration, the leakage rate is ass med to be the lesser actual path y leakage of the two devices. Th 4 hour Completion Time for hydrosta tcally tested line leakage [not on a/closed system] and for secondary co tainment bypass leakage is reas able considering the time required to estore the leakage by isolating th penetration and the relative import ce of secondary containment by ass leakage to the overall contai ent function. For MSIV leakage an 8 hour Completion Time is allow d. The Completion Time of 8 hous for MSIV leakage allows a per' d of time to restore the MSIVs toOPERABLE status given the fact the SIV closure will result in isolation of the main steam line(s) and poten al for plant shutdown. [The 24 hofir Completion Time for BWRI4 STS B 3.6.1.3-8 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 97 of 431
Attachment 1, Volume 11, Rev. 0, Page 98 of 431 PCIVs B 3.6.1.3 BASES ACTIONS (continued) purge valve leakage is acceptable considering the purge valves re in closed so that a gross breach of e containment does not exist.] fT e 72 hour Completion Time for hydrostatically tested line leakage [oa closed system] is acceptable ba ed on the available water seal ex ected to remain as a gaseous fission roduct boundary during the accid nt, and the associated closed system. [The 72 hour Completion Time fo EFCV leakage is acceptable based the Instrument and the small pip diameter of the penetration (ence, reliability) to act as a penetr ion isolation boundary.] VIEWER'S NOTE--- The bracketed options pro ded in ACTION D reflect options i plant design and options in ado ing the associated leakage rate S rveillances. 0e The options (both in ACTIN D and ACTION E) for purge v lve leakage, are based primarily on th design. If leakage rates can be easured separately for each purg valve, ACTION E is intended to ply. This would be required to be ble to implement Required Actio E.3. Should the design allow only fo leak testing both purge valves situltaneously, then the Completion Ti e for ACTION D should include the "24 hours for purge valve leakage" nd ACTION E should be eliminat The option for EFCV s based on the acceptance criteri of SR 3.6.1.3.10. If the acceptance cri Ha is a specific leakage rate (e.g 1 gph) then the Completion Time fo ACTION D should include the "7 hours for EFCV leakage." If the ac eptance criteria for SR 3.6.1.3.10 non-specific (e.g.,
"actuates to the cl ed position") then there is no spe ific leakage criteria and the EFCV Co pletion Time is not adopted.
Similarly, adoptin Completion Times for secondary containment bypass and/or hydrostat Ily tested lines is based on whet er the associated SRs are adopte The additional racketed options for whether the ydrostatically tested line is with or ithout a closed system is predicat d on plant-specific design. If the esign is such that there are not b th types of hydrostaticall tested lines (some with and som without closed systems), the specific' losed system' wording can be re ved and the appropriate 4 or 72 hour ompletion Time retained. In the vent there are both types, the clarifyin wording remains and the bracket are removed. ] BWR/4 STS B 3.6.1.3-9 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 98 of 431
Attachment 1, Volume 11, Rev. 0, Page 99 of 431 PCIVs B 3.6.1.3 BASES ACTIONS (continued) i hpmary p1 lp In the event one or mord containment purg valves are not within the purg va ye eakage limits, purg alve leakage must be restored to 0 within limits or the affected penetration must be isolated. The method of isolation must be by the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a Mclosed and de-activated automatic valve, closed 0 manual valve, and blind flange]. If a purge valve with resilient seals is utilized to satisfy Required Action E.1, it must have been demonstrated to meet the leakage requirements of SR 3.6.1.3' -The specified Completion Time is reasonable, considering that one containment purge;*-EH valve remains closed so that a gross breach of, containment does not 0 exist. Fpnma In accordance with Required Action E.2, this penetration flow path must be verified to be isolated on a periodic basis. The periodic verification is necessary to ensure that containment penetrations required to be isolated following an accident, which are no longer capable of being automatically isolated, will be in the isolation position should an event occur. This The Completion Time of Required Action does not require any testing or valve manipulation.
'once per 31 days for Rather, it involves verification that those isolation devices outside Isolation devices outside pnriarycontainment Is containment and potentially capable of being mispositioned are in the appropriate because the correct positionjl For the isolation devices/inside containment, the/fime undervadministrative perio specified as "prior to entering MOPE 2 or 3 from MODE 4i1f not conrolsabidtyothei perf ted within the previous 92 days@@j/ based on engineerin/judgmentl Misalignment Islow. aans considered reasonable in view f the inaccessibility of We isolation
- , de ices and other administrative contrpIs that will ensure that isolation device misalignment is an unlikely po/tibility.I For the containment purge val with resilient seal that is isolat in accordance with Required A ion E.1, SR 3.6.1.3.7 must be p rformed at least once every [ ] days. his provides assurance that de adation of 0 the resilient seal is dete ed and confirms that the leakag rate of the containment purge val e does not increase during the t.e the penetration is isolatd. The normal Frequency for S .6.1.3.7 is 184 days. Since re reliance is placed on a singl valve while in this Condition, it is p dent to perform the SR more ofn. Therefore, a Frequency of ce per [ ] days was chosen an as been shown to be acceptable , sed on operating experience.
BWR/4 STS B 3.6.1.3-10 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 99 of 431
Attachment 1, Volume 11, Rev. 0, Page 100 of 431 PCIVs B 3.6.1.3 BASES ACTIONS (continued) Required tion.2 is modified by two Notes. Note 1 applies to isolation devices located in high radiation areas and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted. Note 2 applies to isolation devices that are locked, sealed, or otherwise secured in position and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since the function of locking, sealing, or securing components is to ensure that these devices are not inadvertently repositioned. m If any Required Action and associated Completion Time cannot be met in MODE 1, 2, or 3, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. or PCIV(s) required
~1MODE 4 orOEALIn0 If any Required Action and associated Completion Time cannot be me, the unit must be placed in a condition in which the LCO does not apply. it applicable, movement of [recently] assemblies ms be Iraitdfe immej tely suspended. Suspensio o hs civities shall nt reclude comp etion of movement of a to afsaecniin ononri lf applicable,l0c-tion must be immediately initiated to suspend operations e with a potential for draining the reactor vessel (OPDRVs) to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended land valve(s) are-res do- E statuo. If suspending an OPDRV would result in closing thel residual removal 4RHIr E shutown
~ coorin isolation valves, an alternative Required Action is (0 provided to immediately initiate action to restore the valve(s) to OPERABLE status. This allows RHR~o remain in service while actions 0 are being taken to restore the valve. 0
\>(n) 3 BWR/4 STS B 3.6.1.3-11 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 100 of 431
Attachment 1, Volume 11, Rev. 0, Page 101 of 431 PCIVs B 3.6.1.3 BASES SURVEILLANCE [ SR 3.6.1.3.1/ REQUIREMENTS Each [18] inch primary contain ent purge valve is required to be erified sealed closed at 31 day interv Is. This SR is designed to ensur that a gross breach of primary cont nment is not caused by an inadve ent or spurious opening of a prima containment purge valve. Detailed analysis of the purge valves iled to conclusively demonstrate heir ability to close during a LOCA in ti e to limit offsite doses. Primary ontainment purge valves that are seale closed must have motive power o the valve operator removed. This be accomplished by de-energizi g the source of electric power o removing the air supply to the val e operator. In this application, the ter "sealed" has no connotation of Ilak tightness. The 31 day Frequency is result of an NRC initiative, Gen nc Issue B-24 (Ref. 4) related to prima containment purge valve use du ing unit operations. 0 This SR allows a valve that is open under administrative ontrols to not meet the SR during th time the valve is open. Openinga purge valve under administrative ontrols is restricted to one valve i a penetration flow path at a given t. e (refer to discussion for Note 1 f the ACTIONS) in order to effect rep irs to that valve. This allows one urge valve to be opened without res Iting in a failure of the Surveillanc and resultant entry into the ACTI NS for this purge valve, provided the stated restrictions are me}. Condition E must be entered d ing this allowance, and the valve ope ed only as necessary for effectin repairs. Each purge valve in the pene ation flow path may be altematel opened, provided one remains sea ed closed, if necessary, to comple repairs on the penetration. [The SR is mod fled by a Note stating that primary containment purge valves are onl required to be sealed closed in M DES 1, 2, and 3. If a LOCA inside rimary containment occurs in the e MODES, the purge valves may n t be capable of closing before th pressure pulse affects systems do nstream of the purge valves or th release of radioactive material will exceed limits prior to the closing the purge valves. At other times hen the purge valves are requir to be capable of closing (e.g., dunn handling of [recently] irradiated el), pressurization concerns are not pr sent and the purge valves are all ed to be open. ] ] BWR/4 STS B 3.6.1.3-12 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 101 of 431
Attachment 1, Volume 11, Rev. 0, Page 102 of 431 PCIVs B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS (continued) RSO la..3l Incho This SR ensures that th primary containment purg alves are closed as required or, if open, open for an allowable reason. If a purg Orven open in violation of this SR, the valve is considered inoperable. If the inoperable valve is not otherwise known to have excessive leakage when closed, it is not considered to have leakage outside of limits. [The R is also m ie by a Note (Note 1), stating at primary containmentfurge valves/are only required to be closed in ODES 1, 2, and 3. If a OCA 5 insideprimary containment occurs in th se MODES, the purge CaIves may ot be capable of closing before the pressure pulse affect systems dow stream of the purge valves, or th release of radioactive aterial will exc ed limits prior to the purge valve closing. At other time when the pu ge valves are required to be cap ble of closing (e.g., durnng handling of rradiated fuel), pressurization co cerns are not presentbnd the purge v Ives are allowed to be open.] he SR is modified by a Note stating that the SR is not required to be met when the purge vav open for the stated reasons. The Note states that these valves may Whent 18 nch opened for inerting, de-inerting, pressure control, ALARA or air quali and venthalves are considerations for personnel entry, or Surveillances that require the a ve opened they must be aligned to th~e vaves o eopen. The M11 Einch purgevalves are capable of closin reactorpugcpal building plenum and the environment following a LOCA. There Hore, tese valves are allowed vent to be open for limited periods of time. The 31 day Frequency is consistent with other PCIV requirements discussed in SR 3.6.1.3 SR 3.6.1.3 ( This SR verifies that each primary containment isolation manual valve and blind flange that is located outside primary containment and not locked, sealed, or otherwise secured and is required to be closed during g accident gonditions is closed. The SR helps to ensure that post accident (D leakage of radioactive fluids or gases outside the primary containment boundary is within design limits. This SR does not require any testing or valve manipulation. Rather, it involves verification that those PCIVs outside primary containment, and capable of being mispositioned, are in the correct position. Since verification oflv position for PCIVs outside primary containment is 0 relatively easy, the 31 day Frequency was chosen to provide added assurance that the PCIVs are in the correct positions. This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing. BWR/4 STS B 3.6.1.3-13 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 102 of 431
Attachment 1, Volume 11, Rev. 0, Page 103 of 431 PCIVs B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS (continued) Two Notes have been added to this SR. The first Note allows valves and blind flanges located in high radiation areas to be verified by use of administrative controls. Allowing verification by administrative controls is considered acceptable since the pinmay c enait is inerted an access to these areas is typically restrictediduring M-E1, 2, and for ALARA reasons. Therefore, the probability of misalignment of these PCIVs, once they have been verified to be in the proper position, is low. A second Note has been included to clarify that PCIV are open under administrative controls are not required to meet the SR during the time that the PCIVs are open. A These controls consist of stationing a dedicated Individual at the controls of the valve, who Is in continuous communication with the control room. In this way, the penetration can be rapidly Isolated when a need for primary containment Isolation Is Indicated. This SR verifies that each primary containment manual isolation valve and blind flange that is located inside primary containment and not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside the primary containment GP boundary iswithin design limits. For PCIVs inside primary containment, the Frequency dehRd a "prior to entering MODE 2 or 3 from MODE 4 if primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days" is appropriate since these PCIVs are operated under administrative controls and the probability of their misalignment is low. This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing. Two Notes have been added to this SR. The first Note allows valves and blind flanges located in high radiation areas to be verified by use of administrative controls. Allowing verification by administrative controls is considered acceptable since the primary containment is inerted and access to these areas is typically restricted during MODES 1, 2, and 3 for ALARA reasons. Therefore, the probability of misalignment of these PCIVs, once they have been verified to be in their proper position, is low. A second Note has been included to clarify that PCIVs that are open under administrative controls are not required to meet the SR during the time that the PCIVs are open. These controls consist of stationing a dedicated Individual at the controls of the valve. who Is In continuous communication with the control room. In this way. the penetration can be 1rapidly Isolated when a need for primary containment Isolation Is Indicated. BWR/4 STS B 3.6.1.3-14 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 103 of 431
Attachment 1, Volume 11, Rev. 0, Page 104 of 431 PCIVs B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS (continued) SR 3613T4 The traversing incore probe (TIP) shear isolation valves are actuated by explosive charges. Surveillance of explosive charge continuity provides assurance that TIP valves will actuate when required. Other administrative controls, such as those that limit the shelf life of the explosive charges, must be followed. The 31 day Frequency is based on operating experience that has demonstrated the reliability of the explosive charge continuity. SR0 Verifying the isolation time of each power operated, automatic PCIV is within limits is required to demonstrate OPERABILITY. MSIVs may be excluded from this SR since MSIV full closure isolation time is IE- demonstrated by SR 3.6.1.3:. The isolation time test ensures that the valve will isolate in a time period less than or equal to that assumed in the 0 safety analyses. Thelisolati5m-eand Frequency of this aR/re [[in I 3 acodac wt>fi eqieents of the Tetp~ Steric ram orl SR 361.. Lthe 18 Inch< consistent with or~primary containment purge yalves with resilient seals,[ gnatJ and vent K) leakaalbeheA~iirem nts of 10 MF 50, ' I E Appendix J, Option Ref. ,is required to ensure OPERABILITY.
, F >Operating experience has demonstrated thaMistype of seal has the I of this SR Is In potential to degrade in a shorter time p d than do other seal types.
arwordanoe Based on this observation and the imfortance of maintaining this@ Pnrnary penetration leak tight (due topteldrect path between primary containment
/and the environment), a Feuency of 184 days was established.
Additionally, this SR must be performed on within 92 days after opening the valve. The 92 day Freque: was chosen recognizing that cycling the valve could introduce adonal seal degradation (beyond that which occurs to a valve that has,Rt been opened). Thus, decreasing the interval (from 184 days) is a dent measure after a valve has been opened. move to page 8 3.6.1.3-18 } BWR/4 STS B 3.6.1.3-15 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 104 of 431
Attachment 1, Volume 11, Rev. 0, Page 105 of 431 PCIVs B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS (continued) The SR is modified by a te stating that the primary containm purge valves are only require o meet leakage rate testing require nts in MODES 1, 2, and 3. a LOCA inside primary containmen occurs in these MODES, pu e valve leakage must be minimized ensure offsite radiological rele se is within limits. At other times wh the purge valves are required be capable of closing (e.g., during ndling of [recently] irradiated fel), pressurization concerns are not lesent and the purge valves not required to meet any specific le age criteria.] SR3 Verifying that the isolation time of each MSIV is within the specified limits is required to demonstrate OPERABILITY. The isolation time test ensures that the MSIV will isolate in a time period that does not exceed hnd the transient times assumed in the DBA analyses. This ensures that the calculated K) radiological consequences of these events remain within 10 CFR 100 limits. The Frequency of this SR is [in accordancet -Ie requirements Iot the lnservi hng Program orlE onths Automatic PCIVs close on a primary containment isolation signal to prevent leakage of radioactive material from primary containment following a DBA. This SR ensures that each automatic PCIV will actuate Lsolnosition on a prinary containment isolation signal. The overlaps this SR to (i) provide complete testing of the safety function. Theaf iFiff(! J Frequency was developed considering it is prudent that this Surveillance be performed only during a unit outage since isolation of penetrations would eliminate cooling water flow and disrupt the normal operation of many critical components. Operating experience has shown that these 4 components usually pass this Surveillance when performed at the month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. BWR/4 STS B 3.6.1.3-16 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 105 of 431
Attachment 1, Volume 11, Rev. 0, Page 106 of 431 PCIVs B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS (continued) SR 3.6.1.3. 0
- - - --- REV WER'S NOTE-----------
The Surveillance is only allow d for those plants for which NEDO 32977-A, "Excess Flow Che Valve Testing Relaxation," June 000, is applicable. In addition, thlicensee must develop EFCV perf mance criteria and basis to ensee that their corrective action progr m can provide meaningful fee ack for appropriate corrective actins. The EFCV performance c teria and basis must be found accptable by the technical staff. If re uired, an Inservice Testing Progra relief request pursuant to 10 CF 50.55a needs to be approved by e Technical Staff in order to imple ent this Surveillance. Otherwise, ach EFCV shall be verified to act te on an [18] month Frequency. TKe bracketed portions of these Bas s apply to the representative samp as discussed in NEDO-329 -A. This SR requires a demonstration that each Ia represeae-sample of reactor instrumentation line excess flow check valva (EFCV) is ,_ ( OPERABLE by verifying that the valveareduces flow to 51L Wona )(2JOY) simulated instrument line breal ./[The represlntative sample consists of an leapsproximately equal number of EKCVs, s that each EFCV is tested ch at least on e every 10 years (nominal). In a dition, the EFCVs in the sample ar~ representative of the various piah configurations, models, sizes and/operating environments. This en ures that any potentially common problem with a specific type or a lication of EFCV is detected at the earliest possible time.] I This SR provides assurance that the instrumentation line EFCVs will was 1 eror~s tatPredce adiological consequences will rjft be exceeded< J~.Iurag th pot~lted instrument line break event evaluated in Th?,,M month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass this Surveillance when performed at the6[ month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint [TRhe nominal 130 yearnterval is baked On performance testing as dis ussed In ND-32977-8,I "Exctss ~Flow Check Valve Testing Rlbain"Furthermore, iny EFCV failue ,ilbe evaluated to determine/if additional testing in tha test intq val is warranted to ensure oera~l reliability is maintaind Operating experrience has demonstrated tohat these components are hI hly reliable ard ha filures to isolate are very fifrequent. Therefore, ieting of a rersnaie sample was concluded to be acetbe reliailt
-rma BWR/4 STS B 3.6.1.3-17 Rev. 3.0, 03131/04 Attachment 1, Volume 11, Rev. 0, Page 106 of 431
Attachment 1, Volume 11, Rev. 0, Page 107 of 431 PCIVs B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS (continued) 0 {INSERT SR 3.6.1.3.9 from pages B3.6.1.3-19 and B 3.6.1.3-20
') / SR 3.6.1.3.f The TIP shear isolation valves are actuated by explosive charges. An in place functional test is not possible with this design. The explosive squib is removed and tested to provide assurance that the valves will actuate when required. The replacement charge for the explosive squib shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of the batch successfully fired. The EFr months on a STAGGERED TEST BASIS is considered adequate given the administrative controls on replacement charges and the frequent checks of circuit continuity (SR 3.6.1.3 0
{INSERT SR 3.6.1.3.11 from page B 3.6.1.3-15 SR 3.6.1.3.12 This SR ensures that the leak ge rate of secondary containmen bypass leakage paths is less than th specified leakage rate. This pro ides assurance that the assumpt ns in the radiological evaluations of Reference 7 are met. The eakage rate of each bypass leak e path is assumed to be the maxim m pathway leakage (leakage thro gh the worse of the two isolatio valves) unless the penetration is isolated by use of one closed and d -activated automatic valve, close manual valve, or blind flange. In this se, the leakage rate of the isolat d bypass leakage path is assum d to be the actual pathway leaka e through the 0D isolation device. If bo h isolation valves in the penetrati are closed, the actual leakage rate i the lesser leakage rate of the tw valves. The Frequency is requir d by the Primary Containment Le kage Rate Testing Program. This SR imply imposes additional accept ce criteria. [This SR is modified by Note that states that these valve are only required to meet this leakag limit in MODES 1, 2, and 3. In th other conditions, the Reactor Coolan System is not pressurized and sp cific primary containment le age limits are not required. ] [Bypass leak e is considered part of La. 2 - ---- REVIEWER'S NOT Unless sp ifically exempted.] ]
/ I I
BWR/4 STS B 3.6.1.3-18 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 107 of 431
Attachment 1, Volume 11, Rev. 0, Page 108 of 431 PCIVs B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS (continued) SR 3.6.1.3.13 The analyses in References 1 and 7 are based on leakage that i less than the specified leakage rage. Leakage through each MSIV m st be s [11.5] scfh when tested at Pt ([28.8] psig). A Note is added o this SR which states that these val s are only required to meet this le kage limit in MODES 1, 2, and 3. In e other conditions, the Reactor C olant System is not pressurized and specific primary containment I akage limits are not required. This e ures that MSIV leakage is properl accounted for in determining the ov ral1 primary containment leakage rote. The Frequency is required b the Primary Containment Leakag Rate Testing Program. SR 3.6.1.3.14 Surveillance of hyd ostatically tested lines provides as urance that the calculation assumptions of Reference 2 are met. The/acceptance criteria for the combined akage of all hydrostatically testedjines is [1.0 gpm times the total n ber of hydrostatically tested PCIV ] when tested at 1.1 P. ([63.25] prig). The combined leakage rates ust be demonstrated in accordance ith the leakage rate test Frequenc required by the Primary Contaj ment Leakage Rate Testing Progr m. [This SR has een modified by a Note that states that these valves are only require to meet the combined leakage rat in MODES 1, 2, and 3, since this is when the Reactor Coolant System s pressurized and primary containme t is required. In some instances, t e valves are required to be capable otautomatically closing during MOD S other than MODES 1, 2, and 3. However, specific leakage limits are ot applicable in these other MODES r conditions.] ( W~SR 3.6.1.3.i0 (
------- R IEWER'S NOTE-----A--
This SR is only require r those plants with purge valves ' resilient seals allowed to be en during [MODE 1, 2, 3, or 4] a aving blocking ( devices that arot permanently installed on the v es. move to page B 3.6.1-1 } BWR/4 STS B 3.6.1.3-19 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 108 of 431
Attachment 1, Volume 11, Rev. 0, Page 109 of 431 PCIVs B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS (continued) {moN retopages3.6.1.3-18 Verifying each jjtnch primary containment purge alve is blocked to restrict opening to s i5s0required to ensure that the valves can close Junder DBA conditionsiwithmahetires assumie-d -inthe analysis of It (i) 2 E0 HiQ-R nces 1 and . [The SR is modified by a Note statinA that this SR is only required to bmin MODES 1, 2, and 3.] If a LOCA occurs, the purge valves m et close to maintain containment leakag within the values assumed n the accident analysis. Atote tme when purge 1 valves are re ired to be capable of closing (e.g., dun g movement of irradiated fu tssemblies), pressu--at-n oncernsatnoprst.hu the purge v ves can be fully open.l The month Frequency is , appropriate because the blocking devices are typically removed only during a refueling outage. REFERENCES I1. FSAP- ti-apter [15]. 0 AR, TableeZA"Z 0D0D 10 CFR 50, Appendix J, Option [M]. 0 0 0
- 5. ESAA Section6..[ ].
- 6. F R, Sectio 15.1.39]. 0
- 7. SAR, Sect n [6.2].
- 2. USAR, Section 14.72.
- 3. USAR. Section 14.7.3.
- 4. USAR. Section 14.7.2.4. 0
- 5. USAR Section 14.7.3.2.
- 6. USAR Section 14.5.1.
. USAR. Table 5.2-3b.
BWR14 STS B 3.6.1.3-20 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 109 of 431
Attachment 1, Volume 11, Rev. 0, Page 110 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.1.3 BASES, PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs)
- 1. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the plant specific nomenclature, number, reference, system description, analysis description, or licensing basis description.
- 2. The brackets have been removed and the proper plant specific information/value has been provided.
- 3. This paragraph in the Applicable Safety Analyses Section of Bases 3.6.1.3 has been modified since it is incorrect; the DBA analysis does not have a specific assumption for closure time of PCIVs. The analysis assumes the valves will close prior to fuel damage, which is not expected for some time. The closure times of the principle PCIVs are currently specified in the USAR, and are based upon such factors as valve size and valve operator capability. lri addition, the words in ISTS SR 3.6.1.3.6 (ITS SR 3.6.1.3.5) stating that the isolation times are in the IST Program have also been deleted since these times are also located in the USAR.
- 4. This bracketed requirement/information has been deleted because it is not applicable to Monticello.
- 5. Changes have been made to reflect those changes made to the Specification.
- 6. These changes have been made for consistency with similar phrases in other parts of the Bases and/or to be consistent with the Specification.
- 7. Editorial change made for enhanced clarity.
- 8. Typographical/grammatical error corrected.
- 9. Change made to be consistent with the identical phrase in another Bases (i.e.,
Bases 3.6.4.2, Required Actions A.1 and A.2).
- 10. The Reviewers Note is deleted as it is not part of the plant-specific ITS.
- 11. Changes have been made to be consistent with the Specification.
Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 110 of 431
Attachment 1, Volume 11, Rev. 0, Page 111 of 431 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 11, Rev. 0, Page 111 of 431
Attachment 1, Volume 11, Rev. 0, Page 112 of 431 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.6.1.3, PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs) There are no specific NSHC discussions for this Specification. Monticello Page 1 of I Attachment 1, Volume 11, Rev. 0, Page 112 of 431
( C C Monticello ITS Conversion Project Open Items By ITS Section/Specification ITS/CTS 0I # Initiator Revision Description Status Assigned to: Due Date 3.6.1.3 148 Jones ITS 3.6.1.3 requires some PCIVs to be OPERABLE In MODES Tom Demitrack 12/20/2004 other than MODES 1, 2, and 3. It Identifies these PCIVs by stating that they are required when their associated Instrumentation Is required by ITS 3.3.6.1. The Core Team wanted the valves to be Identified In the ITS 3.6.1.3 Applicability section of the Bases. The parenthetical phrase '(I.e.. residual heat removal (RHR) shutdown cooling suction Isolation valves) a) a) was added since these are the two valves that we know of that will 0 be required by ITS 3.3.6.1. When drafting of ITS 3.3.6.1 Is complete, ensure that these are the correct valves, and if not, Identify the correct valves. This also affects ITS 3.6.1.3 Condition 0 F. That Is, if these are the only two valves required by ITS 3.3.6.1, then ITS 3.6.1.3 Condition F should only reference CD Conditions A and B. If they are not the only two valves, then evaluate which Conditions you could be in with the new valves and add any new ones to Condition F. CD 0 0 CD -h ZA) 0
-9 46 Ca)
Mdy Setme 20, 2004 _f Page 3 3
.111onday,September 20, 2004 Page 3 of3 , Volume 11, Rev. 0, Page 114 of 431 ATTACHMENT 4 ITS 3.6.1.4, Drywell Air Temperature Attachment.1, Volume 11, Rev. 0, Page 114 of 431
Attachment 1, Volume 11, Rev. 0, Page 115 of 431 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 11, Rev. 0, Page 115 of 431
, Volume 11, Rev. 0, Page 116 of 431 ITS 3.6.1.4 4-f~po~TS3..1.4~
Page 1 of 1 , Volume 11, Rev. 0, Page 116 of 431
Attachment 1, Volume 11, Rev. 0, Page 117 of 431 DISCUSSION OF CHANGES ITS 3.6.1.4, DRYWELL AIR TEMPERATURE ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES M.1 The CTS does not have any requirements for Drywell Air Temperature. ITS LCO 3.6.1.4 requires drywell average air temperature to be < 135 0F. Appropriate ACTIONS and a Surveillance Requirement are also provided. This changes the CTS by incorporating the requirements of ITS 3.6.1.4. In the event of a design basis accident, with an initial drywell average air temperature less than or equal to the LCO temperature limit, the resultant accident temperature profile assures that the drywell structural temperature is maintained below its design temperature and that required safety related equipment will continue to perform its function. Drywell air temperature satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). This change is acceptable because drywell air temperature is an initial condition of a design basis accident. This change is designated as more restrictive because it adds new requirements to the CTS. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 117 of 431
Attachment 1, Volume 11, Rev. 0, Page 118 of 431 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 118 of 431
Attachment 1, Volume 11, Rev. 0, Page 119 of 431 Drywell Air Temperature 3.6.1 ( 3.6 CONTAINMENT SYSTEMS 3.6.1 9j, Drywell Air Temperature 0D LCO 3.6.1 i Drywell average air temperature shall be *W13,TF. D00 APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell average air A.1 Restore drywell average air 8 hours temperature not within temperature to within limit. limit. B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met. AND B.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1 Jj.1 Verify drywell average air temperature is within limit. 24 hours 0D BWR/4 STS 3.6.1.5-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 119 of 431
Attachment 1, Volume 11, Rev. 0, Page 120 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.1.4, DRYWELL AIR TEMPERATURE
- 1. ISTS 3.6.1.5 is renumbered as ITS 3.6.1.4 since ISTS 3.6.1.4, "Drywell Pressure," is not included in the Monticello ITS.
- 2. The brackets have been removed and the proper plant specific information/value has been provided.
Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 120 of 431
Attachment 1, Volume 11, Rev. 0, Page 121 of 431 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 121 of 431
Attachment 1, Volume 11, Rev. 0, Page 122 of 431 Drywell Air Temperature B 3.6.1 4J 0 B 3.6 CONTAINMENT SYSTEMS B 3.6.1 J Drywell Air Temperature BASES BACKGROUND The drywell contains the reactor vessel and piping, which add heat to the airspace. Drywell coolers remove heat and maintain a suitable environment. The average airspace temperature affects the calculated response to postulated Design Basis Accidents (DBAs). The limitation on the drywell average air temperature was developed as reasonable, based on operating experience. The limitation on drywell air temperature is F used in[fteReferenc *1,safetya 0 APPLICABLE Primary containment performance is evaluated for a spect rea SAFETY . sizes for postulated loss of coolant accidents (LOCAs) (Ref. U. Among 0 ANALYSES the inputs to the design basis analysis is the initial drywell average air C temperature (Rel"j). Analyses assume an initial average drywell air temperature of;135JF. This limitation ensures that the safety analysis 0 remains valid by maintaining the expected initial conditions and ensures that the peak LOCA drywell temperature does not exceed the maximumS gPallowable temperature ofj°I9F (Ref.W. Exceeding this design temperature may result in the degradation of the primary containment structure under accident loads. Equipment inside primary containment required to mitigate the effects of a DBA is designed to operate and be capable of operating under environmental conditions expected for the accident. Drywell air temperature satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO In the event of a DBA, with an initial drywell average air temperature less than or equal to the LCO temperature limit, the resultant accident temperature profile assures that the drywell structural temperature is maintained below its design temperature and that required safety related equipment will continue to perform its function. APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, maintaining drywell' average air temperature within the limit is not required in MODE 4 or 5. BWR/4 STS B 3.6.1.5-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 122 of 431
Attachment 1, Volume 11, Rev. 0, Page 123 of 431 Drywell Air Temperature B 3.6.1 BASES ACTIONS A.1 With drywell average air temperature not within the limit of the LCO, drywell average air temperature must be restored within 8 hours. The Required Action is necessary to return operation to within the bounds of the primary containment analysis. The 8 hour Completion Time is acceptable, considering the sensitivity of the analysis to variations in this parameter, and provides sufficient time to correct minor problems. B.1 and B.2 e 0 If the drywell average air temperature cannot be restored to withirzlimit within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.6.1 0( REQUIREMENTS Verifying that the drywell average air temperature is within the LCO limit ensures that operation remains within the limits assumed for the primary containment analyses. Drywell air temperature is monitored in all quadrants and at various elevations (referenced to mean sea level). Due to the shape of the drywell, a volumetric average is used to determine an accurate representation of the actual average temperature. The 24 hour Frequency of the SR was developed based on operating experience related to drywell average air temperature variations and temperature instrument drift during the applicable MODES and the low probability of a DBA occurring between surveillances. Furthermore, the 24 hour Frequency is considered adequate in view of other indications available in the control room, including alarms, to alert the operator to an abnormal drywell air temperature condition. REFERENCES Er7'SAR, Section F[ 0 0(i)
, Sections[6. .4.1i)0 0 FSAR ion[6.2.1.4.51
- 2. USAR. Section 5.2.3.9.
- 3. USAR. Section 5.2.3.2.
- 4. USAR, Table 5.2-7.
BWRI4 STS B 3.6.1.5-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 123 of 431
Attachment 1, Volume 11, Rev. 0, Page 124 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.1.4 BASES, DRYWELL AIR TEMPERATURE
- 1. Changes have been made to reflect those changes made to the Specification.
- 2. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the plant specific nomenclature, number, reference, system description, analysis description, or licensing basis description.
- 3. The brackets have been removed and the proper plant specific information/value has been provided.
- 4. Typographical/grammatical error corrected.
Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 124 of 431
Attachment 1, Volume 11, Rev. 0, Page 125 of 431 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 11, Rev. 0, Page 125 of 431
Attachment 1, Volume 11, Rev. 0, Page 126 of 431 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.6.1.4, DRYWELL AIR TEMPERATURE There are no specific NSHC discussions for this Specification. Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 126 of 431
, Volume 11, Rev. 0, Page 127 of 431 ATTACHMENT 5 ITS 3.6.1.5, Low-Low Set (LLS) Valves , Volume 11, Rev. 0, Page 127 of 431
Attachment 1, Volume 11, Rev. 0, Page 128 of 431 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 11, Rev. 0, Page 128 of 431
C C C ITS 3.6.1.5 ITS 1* 3.0 UMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS i1 F. Recirculation Pump Trip and Alternate Rod Injection CD Initiation See ITS 3.3.4.1 } n) Whenever the reactor IsIn the RUN mode, the Umiting 0 Conditions for Operation for the Instrumentation listed In CD Table 3.2.5 shal be met. 0 G. Safeguards Bus Voltage Protection I 0
- Whenever the safeguards auxildary electrical power system Is required to be operable by Specification 3.9, a See ITS 3.3.8.1 }
-
CD the Umiting Conditions for Operation for the ._ Instrumentation listed In Table 3.2.6 shall be met. 3.6.1.5 H. jnrumen o to Safety/Relief Valve Low-Low Set - 0 0 03 Applicability hnr the safetyfrelief valves are required to be F
~oprabe bySpeifictio 3.6E the Limiting Uondnions- !-{ See ITS 3.3.6.3 }
0 M.1_ lfor Ope~raionfrteIstrumentation fisted in Table 3.2.7 _ CD
- A (0
;i CD 1. Instrumentation for Control Room HabitabiiUty Protection DX Co
- I 1. Whenever the emergency - _ iltran sem Is _
Se ee iTS 3.3.7.1 N 0ID required to De operable oy Speancaton a 7.1. trne CD to Umltlng Conditions for Operation for the radiation -0 Instrumentation listed InTable 3.2.9 shall be met. (1) to A to 3.214.2 48 8/25/94 Amendment No.45, 30, 65, 89 Page 1 of 4
(7 C IC ITS 3.6.1.5 0 ITS TablSeeTS3.3.6.3} C, Min. No. of Min. No. of Operable or Operable or Total No. of Instru- Operating Instrument Operating ment Channels Per Channels Per Trip Required 0 Function Trip Setting Trip Systems Trip System System Conditions 0 Reactor Scram 2(2) 2A or B or C Detection El ACTIONS A Reactor Coolant 1072 k3/992k :3 psig 2(2) 2 2 AorBo and B System Pressure 1062+/-3/982+/-3 psig for Opening/ 1052 +/-3/972+/-3 psig CD Closing (1) 5_ Discharge Pipe . 30 +/- 1 psid (3) 2(2) 2 2 A or B or C 0 Pressure Inhibit 0 and Position ID Indication (0 CD Inhlbit Timers 10:k1 sec: 2(2) 2 2 A or Bo CD
- A)
Z-9 0 MA CD) 3.214.2 60b 3/31/89 Amendment No. 30,43,62 Page 2 of 4
C c 0 ITS 3.6.1.5 ITS CD Table 3.2.7 (continued) 0 Instrumentation folSafety/Reuef Valve Low-Low SetiLOgici CD Notes. LCO 3.6.1.5 (1) Low-Low set and Inhibit logic Isprovded for three non-Automatic Pressure Relief System Vaves. [Tle three valves have staggered See ITS 3.3.6.3 } 0 set po nts as Inculcat edt. 0 (2) Each valve Is provided with two trip, or actuation, systems. 3 (3) Differential pressure with respect to drywall atmosphere. CD
- Required conditions when minimum conditions for operation are not satisfied.
0 CD A) One trip system may be Inoperable for testing or maintenance for up to 72 hours. If two trip systems cannot be made operable at the end of the 72 hour period. within 24 hours reduce reactor pressure to less than 110 psig and reactor water temperature to -A less than 3451F. CD CD B) With two trip systems inoperable. within 24 hours reduce reactor pressure to leis than 110 psia and reactorwater temperature to than MO F.I 0 ACTION ACTION A ACTIONSB A -less C) One'low-low set vL at the end of them l aybe Inoperable for testing or maintenance for up to day period, "thl hours r oe days.fiilvalve cannot be made operable p a e Xrw_ _ I i 0) -to M. -r. l CD) in 12hour adl 4 Add proposed ACTION B for two or more M. inoperable LLS valves 3.214.2 60c 11/16184 Amendment No. 30 Page 3 of 4
( C C 0 ITS 3.6.1.5 ITS 3.0 LIMING CONDmONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS
- 2) Otherwise, be In Hot Shutdown within the next 12 hours and In Cold Shutdown within 0) 0- the following 24 hours. 0
- c. Any time Irradiated fuel Is In the reactor vessel and reactor water temperature Is above 2120 F See ITS 3.4.5 } CD 0
at least one channel of the required leakage 0 detection Instrumentation shall be operable. If all channels of both systems (Drywell Floor Drain Sump Monitoring System and drywell particulate radioactivity monitoring system) are inoperable, restore at least one channel of the required leakage detection Instrumentation to CD operable status within 1 hour, or be in Hot 0 -A Shutdown within the next 12 hours and In Cold
- 0) Shutdown within the following 24 hours. -A Co
[MDS¶.2 and E. Safety/Relief Valves CD E. Safety/Relief Valves 0
- 1. a. Safety/relief valves shall be tested or replaced
-9' Applicability .ratmg reacor 11 and t pow cordons ant pressure a7greater rature Is areaer than 345 aR tha n v 1psi] he safety each refueling outage In accordance with the Inservice Testing Program.
- l See ITS 3.4.3}
-9' co valve function (SOIT acualon) of sovosafety/reliel b. At least two of the safety/rellef valves shall be valves shall be operable (note: Low-Low Set and CD disassembled and Inspected each refueling ADS requirements are located In Specication outage.
/ 3..H. and 3.5.A. respedivoly)4 Valves shall be set as follows: c The integrity of the safety/reliet valve bellows shall be continuously monitored.
8 valves at s 1120 psig
- d. The operability of the bellows monitoring system W
{See ITS 3.4.3 2. If SpecifIcation 3.6.E.1 Is not met, Initiate an orderly shall be demonstrated each operating cycle. shutdown and have reactor coolant pressure and temperature reduced to 110 psig or less and 345' F 2. Low-Low Set Logic surveillance shall be performed or less within 24 hours. In accordance with Table 4.2.1. 3.6/4.6 127 08/21/03 AmendmentNo.30 62,76,02,0, 11. 122,128. 137 Page 4 of 4
Attachment 1, Volume 11, Rev. 0, Page 133 of 431 DISCUSSION OF CHANGES ITS 3.6.1.5, LOW-LOW SET (LLS) VALVES ADMINISTRATIVE CHANGES A.1 Inthe conversion of the Monticello Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1433, Rev. 3, "Standard Technical Specifications General Electric Plants, BWR/4" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. MORE RESTRICTIVE CHANGES M.1 CTS 3.2.H requires the Low-Low Set (LLS) valves to be OPERABLE when CTS 3.6.E requires the S/RVs to be OPERABLE. CTS 3.6.E.1 requires the S/RVs to be OPERABLE "during power operating conditions and whenever reactor coolant pressure is greater than 110 psig and temperature is greater than 345 0F." CTS Table 3.2-7 Required Condition C requires reducing reactor coolant pressure to less than 110 psig and temperature to less than 3450F within 24 hours if an inoperable LLS valve cannot be restored to OPERABLE status. ITS LCO 3.6.1.5 is applicable in MODES 1, 2, and 3 and ITS 3.6.1.5 ACTION B requires the unit to be in MODE 4. This changes the CTS by requiring the LLS valves to be OPERABLE in MODE 2 < 1%RATED THERMAL POWER (RTP) and in MODE 3 when reactor coolant pressure is less than 110 psig and temperature is less than 345 0F, and requires the unit to exit these new MODES of Applicability when a shutdown is required. The purpose of CTS 3.2.H and CTS 3.6.E is to ensure the appropriate number of LLS valves is OPERABLE whenever there is a potential for pressurization of the reactor and subsequent opening of S/RVs and the purpose of CTS Table 3.2-7 is to provide an action to exit the Applicability of the LCO. This change expands the Applicability to require the LLS valves to be OPERABLE at all times when in MODE 2 instead of when > 1% RTP (the CTS 1.0.0 definition states Power Operation is when reactor power is > 1% RTP), and in MODE 3 when reactor. coolant system temperature is > 2120F instead of when reactor coolant pressure is greater than 110 psig and temperature is greater than 3450F. The LLS valves must be OPERABLE at all times in MODES 2 and 3 because considerable energy may be in the reactor core and the limiting design basis transients are assumed to occur in these MODES. The LLS valves may be required to provide pressure relief to discharge energy from the core until such time that the
- Residual Heat Removal (RHR) System is capable of dissipating the core heat.
Consistent with the Applicability change, the action to shutdown the unit now requires exiting the new Applicability, that is,to go to MODE 4. This change is designated as more restrictive because the LCO will be applicable under more reactor conditions. M.2 ITS SR 3.6.1.5.1 requires verification that each required LLS valve opens when manually actuated every 24 months on a STAGGERED TEST BASIS for each valve solenoid. A Note is included that allows this test to not be performed until Monticello Page 1 of 4 Attachment 1, Volume 11, Rev. 0, Page 133 of 431
Attachment 1, Volume 11, Rev. 0, Page 134 of 431 DISCUSSION OF CHANGES ITS 3.6.1.5, LOW-LOW SET (LLS) VALVES 12 hours after reactor steam dome flow is adequate to perform the test. This Surveillance Requirement is not included in the CTS. This changes the CTS by adding a Surveillance Requirement to verify the LLS valves can be manually actuated every 24 months. This change is acceptable because it helps to ensure each required LLS valve is functioning properly and that no blockage exists in the valve discharge line. This change is designated as more restrictive because it adds a Surveillance Requirement that does not appear in the CTS. M.3 ITS SR 3.6.1.5.2 requires verification that the LLS System actuates on an actual or simulated automatic initiation signal every 24 months. A Note is included that valve actuation may be excluded. This Surveillance Requirement is not included in the CTS. This changes the CTS by adding a Surveillance Requirement to verify the LLS System actuates automatically by an actual or simulated automatic initiation signal every 24 months. This change is acceptable because it helps to ensure the mechanical portions (i.e., solenoids) of the LLS function operate as designed when initiated either by an actual or simulated automatic initiation signal. This change is designated as more restrictive because it adds a Surveillance Requirement that does not appear in the CTS. M.4 CTS Table 3.2-7 Note 1 requires three LLS valves to be OPERABLE, however, CTS Table 3.2-7 Condition C only provides actions for when one LLS valve is inoperable. There are no specified actions to take when two or more LLS valves are inoperable. Therefore, 10 CFR 50.36(c)(2) requires the unit to be shut down until the LCO is met. In addition, no time limit in which to complete the unit shutdown is specified in 10 CFR 50.36(c)(2). ITS LCO 3.6.1.5 ACTION B covers the condition when two or more LLS valves are inoperable (second part of Condition B), and it requires the unit to be in MODE 3 in 12 hours and MODE 4 in 36 hours. This changes the CTS by adding finite times to shut down the unit when it is operating. The purpose of ITS 3.6.1.5 ACTION B is to provide specific compensatory actions for when two or more LLS valves are inoperable. This change is acceptable because it provides the necessary and specific actions to take when the requirements are not met and provides appropriate times to complete the actions. This change is designated as more restrictive because it adds specific times to shut down the unit when two or more LLS valves are inoperable. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None Monticello Page 2 of 4 Attachment 1, Volume 11, Rev. 0, Page 134 of 431
Attachment 1, Volume 11, Rev. 0, Page 135 of 431 DISCUSSION OF CHANGES ITS 3.6.1.5, LOW-LOW SET (LLS) VALVES LESS RESTRICTIVE CHANGES L.1 (Category 3 - Relaxation of Completion Time) CTS Table 3.2-7 Required Condition C allows 7 days to restore one inoperable Low-Low Set valve. ITS 3.6.1.5 ACTION A allows 14 days to restore the inoperable LLS valve to OPERABLE status. This changes the CTS by extending the time to restore an inoperable LLS valve from 7 days to 14 days. The purpose of CTS Table 3.2-7 Required Condition C is to allow time to restore one inoperable LLS valve before requiring a reactor shutdown. The Monticello safety analysis only takes credit for two OPERABLE LLS valves, and three LLS Valves are required OPERABLE by LCO 3.6.1.5. Thus, when one LLS valve is inoperable, a sufficient number of LLS valves remain OPERABLE to meet the analysis assumptions. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing time to repair inoperable features. The Required Actions are consistent with safe operation under the specified Condition, considering the OPERABLE status of the remaining LLS valves. This includes the capacity and capability of remaining LLS valves, a reasonable time for repairs or replacement, and the low probability of a design basis accident occurring during the repair period. This change is designated as less restrictive because additional time is allowed to restore an inoperable LLS valve to OPERABLE status than was allowed in the CTS. L.2 (Category 3 - Relaxation of Completion time) CTS Table 3.2-7 Required Condition C states, in part, that if the LLS valve cannot be made OPERABLE within 7 days then, reactor coolant pressure and temperature must be reduced to less than 110 psig and less than 3450 F in 24 hours. ITS 3.6.1.5 ACTION B requires the unit to be in MODE 3 in 12 hours and MODE 4 in 36 hours. This changes the CTS by requiring the unit to be in MODE 3 in 12 hours and extends the time to be outside of the Applicability of the Specification from 24 hours to 36 hours. The change to be in MODE 4 in lieu of the current requirement to reduce reactor coolant pressure to less than 110 psig and reactor coolant temperature to less than 3450F is discussed in DOC M.1. The purpose of CTS Table 3.2-7 Required Condition C is to place the unit outside of the Applicability of the Specification within a reasonable amount of time. This change is acceptable because the Completion Time is consistent with safe operation under the specified Condition, considering the OPERABLE status of the redundant systems or features. This includes the capacity and capability of remaining systems or features, a reasonable time for repairs or replacement, and the low probability of a transient occurring during the allowed Completion Time. The allowed Completion Times are reasonable, based on operating experience, to reach required unit conditions from full power conditions in an orderly manner and without challenging unit systems. This change is acceptable because it requires the unit to be in an intermediate condition (MODE 3) sooner than is currently required (12 hours versus 24 hours). This portion of the change reduces the time the unit would be allowed to continue to operate in MODES I and 2 once the condition is identified. The consequences of a pressurization event are significantly reduced when the reactor is shutdown and a controlled Monticello Page 3 of 4 Attachment 1, Volume 11, Rev. 0, Page 135 of 431
Attachment 1, Volume 11, Rev. 0, Page 136 of 431 DISCUSSION OF CHANGES ITS 3.6.1.5, LOW-LOW SET (LLS) VALVES. cooldown is already in progress. This change is designated as less restrictive because additional time is allowed to place the unit outside the LCO Applicability than is allowed in the CTS. Monticello Page 4 of 4 Attachment 1, Volume 11, Rev. 0, Page 136 of 431
Attachment 1, Volume 11, Rev. 0, Page 137 of 431 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 137 of 431
Attachment 1, Volume 11, Rev. 0, Page 138 of 431 LLS Valves 3.6. 1A CTS (0 3.6 CONTAINMENT SYSTEMS 3.2.H 3.6.1 Low-Low Set (LLS) Valves 0 Table 3.2-7 LCO 3.6.1 The LLS function of [Eirl safety/relief valves shall be OPERABLE. Note I 0 3.2.H, APPLICABILITY: MODES 1, 2, and 3. Table 3.2-7 Condition C, 3.6.E.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Table 3.2-7 A. One LLS valve A.1 Restore LLS valve to 14 days Condition C inoperable. OPERABLE status. Table 3.2-7 B. Required Action and B.1 Be in MODE 3. 12 hours Condition C, DOC M.4 associated Completion Time of Condition A not AND met. B.2 Be in MODE 4. 36 hours OR Two or more LLS valves inoperable. BWR/4 STS 3.6.1.6-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 138 of 431
Attachment 1, Volume 11, Rev. 0, Page 139 of 431 LLS Valves 1 3.6.5 03 CTS SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DOC M.2 SR 3.6.1 .1 ------NOTE ---- ar-- Not required to be performed until 12 hours after 0 reactor steam press e anq flow i adequate to perform the test. 0 Verify each LLS valve opens when manually actuated. Smoths~n a STAGGERED 0D TEST BASIS for each valve solenoicd DOC M.3 SR 3.6.1 t2D -- a------- -NOTEPA---- 0D Valve actuation may be excluded. Verify the LLS System actuates on an actual or simulated automatic initiation signal. dmonths 0 BWR/4 STS 3.6.1.6-2 Rev. 3.0, 03/31104 Attachment 1, Volume 11, Rev. 0, Page 139 of 431
Attachment 1, Volume 11, Rev. 0, Page 140 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.1.5, LOW-LOW SET (LLS) VALVES
- 1. ISTS 3.6.1.6 is renumbered as ITS 3.6.1.5 since ISTS 3.6.1.4, "Drywell Pressure," is not included in the Monticello ITS.
- 2. The brackets have been removed and the proper plant specific information/value has been provided.
- 3. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the plant specific nomenclature, number, reference, system description, analysis description, or licensing basis description.
- 4. A minimum reactor steam pressure is not necessary to prevent damage to the S/RVs. Only an adequate reactor steam flow is necessary to properly test the S/RVs.
(i.e., as long as the bypass valves are controlling reactor steam pressure, the S/RVs can be tested safely). Monticello Page 1 of I Attachment 1, Volume 11, Rev. 0, Page 140 of 431
Attachment 1, Volume 11, Rev. 0, Page 141 of 431 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 141 of 431
Attachment 1, Volume 11, Rev. 0, Page 142 of 431 LLS Valves
- ~~B 3.6. 1 0
B 3.6 CONTAINMENT SYSTEMS B Low-Low Set (LLS) Valves 03 BASES BACKGROUND The safety/relief valves (S/RVs) can actuate in either the safety mode, the Automatic Depressurization System mode, or the LLS mode. In the LLS mode (or power actuated mode of operation), a pneumatic diaphragm s second stge 1_ and stem assembly overcomes the spring orce and opens th pilot valve. As in the safety mode, opening th_% pilot valve allows a differential 0D pressure to develop across the main valve piston and opens the main valve. The main valve can stay open with valve inlet steam pressure as low as 5Qj psig. Below this pressure, steam pressure may not be 0 inI m-a sufficient to hold the main valve open against the spring force of the.l valves. The pneumatic operator is arranged so that its malfunction will mai not prevent the valve PRfrom lifting if steam inlet pressure exceeds the safety mode pressure setpoints. E ~ of the S/RVs are equipped to provide the LLS function. The LLS logic causes the LLS valves to be opened at a lower pressure than the 0 relief or safety mode pressure setpoints and stay open longer, so that [I
- isrioreAhth 60 ,,is prevented on subsequent actuations.
Therefore, the LLS function prevents excessive short duration S/RV 0 cycles with valve actuation at the relief setpoint. Each S/RV discharges steam through a discharge line and quencher to a location near the bottom of the suppression pool, which causes a load on the suppression pool wall. Actuation at lower reactor pressure results in a lower load. APPLICABLE The LLS relief mode functions to ensure that the containment design SAFETY basis o one operatin qsise ent actuations is me. In other ANALYSES words, multiple simultaneous openings of 4S/RVs (following the initial InoLS no subsequent opening), and the corresponding higher load ed. The safety SoVsend a analysis demonstrates that the LLS functions to avoid the induced thrust subsequent loads on the S/RV discharge line resulting from "subsequent actuations"
>5.75 seconds for of the S/RV during Design Basis Accidents (DBAs). Furthermore, the the LLSvaves ,LLS function justifies the primary containment analysis assumption that simultaneous S/RV openings occur only on the initial actuation for DBAs.
Eventhoug h LLS S/RVs are specified, r1LLS S/RVs (0D0 operate in any DBA analysis. 7 I
-nly w l are requiredto]
LLS valves satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). BWR/4 STS B 3.6.1.6-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 142 of 431
Attachment 1, Volume 11, Rev. 0, Page 143 of 431 LLS Valves B 3.6.1. 0D BASES LCO E LLS valves are required to be OPERABLE to satisfy the 0 assumptions of the safety analyses (Ref. 1). The requirements of this LCO are applicable to the mechanical and electrical/pneumatic capability of the LLS valves to function for controlling the opening and closing of the S/RVs. APPLICABILITY In MODES 1, 2, and 3, an event could cause pressurization of the reactor and opening of S/RVs. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the LLS valves OPERABLE is not required in MODE 4 or 5. ACTIONS A.1 With one LLS valve inoperable, the remaining OPERABLE LLS valves are adequate to perform the designed function. However, the overall reliability is reduced. The 14 day Completion Time takes into account the redundant capability afforded by the remaining LLS valves and the low probability of an event in which the remaining LLS valve capability would be inadequate. B.1 and B.2 If two or more LLS valves are inoperable or if the inoperable LLS valve cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE REQUIREMENTS 0D A manual actuation of each LLS valve is performed to verify that the valve and solenoids are functioning properly and no blockage exists in the valve E ~discharge line. This can be demonstrated by the response of the turbine con obypass valva by a change in the measured steam flow, or by any other method that is suitable to verify steam flow. steam dome pressure wstbe available to perform t IAde e reactor est to avoid 0D damaging the Av .Adequate pressure at whi s test is to be performes [920] psig (the pressure remended by the valve / BWR/4 STS B 3.6.1.6-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 143 of 431
Attachment 1, Volume 11, Rev. 0, Page 144 of 431 LLS Valves B 3.6.1 i BASES SURVEILLANCE REQUIREMENTS (continued) [Sufficient time is therefore allowed after the required flow Is [achieved to perform this test. ( manuta r). Also] $dequate steam flow must be passing through the Imain urmine od turbine bypass valves to continue to control reacto 6one pressure when the LLS valves divert steam flow upon opening. Adequate l i steam flow is represented p ast turbine bypass valvFeu ernL tte erfm T eor the 10F wb/hs conluddmonth Frequency was based on ta me ys V test i cs s req u ired t o the Sueillan ce wever, Nandtstehmay NotbeaaillFrequency o months ona STAGGERED required toube TEST BASIS ensures that each solenoid for each m s alternatessure T hoursater reactor th tested. Operating experience has shown theat these components usual steam flow Is adequte pass the Surveillance when performed at thflM month Frequencyi t Therfore, he Frequency was concluded to be acceptable from a reliability standpoint. thSstrinsvoesueroecionaecerfidbyReerncn2pro allformed fumn al c tSince o s ignaressure d,1 is ousi al to dtlr required crmr Surveillancally the frad1ro2h uo et
.y not be available d g a unit outage, t receipt pecfoed durini the sigals following a unit onalteterforme tis allowedtpriotho performing the test because valve of t f i the setpoints fodsgerdture protection aren verified by Reference pror fhe auto valve installation.sger adequate L ctor steam dome pressur TId ctua.owSntetLLs required flow is m F is preparef fand or perfogr 0hetes reached Is sufficient to achieve stable _L 3.6.1aata econditionstforaSR testing and provides a reasonable time to comvpeliete thm The LLSo designated S/RVs are required to actuate automatically upon receipt of specific initiation signals. A system functional test is performed to verify that the mechanical portions (i.e., solenoids) of the LLS function operate as designed when initiated either by an actual or simulated rCO 3.3.6.3. utomatic initiation signal. The LOGIC SYSTEM FUNCTIONAL TEST in Instrumentation(,' S overlaps this SR to provide complete testing of the safety )
function.
, hnmonth Frequency is based on the need to perform this (i0 8 Surveillance under the conditions that apply during a plant outage and the
\ potential for an unplanned transient if the Surveillance were performed
\ with the reactor at power. Operating experience has shown these
\ components usually pass the Surveillance when performed at the 3gmonth Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
0 This SR is modified by a Note that excludes valve actuation. This prevents a reactor pressure vessel pressure blowdown. BWR/4 STS B 3.6.1.6-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 144 of 431
Attachment 1, Volume 11, Rev. 0, Page 145 of 431 LLS Valves B 3.6.1. 3 BASES 0 REFERENCES IEYJ1. I*SAR, Section[[ 0
- 2. ASM . oier and re 9 2Operaton and Maintenance (OM) Codel e o e, Section l n)
BWR/4 STS B 3.6.1.6-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 145 of 431
Attachment 1, Volume 11, Rev. 0, Page 146 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.1.5 BASES, LOW-LOW SET (LLS) VALVES
- 1. Changes have been made to reflect changes made to the Specification.
- 2. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the plant specific nomenclature, number, reference, system description, analysis description, or licensing basis description.
- 3. The brackets have been removed and the proper plant specific information/value has been provided.
- 4. Editorial change made for enhanced clarity or to be consistent with similar statements in other places in the Bases.
- 5. Typographical error corrected.
- 6. Changes made to be consistent with the Bases for similar Surveillances in ITS 3.4.3 and ITS 3.5.1.
Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 146 of 431
Attachment 1, Volume 11, Rev. 0, Page 147 of 431 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 11, Rev. 0, Page 147 of 431
Attachment 1, Volume 11, Rev. 0, Page 148 of 431 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.6.1.5, LOW-LOW SET (LLS) VALVES There are no specific NSHC discussions for this Specification. Monticello Page 1 of I Attachment 1, Volume 11, Rev. 0, Page 148 of 431
Attachment 1, Volume 11, Rev. 0, Page 149 of 431 ATTACHMENT 6 - ITS 3.6.1.6, Reactor Building-To-Suppression Chamber Vacuum Breakers Attachment 1, Volume 11, Rev. 0, Page 149 of 431
Attachment 1, Volume 11, Rev. 0, Page 150 of 431 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 11, Rev. 0, Page 150 of 431
( C C ITS 3.6.1.6 ITS 0 ITS 3.0 WAMING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REOUIREMENTS i 3.6.1.6 3. Pressure Suppression Chamber- Reactor Bulding 3. Pressure Suppression Chamber . Reactor Building ID LCO 3.6.1.6 Vacuum Breakers
. Except as specified In 3.7.A.3.b below SR SR 3.6.1.6.2
)
Vacuum Breakers a. la. e rrt"Ol4_X^ lAdd propose-dSR 3.161 IIin piresureauppFRUsiun cnariner-re10anor pressure suppression chamber-reactor buildit SR 3.6.1.6.33i building vacuum breakers and associated vacuum breakers shall be operable at afl times Instrumentation Including set point shall be 0 Applicability when the rima containment checked for proper operation every three 0 M.1 rur Teet point of the differential months. [pressure istnumentatlon which actuates the CD SR36.1..3 ----- Jpressure suppression charnber-reactor building SR 3.6.1.6.3 vacuum breakers shall be s0.5 psL CD
- b. From and after the date that ot the M pressure suppression ohambr-reactr builin ACTIONS A vacuum breakers is made or found to be 0 and C Inoperable for any reason, reaeor operation Is 0 Ei
-o i -- M.3 CD made operable, provided that the repair CD procedure does not violate primay containment Integrity.. Add proposed ACTION A for an open vacuum breaker in both lines
-9' .4 -:21
- c. If requirements of 3.7.A3 cannot be met the -Add proposed ACTIONS B and D r ACTION E reactor shal be laced In a Cold Shutdown condition wihin hours.
IDE 3 in 12hor n 3.7/4.7 163 01/28105 Amendment No. O, 43 o, 141 Page 1 of I
Attachment 1, Volume 11, Rev. 0, Page 152 of 431 DISCUSSION OF CHANGES ITS 3.6.1.6, REACTOR BUILDING-TO-SUPPRESSION CHAMBER VACUUM BREAKERS ADMINISTRATIVE CHANGES A.1 In the conversion of the Monticello Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1433, Rev. 3, "Standard Technical Specifications General Electric Plants, BWR/4" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 CTS 3.7.A.3.a requires 'two" reactor building-to-suppression chamber vacuum breakers to be OPERABLE. ITS LCO 3.6.1.6 requires "each" reactor building-to-suppression chamber vacuum breakers to be OPERABLE. This changes the CTS by using the term "each" instead of the actual number of vacuum breakers. The purpose of CTS 3.7.A.3.a is to ensure the appropriate number of reactor building-to-suppression chamber vacuum breakers are OPERABLE to prevent possible damage to the integrity of the primary containment that could result from a negative pressure in the suppression chamber or drywell in excess of the primary containment design limit. CTS 3.7.A.3.a requires "two" reactor building-to-suppression chamber vacuum breakers to be OPERABLE, however the CTS Bases define a vacuum breaker as two in-series valves. ITS 3.6.1.6 defines a vacuum breaker as an individual valve. Thus, requiring "each" vacuum breaker to be OPERABLE in the ITS requires four total valves to be OPERABLE, consistent with the CTS requirement. This change is designated as administrative since it does not result in any technical change to the CTS. MORE RESTRICTIVE CHANGES M.1 CTS 3.7.A.3.a is applicable when Primary Containment Integrity is required. CTS 3.7.A.2.a.(1), specifies that Primary Containment Integrity is required at all times when the reactor is critical or when reactor water temperature is above 212 0F and fuel is in the reactor vessel. ITS 3.6.1.6 is applicable in MODES 1, 2, and 3. This changes the CTS by requiring the reactor building-to-suppression chamber vacuum breakers to be OPERABLE in MODE 2 when reactor water temperature is less than or equal to 212 0F. The purpose of CTS 3.7.A.3.a is to ensure the reactor building-to-suppression chamber vacuum breakers are OPERABLE to prevent possible damage to the integrity of the primary containment that could result from a negative pressure in the suppression chamber or drywell in excess of the primary containment design limit. The reactor building-to-suppression chamber vacuum breakers are required to be OPERABLE during MODES 1, 2, and 3 when a design basis accident could cause a release of radioactive material to the primary containment. In MODES I and 3, the reactor coolant temperature will always be above 2120F. In MODE 2, the reactor coolant temperature may be less than or equal to 2120 F when the reactor is subcritical but control rods are withdrawn. Monticello Page 1 of 4 Attachment 1, Volume 11, Rev. 0, Page 152 of 431
Attachment 1, Volume 11, Rev. 0, Page 153 of 431 DISCUSSION OF CHANGES ITS 3.6.1.6, REACTOR BUILDING-TO-SUPPRESSION CHAMBER VACUUM BREAKERS Therefore, it is necessary and acceptable to require the reactor building-to-suppression chamber vacuum breakers to be OPERABLE. This change is designated as more restrictive because the LCO will be applicable under more reactor operating conditions than in the CTS. M.2 ITS SR 3.6.1.6.1 requires verification that each vacuum breaker is closed every 14 days. Two Notes are included that specify the Surveillance is not required to be met if 1) the vacuum breakers are open during Surveillances; or 2) if the vacuum breakers are open when performing their intended function. This Surveillance Requirement is not included in the CTS. This changes the CTS by adding a Surveillance Requirement to verify each vacuum breaker is closed every 14 days. This change is acceptable because it helps to ensure each vacuum breaker is in its normally closed position. This helps ensure a leak path does not exist for the primary containment. This change is designated as more restrictive because it adds a Surveillance Requirement that is not required in the CTS. M.3 CTS 3.7.A.3.b allows seven days to restore one reactor building-to-suppression chamber vacuum breaker found inoperable for any reason. Under similar inoperabilities, ITS 3.6.1.6 ACTIONS A and C provide 72 hours to restore the inoperable vacuum breaker(s) (i.e., one of the two vacuum breakers in a line open or one or two vacuum breakers in a line inoperable for opening) to OPERABLE status. This changes the CTS by reducing the Completion Time to restore an inoperable vacuum breaker from,7 days to 72 hours. The purpose of CTS 3.7.A.3.b is to ensure the vacuum breakers support the leak tight requirements for the primary containment boundary, and provide sufficient vacuum relief when the primary containment depressurizes below the reactor building pressure. This change is acceptable because the proposed Completion Times take into account the redundancy capability afforded by the remaining vacuum breakers the fact that the OPERABLE vacuum breaker in the affected line is closed, and the low probability of an event occurring that would require the vacuum breakers to be OPERABLE during this period. This change is designated as more restrictive because the ITS Completion Time is less than the Completion Time that appears in the CTS. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None Monticello Page 2 of 4 Attachment 1, Volume 11, Rev. 0, Page 153 of 431
Attachment 1, Volume 11, Rev. 0, Page 154 of 431 DISCUSSION OF CHANGES ITS 3.6.1.6, REACTOR BUILDING-TO-SUPPRESSION CHAMBER VACUUM BREAKERS LESS RESTRICTIVE CHANGES L.1 (Category 4 - Relaxation of Required Action) CTS 3.7.A.3.b allows an inoperability in only one reactor building-to-suppression chamber vacuum breaker line for any reason. In addition, the vacuum breakers in the line must still be ensuring primary containment integrity is met (i.e., both valves in the line cannot be open). ITS 3.6.1.6 ACTIONS Note and ACTIONS A, B, and D, specify ACTIONS for when both lines have inoperable reactor building-to-suppression chamber vacuum breaker(s) and when one line has both vacuum breakers open. Specifically, ITS 3.6.1.6 ACTION A in combination with the ACTIONS Note will allow each line to have one vacuum breaker open for up to 72 hours, ITS 3.6.1.6 ACTION B will allow one or both lines to have two vacuum breakers open for up to 1 hour, and ITS 3.6.1.6 ACTION D will allow both lines to have one or more vacuum breakers inoperable for opening for up to 1 hour. This changes the CTS by allowing multiple vacuum breaker inoperabilities based on the reason for and proximity of the inoperability, and adding an ACTIONS Note to permit separate condition entry for each relief line. The purpose of CTS 3.7.A.3.b is to ensure each vacuum breaker (in each of two parallel vacuum breaker lines) is closed to maintain the primary containment OPERABLE for a DBA LOCA and is capable of opening to prevent depressurization of the primary containment below the reactor building pressure. ITS 3.6.1.6 ACTIONS Note and ACTIONS A, B, and D,have been added to specify ACTIONS consistent with the type of inoperability and the effect on maintaining at least one reactor building-to-suppression chamber line OPERABLE for opening and both lines closed. ITS 3.6.1.6 ACTION A in combination with the ACTIONS Note ensures that for each line, any one open vacuum breaker is closed within 72 hours. ITS 3.6.1.6 ACTION B ensures primary containment leak tightness is restored within 1 hour consistent with requirements of LCO 3.6.1.1, Primary Containment. ITS 3.6.1.6 ACTION D ensures that for two vacuum breaker lines with one or more vacuum breakers inoperable for opening that one line is restored to OPERABLE status within 1 hour to prevent primary containment depressurization below reactor building pressure. This change is acceptable because the Required Actions are consistent with the analyses requirements to maintain the Primary Containment OPERABLE, and it specifies remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing time to repair inoperable features. The Required Actions are consistent with safe operation under the specified Condition, considering the OPERABLE status of the remaining vacuum breakers. This includes the capacity and capability of the remaining vacuum breakers, a reasonable time for repairs or replacement, and the low probability of a design basis accident occurring during the repair period. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. L.2 (Category 3 - Relaxation of Completion Time) CTS 3.7.A.3.c requires the unit to be placed in the cold shutdown condition within 24 hours if the inoperable reactor building-to-suppression chamber vacuum breaker cannot be restored to OPERABLE within 7 days. ITS 3.6.1.6 ACTION E requires the unit be in Monticello Page 3 of 4 Attachment 1, Volume 11, Rev. 0, Page 154 of 431
Attachment 1, Volume 11, Rev. 0, Page 155 of 431 DISCUSSION OF CHANGES ITS 3.6.1.6, REACTOR BUILDING-TO-SUPPRESSION CHAMBER VACUUM BREAKERS MODE 3 in 12 hours and in MODE 4 in 36 hours. This changes the CTS by requiring the unit to be in MODE 3 in 12 hours and by extending the time to be in cold shutdown (i.e., MODE 4) from 24 hours to 36 hours. The purpose of CTS 3.7.A.3.c is to place the unit outside the Applicability of the Specification within a reasonable amount of time. This change is acceptable because the Completion Time is consistent with the safe operation under the specified Condition, considering the OPERABLE status of the redundant systems or features. This includes the capacity and capability of remaining systems or features, a reasonable time for repairs or replacement, and the low probability of a DBA occurring during the allowed Completion Time. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. This change is also acceptable because it requires the unit to be in an intermediate condition (MODE 3) sooner than is currently required (12 hours versus 24 hours). This portion of the change reduces the amount of time the unit would be allowed to continue to operate in MODES 1 and 2 once the condition is identified. The consequences of a pressurization event are significantly reduced when the reactor is shutdown and a controlled cooldown is already in progress. This change is designated as less restrictive because additional time is allowed to place the unit outside the LCO Applicability than is allowed in the CTS. Monticello Page 4 of 4 Attachment 1, Volume 11, Rev. 0, Page 155 of 431
Attachment 1, Volume 11, Rev. 0, Page 156 of 431 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 156 of 431
Attachment 1, Volume 11, Rev. 0, Page 157 of 431 Reactor Building-to-Suppression Chamber Vacuum Breakers 3.6.11 0D CTS 3.6 CONTAINMENT SYSTEMS 3.7.A3 3.6.1 Reactor Building-to-Suppression Chamber Vacuum Breakers 3.7.A.3.a LCO 3.6.1 - Each reactor building-to-suppression chamber vacuum breaker shall be OPERABLE. 3.7.A.3-a APPLICABILITY: MODES 1, 2, and 3. ACTIONS
-NOTE.
DOC L.1 Separate Condition entry is allowed for each line. CONDITION REQUIRED ACTION COMPLETION TIME 3.7.A.3.b, A. One or more lines with A.1 Close the open vacuum 72 hours one reactor building-to- breaker. suppression chamber vacuum breaker not closed. DOC L.1 B. One or more lines with B.1 Close one open vacuum 1 hour two reactor building-to- breaker. suppression chamber vacuum breakers not closed.
-t 3.7A.3.b C. One line with one or C.1 Restore the vacuum 72 hours .
more reactor building-to- breaker(s) to OPERABLE suppression chamber status. vacuum breakers inoperable for opening. DOC L.1 D. Two [or ore lines with one or more reactor D.1 Restore all vacuum breakers intone line to 1 hour 0 building-to-suppression OPERABLE status. chamber vacuum breakers inoperable for opening. BWR/4 STS 3.6.1.7-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 157 of 431
Attachment 1, Volume 11, Rev. 0, Page 158 of 431 Reactor Building-to-Suppression Chamber Vacuum Breakers 3.6.1.T (0
-T ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME 3.7.A.3.c E. Required Action and E.1 Be in MODE 3. 12 hours Associated Completion Time not met. AND E.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DOC M.2 SR 3.6.1 .1 --- NOTES------- (0
- 1. Not required to be met for vacuum breakers that are open during Surveillances.
- 2. Not required to be met for vacuum breakers open when performing their intended function.
Verify each vacuum breaker is closed. 14 days 4.7.A.3.a SR3.6.1.e21 Perform a functional test of each vacuum breaker. R92]days (00 317A.3.b. 4.7.A3.a SR 3.6.1. .3 Verify the opening setpoint of each vacuum breaker J[1 8] Xbonthsl (DO is S0.5 psid. BWR/4 STS 3.6.1.7-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 158 of 431
Attachment 1, Volume 11, Rev. 0, Page 159 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.1.6, REACTOR BUILDING-TO-SUPPRESSION CHAMBER VACUUM BREAKERS
- 1. ISTS 3.6.1.7 is renumbered as ITS 3.6.1.6 since ISTS 3.6.1.4, "Drywell Pressure," is not included in the Monticello ITS.
- 2. The brackets have been removed and the proper plant specific information/value has been provided.
Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 159 of 431
Attachment 1,Volume 11, Rev. 0, Page 160 of 431 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 160 of 431
Attachment 1, Volume 11, Rev. 0, Page 161 of 431 Reactor Building-to-Suppression Chamber Vacuum Breakers B 3.61 0 B 3.6 CONTAINMENT SYSTEMS 3.6.1, Reactor Building-to-Suppression Chamber Vacuum Breakers 0 BASES BACKGROUND The function of the reactor building-to-suppression chamber vacuum breakers is to relieve vacuum when primary containment depressurizes below reactor building pressure. If the drywell depressurizes below reactor building pressure, the negative differential pressure is mitigated by flow through the reactor building-to-suppression chamber vacuum breakers and through the suppressionochamber-to-drywell vacuum 0 breakers. The design of the external (reactor building-to-suppression chamber) vacuum relief provisions consists of two vacuum breakers (a svacuum breaker and an air operated butterfly valve), located in series in Fepar -" tithe suppression chamber swtch-each of tw lines from the reactor buildingby, facommon 20In line a .l acted differential pressur JThe
, 2 on vacuum breaker is self actuating and can be eRA operated for testing purposes. The two vacuum breakers in series must be closed to m ai snng check a leak tight primary containment boundary. IEe}
A negative differential pressure across the drywell wall is caused byErt A depressurization of the drywell. Events that cause this[i~] depressurization are cooling cycles, inadvertentfprimarv ceniainmen tl(&) spray actuation, and steam condensation in the event of a primary system rupture. Reactor building-to-suppression chamber vacuum breakers prevent an excessive negative differential pressure across the primary containment boundary. Cooling cycles result in minor pressure transients in the drywell, which occur slowly and are normally controlled by he in F ventilation equipment. Inadvertent spray actuation results in a more significant pressure transient and becomes important in sizing the external (reactor building-to-suppression chamber) vacuum breakers. The external vacuum breakers are sized on the basis of the air flow from the secondary containment that is required to mitigate the depressurization transient and limit the maximum negative containment (drywell and suppression chamber) pressure to within design limits. The maximum depressurization rate is a function of the primary containment spray flow rate and temperature and the assumed initial conditions of the primary containment atmosphere. Low spray temperatures and atmospheric conditions that yield the minimum amount of contained noncondensible gases are assumed for conservatism. BWR/4 STS B 3.6.1.7-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 161 of 431
Attachment 1, Volume 11, Rev. 0, Page 162 of 431 Reactor Building-to-Suppression Chamber Vacuum Breakers B 3.6.1 BASES APPLICABLE Analytical methods and assumptions involving the reactor building-to-SAFETY suppression chamber vacuum breakers are presented in Reference 1 as ANALYSES part of the accident response of the containment systems. Internal (suppressiorEchamber-to-drywell) and external (reactor building-to-suppression chamber) vacuum breakers are provided as part of the primary containment to limit the negative differential pressure across the drywell and suppression chamber walls, which form part of the primary containment boundary. The safety analyses assume the external vacuum breakers to be closed initially and to be fully open atJj.5M psid (Ref. 1). Additionally, of the two reactor building-to-suppression chamber vacuum breakerslone is Inaline(i assumed to fail in a closed position to satisfy the single active failure criterion. Design Basis Accident (DBA) analyses require the vacuum breakers to be closed initially and to remain closed and leak tight with positive primary containment pressure. E cases were considered in the safety analyses to determine the adequacy of the external vacuum breakers: [ flla. A sm rea loss of coolant accident followed by actuation of both 3 spray loop )
- b. Inadvertent actuation of one r tainment spray loop during Inormal operation,-
jI. Inadvertent actuation of botN nrinary ainmenl spray loops during normal operatio i -_______
- d. A postulated DBA assuming E ergerfcy Core Cooling Systems (ECCS) runout flow ondensation effectiveness of 50%, and l A postulated DBA assuming ECCS runout flow with a condensation effectiveness of 100%.
The results of these[ cases show that the external vacuum breakers, with an opening setpoint of 10.-5 psid, are capable of maintaining the differential pressure within design limits. The reactor building-to-suppression chamber vacuum breakers satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). BWR/4 STS B 3.6.1.7-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 162 of 431
Attachment 1, Volume 11, Rev. 0, Page 163 of 431 Reactor Building-to-Suppression Chamber Vacuum Breakers B 3.6. BASES LCO All reactor'building-to-suppression chamber vacuum breakers are required to be OPERABLE to satisfy the assumptions used in the safety alyses. The requirement ensures that the two vacuum breakers (i) swing check vacuum breaker and air operated butterfly valve) in each of the two lines from the reactor building to the suppression chamber airspace are closed (except during testing or when performing their intended function). Also, the requirement ensures both vacuum breakers in each line will open to relieve a negative pressure in the suppression chamber. APPLICABILITY In MODES 1, 2, an , BA could cause pressuriz primary containm . ODES 1,2, and 3, the ion Pool Srav Svste l i.flaGied to be OPERABLE to rt eefcso B~ ~cessivIl [negative pressure inside mar containment coul inadvertent intainolh .Therefore, the vacuum brear required to beO LEn MODES 1, 2, and 3, wh~ ppressior Pool Sp em is required to be OPERABL itigate the effects vertent actuation of the SupesolSray SVstem. ER/h MODES 1,2, and 3, a DBA could result in excessive negative C differential pressure across the drywell wall caused by the rapid depressurization of the drywell. The event that results in the limiting rapid
, after the depressurization of the drywell is the primary system rupture, which suppression cshamber-to purges the drywell of air and fills the drywell free airspace with steam.
drywell vacuum breakers Subseauent condensation of the steam would result in depressurization differential pressuie of the drywel. The limiting pressure and temperature of the primary between the suppression system prior to a DBA occur in MODES 1, 2, and 3.* would result in { depressurizationfothO In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining reactor building-to-suppression chamber vacuum breakers OPERABLE is not required in MODE 4 or 5. ACTIONS A Note has been added to provide clarification that, for the purpose of this LCO, separate Condition entry Is allowed for eachl penetrF595F-now pat". reactor building-to-suppression chamber vacuum breaker line A.1 With one or more vacuum breakers not closed, the leak tight primary containment boundary may be threatened. Therefore, the inoperable vacuum breakers must bel restored to OsERA Eatus or the open I Ivacuureaked closed within 72 hours. The 72 hour Completion Time is BWR/4 STS B 3.6.1.7-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 163 of 431
Attachment 1, Volume 11, Rev. 0, Page 164 of 431 Reactor Building-to-Suppression Chamber Vacuum Breakers B 3.6.1 5 BASES ACTIONS (continued) consistent with requirements for inoperable suppressiore~arber-to, ( drywell vacuum breakers in LCO 3.6.1.@7"Suppressio vChamber-to- U( i) Drywell Vacuum Breakers." The 72 hour Completion Time takes into account the redundancy capability afforded by the remaining breakers, the fact that the OPERABLE breaker in each of the lines is closed, and the low probability of an event occurring that would require the vacuum breakers to be OPERABLE during this period. B.1 With one or more lines with two vacuum breakers not closed, primary containment integrity is not maintained. Therefore, one open vacuum breaker must be closed within 1 hour. This Completion Time is consistent with the ACTIONS of LCO 3.6.1.1, "Primary Containment," which requires that primary containment be restored to OPERABLE status within 1 hour. C.1 With one line with one or more vacuum breakers inoperable for opening, the leak tight primary containment boundary is intat Teaility to and the remaining OPERABLE vacuum breakers In the other line mthreateaned Ivpt that causes a containment deprsyzains 0 are capable of providing theteehowever, if both vacuum breakers in alesonvacuum the vacuum relief function. However, overall system breaker peyietration-are not OPER6BLB. Therefore, the inoperable reliablity is reduced because a single failure vacuum breaker must be restored to OPERABLE status within 72 hours. (to open) In one of the This is consistent with the Completion Time for Condition A and the fact vacuum breakers in the other line results In a loss that the leak tight primary containment boundary is being maintained. of the vacuum breaker function D.1 With two or ore lines with one or more vacuum breakers inoperable for opening, the primary containment boundary is intact. However, in the 0 event of a containment depressurization, the function of the vacuum breakers is lost. Therefore, all vacuum breakers ingoneM line must be restored to OPERABLE status within 1 hour. This Completion Time is 0 consistent with the ACTIONS of LCO 3.6.1.1, which requires that primary containment be restored to OPERABLE status within 1 hour. BWR/4 STS B 3.6.1.7-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 164 of 431
Attachment 1, Volume 11, Rev. 0, Page 165 of 431 Reactor Building-to-Suppression Chamber Vacuum Breakers B 3.6-1J BASES ACTIONS (continued) E.1 and E.2 If all the vacuum breakers infondj line cannot be closed or restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.6.1 17.1 REQUIREMENTS ( Each vacuum breaker is verified to be closed to ensure that a potential breach in the primary containment boundary is not present. This Surveillance is performed by observing local or control room indications of vacuum breaker positionlo-rby verifying a diffejseffial pressure of l [0.5] psid is ma~ned between the reactp l6uilding and sutpression I D chambe. The 14 day Frequency is based on engineering judgment, is considered adequate in view of other indications of vacuum breaker status available to operations personnel, and has been shown to be acceptable through operating experience. Two Notes are added to this SR. The first Note allows reactorMEo- Z ) suppression chamber vacuum breakers opened in conjunction with the performance of a Surveillance to not be considered as failing this SR. These periods of opening vacuum breakers are controlled by plant procedures and do not represent inoperable vacuum breakers. The second Note is included to clarify that vacuum breakers open due to an actual differential pressure are not considered as failing this SR. SR 3.6.1 .2L Each vacuum breaker must be cycled to ensure that it opens properly to perform its design function and returns to its fully closed position. This ensures that the safety analysis assumptions are valid. TheQ92 day Frequency of this SR was developed based upon Inservice Testing Program requirements to perform valve testing at least once every T92 days. BWR/4 STS B 3.6.1.7-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 165 of 431
Attachment 1, Volume 11, Rev. 0, Page 166 of 431 Reactor Building-to-Suppression Chamber Vacuum Breakers B 3.6.1 1 0 BASES SURVEILLANCE REQUIREMENTS (continued) addF6- 0 Demonstration of vacuum breaker opening setpoint is necessary to ensure that the safety analysis assumption regarding vacuum breaker full open differential pressure of sIIO.51 psid is valid. lThe ] month Frequency is ased on the need to perform this eillance under the conditions that a ply during a plant outage and e potential for an unplanned tra eset~if thp Surveillancewr rfmd with the reactor at power. For his uniteffontFrequency has been shown to be acce ta ebased on operating experienc an Is u erjustified because of other svreillances performed at shorte requencies that convey the pro functioning status of each va m breake. REFERENCES LUJ .IAScinf r "~ ,00 BWR/4 STS B 3.6.1.7-6 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 166 of 431
Attachment 1, Volume 11, Rev. 0, Page 167 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.1.6 BASES, REACTOR BUILDING-TO-SUPPRESSION CHAMBER VACUUM BREAKERS
- 1. Changes have been made to reflect changes made to the Specifications.
- 2. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the plant specific nomenclature, number, reference, system description, analysis description, or licensing basis description.
- 3. The brackets have been removed and the proper plant specific information/value has been provided.
- 4. Inadvertent actuation of the suppression pool spray system is not the main concern for depressurizing the drywell, a LOCA inside the drywell is the main concern. In addition, actuation of suppression pool spray is not a major concern in determining the adequacy of the vacuum breakers at Monticello, the drywell spray is the more limiting system. Therefore, this section has been revised to place emphasis on the proper events.
.5. These punctuation corrections have been made consistent with the Writers Guide for the Standard Technical Specifications, NEI 01-03, Section 5.1.3.
- 6. Typographical error corrected.
- 7. Change made for clarity. The term "rapid" does not apply to cooling cycles, since the description of cooling cycles states that it occurs slowly.
- 8. The vacuum breakers are accessible during power operations and indicators are provided locally and in the control room. Therefore, Monticello does not believe this alternate method of determining position is necessary.
- 9. Changes have been made to be consistent with the Specification.
Monticello Page 1 of I Attachment 1, Volume 11, Rev. 0, Page 167 of 431
Attachment 1,Volume 11, Rev. 0, Page 168 of 431 Specific No Significant Hazards Considerations (NSHCs) Attachment 1,Volume 11, Rev. 0, Page 168 of 431
Attachment 1, Volume 11, Rev. 0, Page 169 of 431 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.6.1.6, REACTOR BUILDING-TO-SUPPRESSION CHAMBER VACUUM BREAKERS There are no specific NSHC discussions for this Specification. Monticello Page 1 of I Attachment 1, Volume 11, Rev. 0, Page 169 of 431
Attachment 1, Volume 11, Rev. 0, Page 170 of 431 ATTACHMENT 7 ITS 3.6.1.7, Suppression Chamber-To-Drywell Vacuum Breakers Attachment 1, Volume 11, Rev. 0, Page 170 of 431
Attachment 1, Volume 11, Rev. 0, Page 171 of 431 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 11, Rev. 0, Page 171 of 431
(7 C C ITS 3.6.1.7 ITS 0 ITS
, 3.0 UMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS
- 4. Pressure Suppression Chamber-Drywell Vacuum 4. Pressure Suppression Chamber-Drywell Vacuum 03 a) Breakers Breakers 0 Applicablity 0
- a. Operability and full closure of the 0CD LCO 3.6.1.7 drywellsupprassion charnber vacuum breakers CD shall be verifled by parformanace of the/
a 0 S SR 3.6.1.7.1 Note 1 (1) Monthly each operable drywell-suppresslon 0 chamber vacuum breaker shall be exercised through an opening-closing S ACTION B cycle. r 0 See ITS 3.6.1.11 tD l . _ . . , tD 3 (2) Once each operating cycle, drywell to 0 suppression chamber leakage shall be -u -U demonstrated to be lass than that -a I eOnUient to a one-ineh diameter orfleeC I iU anc idfvacuum breff~er shl be s 03 CD _ sected. nminent ee uirade ACTION A
- -4 tD to ACTION B 0
6 tD -9' iL
.3.6.1.7.3 (4) Once each roWer eycse me vacuum _ co breakers shall be tested to determine that CD SR 3.6.1.7.1 d.
the force required to open each valve from Note 3 fully closed to fully open does not exceed that equivalent to 0.6 ng uppssionchamber aceof valv dle. (Containment access r ured.) 3.7/4.7 164 01/28/05 Amendment No. 8, 36, 80, 104, 141 Page 1 of 3
(C C ITS 3.6.1.7 C 0 ITS ITS 3.0 LIMING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REOUIREMENTS I
- e. One posrZon alarm circan be inoperable b. When the position of any drywell-suppression 0) provl rfg that the red daft position alar ACTION B chamber vacuum breaker valve Is Indicated to c rgtis operable. Both pos~tlon alarm Crcuit be not fully dosed at a time when such closure 0 ieto suppressionbe.I1 ted 0 ry be inoperagh for a period not I xceed Is req u CD 0
/seven days p eied that all vacuy breakers operabsa l'are
-.4 L02 113) die-r-en Tif demo Fa ated to be less cysha>
hatshownon 2- f. If requirements of 3.7.A.4 cannot be met the SR 3.6.1.7.1 evidence of subsequent operalion o the ACTON C second Inoperable valve until the Inoperable valve is 0 reactor shall be placed in a Cold Shutdown I condIion within hours. Frequency restored to a normal condition. 0 3 MODE3 in 12 c. When positIon a cdrcuIts are made hours and to be mnope the control panesti le, 0 36 L~~j~dicator light la us shall be record daily 1o0s -L > / detet diaaqs in the vacuum breker wslton to
;a 0 I ___________________________________________
CD Primary Containment Oxygen Concentration 5. Primary Containment Oxygen Concentration -U Whenever Inertling Is required, the primary See ITS 3.6.3.1 }
- a. The primary containment atmosphere shall be containment oxygen concentration shall be CD reduced to less than 4% oxygen by volume with measured and recorded on a weekly basis.
nitrogen gas whenever the reactor Is in the run - to mode. except as specified In 3.7.A.5.b. 0) CD
- b. Within the 24-hour period after Thermal Power CD,
_-4 Is > 15% Rated Thermal Power following C4 -9' startup, to 24 hours prior to reducing Thermal 0s Power to < 15% Rated Thermal Power prior to the next scheduled reactor shutdown, the containment atmosphere oxygen concentration shall be reduced to less than 4% by volume, and maintained In this condition. I 3.7/4.7 165 09/23/02 Amendment No. 64 130 Page 2 of 3
( C C ITS 3.6.1.7 0 w 0.4 Figure 3.7/
/ 0) 0 0 .DtFF£RENTIL PI SURE DECAY BETWEEN THE AND LL .DRYWELL WITH A SHIM HOLDtNG 0 VACUUM BREAKER .EACH Il 6 INCH OPEN AT THE CD BOTTOM. E PERFORMED 5.10.73. .
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200 191 1/9/81 Amendment No. 0 Page 3 of 3
Attachment 1, Volume 11, Rev. 0, Page 175 of 431 DISCUSSION OF CHANGES ITS 3.6.1.7, SUPPRESSION CHAMBER-TO-DRYWELL VACUUM BREAKERS ADMINISTRATIVE CHANGES A.1 In the conversion of the Monticello Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1433, Rev. 3, "Standard Technical Specifications General Electric Plants, BWR/4" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 CTS 4.7.A.4.a.(4) requires the opening setpoint of the vacuum breakers to be tested once each "operating cycle." ITS SR 3.6.1.7.3 requires a similar verification every "24 months", that each vacuum breaker opening setpoint is less than or equal to 0.5 psid. This changes the CTS by changing the Frequency from "operating cycle" to "24 months." This change is acceptable because the current "operating cycle" is "24 months." In letter L-MT-04-036, from Thomas J. Palmisano (NMC) to the USNRC, dated June 30, 2004, NMC has proposed to extend the fuel cycle from 18 months to 24 months and at the same time has performed an evaluation in accordance with Generic Letter 91-04 to extend the unit Surveillance Requirements from 18 months to 24 months. CTS 4.7.A.4.a.(4) was included in this evaluation. This change is designated as administrative because it does not result in any technical changes to the CTS. MORE RESTRICTIVE CHANGES M.1 CTS 3.7.A.4.a is applicable when Primary Containment Integrity is required. CTS 3.7.A.2.a.(1) specifies that Primary Containment Integrity is required at all times when the reactor is critical or when the reactor water temperature is above 212 0F and fuel is in the reactor vessel. ITS 3.6.1.7 is applicable in MODES 1, 2, and 3. This changes the CTS by requiring the suppression chamber-to-drywell vacuum breakers to be OPERABLE in MODE 2 when reactor water temperature is less than or equal to 212 0FR The purpose of CTS 3.7.A.4.a is to ensure the suppression chamber-to-drywell vacuum breakers are OPERABLE to prevent possible damage to the integrity of the primary containment that could result from a negative pressure in the suppression chamber or drywell in excess of the primary containment design limit. The suppression chamber-to-drywell vacuum breakers are required to be OPERABLE during MODES 1, 2, and 3 when a design basis accident could cause a release of radioactive material to the primary containment. In MODES 1 and 3, the reactor coolant temperature will always be above 212 0F. In MODE 2, the reactor coolant temperature may be less than or equal to 2120 F when the reactor is subcritical but control rods are withdrawn. Therefore, it is necessary and acceptable to require the suppression chamber-to-drywell vacuum breakers to be OPERABLE. This change is designated as more restrictive because the LCO will be applicable under more reactor operating conditions than in the CTS. Monticello Page 1 of 7 Attachment 1, Volume 11, Rev. 0, Page 175 of 431
Attachment 1, Volume 11, Rev. 0, Page 176 of 431 DISCUSSION OF CHANGES ITS 3.6.1.7, SUPPRESSION CHAMBER-TO-DRYWELL VACUUM BREAKERS M.2 ITS SR 3.6.1.7.1 requires verification that each vacuum breaker is closed every 14 days. Two Notes are included that specify the Surveillance is not required to be met: 1) the vacuum breakers are open during Surveillances; or 2) if the vacuum breakers are open when performing their intended function. This Surveillance Requirement is not included in the CTS. This changes the CTS by adding a Surveillance Requirement to verify each vacuum breaker is closed every 14 days. This change is acceptable because it helps to ensure each vacuum breaker is correctly positioned. This helps ensure a bypass leak path does not exist for the primary containment. This change is designated as more restrictive because it adds a Surveillance Requirement that is not required in the CTS. M.3 CTS 3.7.A.4.b allows a suppression chamber-to-drywell vacuum breaker to be not fully closed by indication provided drywell-to-suppression chamber differential pressure decay does not exceed values shown on Figure 3.7-1. CTS 3.7.A.4.c allows vacuum breakers to be inoperable provided a determination that the inoperable vacuum breaker is closed. Neither of these Actions provides a time to complete the determination. However, CTS 4.7.A.4.b states that when the position of any suppression chamber-to-drywell vacuum breaker is indicated to be not fully closed, "the drywell to suppression chamber differential pressure decay shall be demonstrated to be less than shown on Figure 3.7.1 immediately..." Thus, CTS 4.7.A.4.b is the requirement that ensures the provisions of CTS 3.7.A..4.b and c are met. Furthermore, CTS 4.7.A.4.b does not specify a completion time to perform the demonstration; only that it is begun "immediately." ITS 3.6.1.7 ACTION B imposes a requirement to close the open vacuum breaker in 12 hours. This changes the CTS by specifying a 12 hour Completion Time to close an open vacuum breaker and deletes the requirement to initiate to demonstration immediately. The change that moves the method of determining a vacuum breaker is closed (by performing a drywell-to-suppression chamber differential pressure decay test) is discussed in DOC LA.1. The purpose of CTS 3.7.A.4.b, 3.7.A.4.c, and 4.7.A.4.b is to ensure all vacuum breakers support the leak tight primary containment boundary. The actions to demonstrate the drywell-to-suppression chamber differential pressure decay does not exceed values shown on Figure 3.7-1 does not address the completion time for the demonstration. This change is acceptable because if the vacuum breaker position indication is not reliable, the proposed Completion Time provides adequate time to perform testing to verify the vacuum breaker is closed. Additionally, this short time is allowed to close the vacuum breaker due to the low probability of an event occurring that would pressurize the primary containment. This change is designated as more restrictive because the ITS Completion Time is less than the Completion Time that appears in the CTS. Furthermore, in order to ensure completion of the drywell-to-suppression chamber differential pressure decay test within 12 hours, action (i.e., actions to set up the equipment necessary to perform the test) must begin immediately. Thus, there is no reason to include the current "immediate" requirement and this portion of the change is an administrative change. M.4 CTS 3.7.A.4.a requires all eight suppression chamber-to-drywell vacuum breakers to be OPERABLE and closed. However, CTS 3.7.A.4.c specifies up to Monticello Page 2 of 7 Attachment 1, Volume 11, Rev. 0, Page 176 of 431
Attachment 1, Volume 11, Rev. 0, Page 177 of 431 DISCUSSION OF CHANGES ITS 3.6.1.7, SUPPRESSION CHAMBER-TO-DRYWELL VACUUM BREAKERS two suppression chamber-to-drywell vacuum breakers may be inoperable provided that: 1)the vacuum breakers are fully closed and at least one alarm circuit is OPERABLE; or 2) the vacuum breaker is secured in the closed position or replaced by a blank flange. ITS LCO 3.6.1.7 requires seven suppression chamber-to-drywell vacuum breakers to be OPERABLE for opening and eight suppression chamber-to-drywell vacuum breakers are closed. ITS 3.6.1.7 ACTION A states that when one of the seven required suppression chamber-to-drywell vacuum breakers is inoperable, it must be restored to OPERABLE status within 72 hours. This changes the CTS by increasing the number of vacuum breakers required to be OPERABLE for opening from six to seven, and providing a Completion Time of 72 hours to restore an inoperable vacuum breaker when one of the seven required vacuum breakers is inoperable. The change to the manner in which the vacuum breakers are determined to be closed is discussed in DOC L.1. The purpose of CTS 3.7.A.4.a is to ensure sufficient vacuum breaker capacity is available to assure the suppression chamber-to-drywell negative differential pressure remains below the design value. This change is acceptable because the LCO requirements continue to ensure that the required suppression chamber-to-drywell vacuum breakers are maintained consistent with the safety analyses and licensing basis. This safety analysis specifies six vacuum breakers are adequate to provide the negative pressure protection requirement for each of the three events (listed in the ITS 3.6.1.7 Bases) evaluated. The current licensing basis also only requires six vacuum breakers to be OPERABLE, since CTS 3.7.A.4.c allows unlimited operation with two of the eight vacuum breakers inoperable, provided the two inoperable vacuum breakers are closed. Thus, seven vacuum breakers are specified in ITS LCO 3.6.1.7 to ensure that, assuming a single failure of a required vacuum breaker, the remaining vacuum breakers are capable of providing the vacuum relief function. ITS 3.6.1.7 ACTION A requires that when one of the seven required vacuum breakers are inoperable, it must be restored within 72 hours. Thus, unlimited operation with the minimum assumed number of vacuum breakers OPERABLE is not allowed. Therefore, this change is designated as more restrictive because more stringent LCO Requirements and ACTIONS are being applied in the ITS than were applied in the CTS. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA.I (Type 3 - Removal of Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 3.7.A.4.a requires all suppression chamber-to-drywell vacuum breakers be closed "as indicated by the position indication system." CTS 3.7.A.4.b states that a suppression chamber-to-drywell vacuum breaker may be nonfully closed 'as indicated by the position indication and alarm system provided that drywell to suppression chamber differential pressure decay does not exceed that shown on Figure 3.7.1." CTS 4.7.A.4.b states that when Monticello Page 3 of 7 Attachment 1, Volume 11, Rev. 0, Page 177 of 431
Attachment 1, Volume 11, Rev. 0, Page 178 of 431 DISCUSSION OF CHANGES ITS 3.6.1.7, SUPPRESSION CHAMBER-TO-DRYWELL VACUUM BREAKERS the position of any suppression chamber-to-drywell vacuum breaker is indicated to be not fully closed, "the drywell to suppression chamber differential pressure decay shall be demonstrated to be less than shown on Figure 3.7.1." ITS 3.6.1.7 does not include the details for determining how the vacuum breakers are to be determined closed. This changes the CTS by relocating these details to the ITS Bases. The removal of these details for ensuring the vacuum breakers are closed from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement to maintain close or close the vacuum breakers if open. Also, this change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5 of the ITS. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed. LA.2 (Type 3 - Removal of Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 4.7.A.4.a.(4) requires a verification that the vacuum breaker opening force not exceed an equivalent of 0.5 psid "acting on the suppression chamber face of the valve disc. (Containment access required.)" ITS SR 3.6.1.7.3 continues to require a verification that the vacuum breaker opening force is < 0.5 psid, but the location of the force (acting on the suppression chamber face of the valve disc) is not specified. In addition, the information that performance of the SR requires containment access is also not specified. This changes the CTS by relocating these details to the ITS Bases. The removal of these details for the location of the opening force and that containment access is required to perform the test from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the allowance to verify the opening force of the vacuum breakers. Also, this change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5 of the ITS. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed. LA.3 (Type 6 - Removal of LCO, SR, or other TS requirement to the TRM, USAR, ODCM, OQAP, IST Program, or I1P) CTS 4.7.A.4.a.(2) states, in part, that "Once each operating cycle ... each vacuum breaker shall be visually inspected." ITS 3.6.1.7 does not include this requirement. This changes the CTS by relocating this Surveillance Requirement to the Inservice Testing Program. The removal of this Surveillance Requirement from the Technical Specifications is'acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and Monticello Page 4 of 7 Attachment 1, Volume 11, Rev. 0, Page 178 of 431
Attachment 1, Volume 11, Rev. 0, Page 179 of 431 DISCUSSION OF CHANGES ITS 3.6.1.7, SUPPRESSION CHAMBER-TO-DRYWELL VACUUM BREAKERS safety. ITS 3.6.1.7 still requires the suppression chamber-to-drywell vacuum breakers to be OPERABLE, and includes Surveillance Requirements that ensure the vacuum breakers are closed and open at the prescribed differential pressure. Also, this change is acceptable because this type of Surveillance Requirement will be adequately controlled in the Inservice Testing Program, which is controlled by 10 CFR 50.55a. This change is designated as a less restrictive removal of detail change because a requirement is being removed from the Technical Specifications. LA.4 (Type 3 - Removal of Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 3.7.A.4.d allows the suppression chamber-to-drywell vacuum breakers to be cycled, one at a time, during containment inerting and de-inerting operations "to assist in purging air or nitrogen from the suppression chamber vent header." ITS SR 3.6.1.7.1 Note 3 continues to provide the same allowance, but does not specify the purpose of the allowance. This changes the CTS by relocating this detail to the ITS Bases. The removal of this detail for allowing the suppression chamber-to-drywell vacuum breakers to be cycled, one at a time, during containment inerting and deinerting operations from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the allowance to cycle the vacuum breakers. Also, this change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5 of the ITS. This program provides for the evaluation of changes to ensure the Bases are properly controlled; This change is designated as a less restrictive removal of detail change because information relating to system design is being removed. LESS RESTRICTIVE CHANGES L.1 (Category 4 - Relaxation of Required Action) CTS 3.7.A.4.c allows two of the eight suppression chamber-to-drywell vacuum breakers to be inoperable, provided that: 1) the vacuum breakers are fully closed and at least one alarm circuit is OPERABLE; or 2) the vacuum breaker is secured in the closed position or replaced by a blank flange. ITS 3.6.1.7 continues to require all eight suppression chamber-to-drywell vacuum breakers to be closed and provides actions for when a vacuum breaker is inoperable for opening (ITS 3.6.1.7 ACTION A), however, the ITS does not include the additional requirements that when a suppression chamber-to-drywell vacuum breaker is inoperable for opening, at least one alarm circuit is OPERABLE for the inoperable vacuum breaker or the Inoperable vacuum breaker is secured in the closed position or replaced by a blank flange. This changes the CTS by deleting these additional requirements when a suppression chamber-to-drywell vacuum breaker is inoperable for opening. The purpose of CTS 3.7.A.4.c is to ensure the inoperable suppression chamber-to-drywell vacuum breaker is not also open, thus affecting the pressure suppression function of the primary containment. This change is acceptable Monticello Page 5 of 7 Attachment 1, Volume 11, Rev. 0, Page 179 of 431
Attachment 1, Volume 11, Rev. 0, Page 180 of 431 DISCUSSION OF CHANGES ITS 3.6.1.7, SUPPRESSION CHAMBER-TO-DRYWELL VACUUM BREAKERS because the LCO requirements continue to ensure that the required suppression chamber-to-drywell vacuum breakers are maintained closed consistent with the safety analysis assumptions. The added requirements are unnecessary since the inability of the vacuum breaker to open does not affect the vacuum breakers ability to be verified closed. In addition, ITS SR 3.6.1.7.1 requires a 14 day verification that all vacuum breakers are closed. This SR will provide adequate verification of vacuum breaker position. This change is also acceptable because ITS 3.6.1.7 ACTION A is consistent with the safe operation under the specified Condition, considering the OPERABLE status of the redundant systems or features. This includes the capacity and capability of remaining systems or features, a reasonable time for repairs or replacement, and the low probability of a DBA occurring during the allowed Completion Time. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. L.2 (Category I - Relaxation of LCO Requirement) CTS 3.7.A.4.e, 4.7.A.4.a.(3), and 4.7.A.4.c specify requirements for vacuum breaker position alarm circuits. ITS 3.6.1.7 does not include position indication-only or alarm circuits as a requirement for vacuum breaker OPERABILITY. This changes the CTS by deleting the requirements for vacuum breaker position indication-only and alarm circuits. The purpose of CTS 3.7.A.4.e, 4.7.A.4.a.(3), and 4.7.A.4.c is to help ensure that the vacuum breakers are in the closed position. However, the vacuum breaker position indication instrumentation does not necessarily relate directly to the respective system OPERABILITY. The BWR ISTS, NUREG-1433, Revision 3, does not specify indication-only or alarm equipment to be OPERABLE to support OPERABILITY of a system or component. Control of the availability of, and necessary compensatory activities if not available, for indications and alarm instruments are addressed by plant operational procedures and policies. Vacuum breaker position is required to be known to be able to satisfy the ITS 3.6.1.7 Surveillance Requirements (SR 3.6.1.7.1, SR 3.6.1.7.2, and SR 3.6.1.7.3) for the vacuum breakers. If position indication and alarm is not available and vacuum breaker position can not be determined, then the Surveillance Requirements cannot be satisfied and the appropriate actions must be taken for inoperable vacuum breakers in accordance with the ACTIONS of ITS 3.6.1.7. As a result, the requirements for the vacuum breaker position indication are adequately addressed by the requirements of ITS 3.6.1.7 and the associated SRs and their deletion from the Technical Specifications is acceptable. This change is designated as less restrictive because less stringent LCO requirements are being applied in the ITS than were applied in the CTS. L.3 (Category 3 - Relaxation of Completion Time) CTS 3.7.A.4.f requires the unit to be placed in the cold shutdown condition within 24 hours if the requirements of CTS 3.7.A.4 are not met. ITS 3.6.1.7 ACTION C requires the unit be in MODE 3 in 12 hours and in MODE 4 in 36 hours. This changes the CTS by requiring the unit to be in MODE 3 in 12 hours and by extending the time to be in cold shutdown (i.e., MODE 4) from 24 hours to 36 hours. Monticello Page 6 of 7 Attachment 1, Volume 11, Rev. 0, Page 180 of 431
Attachment 1, Volume 11, Rev. 0, Page 181 of 431 DISCUSSION OF CHANGES ITS 3.6.1.7, SUPPRESSION CHAMBER-TO-DRYWELL VACUUM BREAKERS The purpose of CTS 3.7.A.4.f is to place the unit outside the Applicability of the Specification within a reasonable amount of time. This change is acceptable because the Completion Time is consistent with the safe operation under the specified Condition, considering the OPERABLE status of the redundant systems or features. This includes the capacity and capability of remaining systems or features, a reasonable time for repairs or replacement, and the low probability of a DBA occurring during the allowed Completion Time. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. This change is also acceptable because it requires the unit to be in an intermediate condition (MODE 3) sooner than is currently required (12 hours versus 24 hours). This portion of the change reduces the amount of time the unit would be allowed to continue to operate in MODES I and 2 once the condition is identified. The consequences of a pressurization event are significantly reduced when the reactor is shutdown and a controlled cooldown is already in progress. This change is designated as less restrictive because additional time is allowed to place the unit outside the LCO Applicability than is allowed in the CTS. Monticello Page 7 of 7 Attachment 1, Volume 11, Rev. 0, Page 181 of 431
Attachment 1, Volume 11, Rev. 0, Page 182 of 431 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 182 of 431
Attachment 1, Volume II, Rev. 0, Page 183 of 431 Suppression Chamber-to-Drywell Vacuum Breakers 3.6.1E crs
~ 0 3.6 CONTAINMENT SYSTEMS 3.7.A.4 3.6.1. Suppression Chamber-to-Drywell Vacuum Breakers 0 3.7.A4.a LCO 3.6.1.f I suppression chamber-to-drywell vacuum breakers shall be OPERABLE for opening.
0 AND suppression chamber-to-drywell vacuum breakers shall be 0 closed except when peiTfr intended TuncZTor. 0 APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 3.7.A.4.c A. One required A.1 Restore one vacuum 72 hours suppression chamber-to- breaker to OPERABLE drywell vacuum breaker status. inoperable for opening. 0 1- t
.l12 3.7-A.4.b,B. One suppression B.1 Close the open vacuum 3.7A4.c, chamber-to-drywell breaker.
4.7.A4.b vacuum breaker not closed.
+ 4 3.71A.4.f. C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time not met. AND C.2 Be in MODE 4. 36 hours BWR/4 STS 3.6.1.8-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 183 of 431
Attachment 1, Volume 11, Rev. 0, Page 184 of 431 Suppression Chamber-to-Drywell Vacuum Breakers 3.6.1 . (0 CTS SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DOC SR 3.6.1.f.1 E}) ~-NO40 -- _ M.2. 3.7.A.4d Not required to be met for vacuum breakers that are
- oDen during Surveillances.
Verify each vacuum breaker is closed. 14 days
- 3. No required to be met for vacuum breakers being cycfed, one at a time, during primary containment Inerting and de-Inerting operations. AND.
2.Not required to be met for vacuum breakers performing their Intended function. Within eA'- hours disch rge of/ stea to the suppression/ cha'mber fror, the s fety/relief lvo any operation that causes the drywell-to-suppression chamber differential pressure to be reduced by
ŽP'.5j psid r
/
ff any vacuum breaker position indicator does not Indicate Lclosed BWR/4 STS '3.6.1.8-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 184 of 431
Attachment 1, Volume 11, Rev. 0, Page 185 of 431 Suppression Chamber-to-Drywell Vacuum Breakers
- RRSt 3.6SL ()
CTS SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY 4.7A.4.a.(1)SR 3.6.1 .2 Perform a functional test of each required vacuum breaker. 31 days 0 Wi in I hours 0 fol0win an o erati that ckuse any of e 9acuu brea rs tooD n 4.7A.4.a.(4) SR 3.6.1,.3 Verify the opening setpoint of each required vacuum breaker is 40.1 psid. k4J months 00 BWR/4 STS 3.6.1.8-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 185 of 431
Attachment 1, Volume 11, Rev. 0, Page 186 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.1.7, SUPPRESSION CHAMBER-TO-DRYWELL VACUUM BREAKERS
- 1. ISTS 3.6.1.8 is renumbered as ITS 3.6.1.7 since ISTS 3.6.1.4, "Drywell Pressure," is not included in the Monticello ITS.
- 2. The brackets have been removed and the proper plant specific information/value has been provided.
- 3. A portion of the second part of the LCO statement ("except when performing their intended function") has been moved to ITS SR 3.6.1.7.1 in the form of a Note.
Placing this allowance in a Surveillance Requirement Note is consistent with a similar allowance in BWR/4 ISTS SR 3.6.1.7.1 for the reactor building-to-suppression chamber vacuum breakers. This change is also consistent with a similar change allowed by the NRC in the Dresden, Quad Cities, and FitzPatrick ITS conversions. Due to this addition, the existing Note to ITS SR 3.6.1.7.1 has been numbered.
- 4. Note 3 has been added to ITS SR 3.6.1.7.1, consistent with a current licensing basis allowance (shown in CTS 3.7.A.4.d).
- 5. The second Frequency to ISTS SR 3.6.1.8.1 requires the vacuum breakers to be verified closed after they may have been opened. However, CTS 4.7.A.4.b only requires this situational verification that the vacuum breakers are closed if there is evidence of operation of the vacuum breakers and a vacuum breaker position indication is not indicating closed. In addition, the CTS does not specify how soon after vacuum breaker operation the verification must be made. Thus, ITS SR 3.6.1.7.1 second Frequency has been changed to "Within 12 hours after any operation that causes the drywell-to-suppression chamber differential pressure to be reduced by > 0.5 psid if any vacuum breaker position indicator does not indicate closed." This ISTS Frequency is not needed and has not been included in ITS SR 3.6.1.7.1 if the vacuum breaker position indicators are showing the vacuum breakers to be closed. There are many other instances where valves are required to be closed, and verified closed on a periodic basis. If these other valves are cycled (e.g., ECCS valves) plant administrative controls ensure they are left in the correct position; a special Frequency of the Surveillance Is not required. In addition, these vacuum breakers have local position indication with alarms in the control room, which are monitored by control room operators. If conditions exist for the vacuum breakers to be potentially opened (e.g., venting the drywell), control room operators would be alert to the possibility and ensure the vacuum breakers were closed at the completion of the evolution. Thus, the only time that the verification is necessary is when the vacuum breakers open and their indicators are not functioning properly. In this case, a verification needs to be performed that the vacuum breakers re-closed after opening. As stated in the ITS Bases, this verification consists of a differential pressure test, and the time necessary to perform the test is approximately 12 hours.
This 12 hour time is also consistent with the proposed Completion Time for Required Action B.1 (see JFD 7).
- 6. The second and third Frequencies to ISTS SR 3.6.1.8.2 require a functional test of the vacuum breakers (i.e., cycle the vacuum breakers) within 12 hours after an operation (i.e., S/RV opening) that may have caused them to cycle, or after the vacuum breakers have cycled. In a September 8, 1992 memorandum to C.I. Grimes from C.E. McCracken, the only basis for this Frequency is given as..."in case the event caused damage to one or more vacuum breakers." Since the vacuum breakers are designed to operate and are assumed to function after a LOCA Monticello Page 1 of 2 Attachment 1, Volume 11, Rev. 0, Page 186 of 431
Attachment 1, Volume 11, Rev. 0, Page 187 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.1.7, SUPPRESSION CHAMBER-TO-DRYWELL VACUUM BREAKERS blowdown, their operation as designed after some minor steam release from the S/RVs or change in internal pressure should not raise questions regarding immediate OPERABILITY of the vacuum breakers. Furthermore, steam quenching from the discharge of an S/RV is enhanced by the T-quenchers. Steam discharged to the torus, resulting in increased wetwell pressure and vacuum breaker opening, may pose a long term equipment degradation, rather than any immediate OPERABILITY concern. The 12 hour Frequency would be meaningless to detect long term degradation, while the normal 31 day Frequency would more than suffice for this
. concern. In addition, local position indication and redundant control room indication are provided for each vacuum breaker, as well as a common control room alarm, such that the control room operators would be alerted to the possibility of a stuck open vacuum breaker and would take the appropriate action (e.g., close the vacuum breaker) to ensure isolation capability is maintained. -Therefore, this Frequency, which is not required in the current Technical Specifications for Monticello, has not been added to the Monticello ITS.
- 7. ISTS 3.6.1.8 Required Action B.1 requires an open suppression chamber-to-drywell vacuum breaker to be closed within 2 hours. As stated in the ISTS 3.6.1.8 Bases, an alternate method of verifying the vacuum breakers are closed is to perform a differential pressure test between the suppression chamber and drywell. Monticello has reviewed the time it would take to perform this type of test, which is also currently allowed in CTS 4.7.A.4.b, and has determined that it would take approximately 12 hours. Therefore, a 12 hour Completion Time has been proposed
-for ITS 3.6.1.7 Required Action B.1. This proposed time is more restrictive than is currently required, since no finite time to perform this test is provided in the CTS (see ITS 3.6.1.7 Discussion of Change M.3).
Monticello Page 2 of 2 Attachment 1, Volume 11, Rev. 0, Page 187 of 431
Attachment 1, Volume 11, Rev. 0, Page 188 of 431 Improved Standard Technical Specifications (ISTS) Bases Markup' and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 188 of 431
Attachment 1, Volume 11, Rev. 0, Page 189 of 431 Suppression Chariber-to-Drywell Vacuum Breakers INT B 3.63C B 3.6 CONTAINMENT SYSTEMS B 3.6.1sp Suppression Chamber-to-Drywell Vacuum Breakers (0 BASES BACKGROL IND The function of the suppressionrchamber-to-drywell vacuum breakers is (C to relieve vacuum in the drywell. There ar internal vacuum breakers (2 located on the vent header of the vent system between the drywell and the suppression chamber, which allow air and steam flow from the suppression chamber to the drywell when the drywell is at a negative pressure with respect to the suppression chamber. Therefore, suppression chamber-to-drywell vacuum breakers prevent an excessive Fsuppression ciiamber- l negative differential pressure across there drywell boundary. Each vacuum breaker isa self actuating valve, similar to a check valve, which can be remotely operated for testing purposes. A negative differential pressure across the drywell wall is caused byE4afi depressurization of the drywell. Events that cause thisgraid 0 depressurization are cooling cycles, inadvertent drywell spray actuation, and steam condensation from sprays or subcooled water reflood of a break in the event of a primary system rupture. Cooling cycles result in minor pressure transients in the drywell that occur slowly and are normally controlled byhe an ventilation equipment. Spray actuation or spill of subcooled water out of a break results in more 0 significant pressure transients and becomes important in sizing the internal vacuum breakers. In the event of a primary system rupture, steam condensation within the drywell results in the most severe pressure transient. Following a primary system rupture, air in the drywell is purged into the suppression chamber free airspace, leaving the drywell full of steam. Subsequent condensation of the steam can be caused in two possible ways, namely, Emergency Core Cooling Systems flow from a recirculation line break, or drywell spray actuation following a loss of coolant accident (LOCA). These two cases determine the maximum depressurization rate of the drywell. In addition, the waterleg in the Mark I Vent System downcomer is controlled by the drywell-to-suppression chamber differential pressure. If the drywell pressure is less than the suppression chamber pressure, there will be an increase in the vent waterleg. This will result in an increase in the water clearing inertia Inthe event of a postulated LOCA, resulting in an increase in the peak drywell pressure. This in turn will result in an increase in the pool swell dynamic loads. The internal vacuum breakers limit the height of the waterleg in the vent system during normal operation. BWR/4 STS B 3.6.1.8-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 189 of 431
Attachment 1, Volume 11, Rev. 0, Page 190 of 431 Suppression Chamber-to-Drywell Vacuum Breakers B 3.6.1.V1 (0 BASES APPLICABLE Analytical methods and assumptions involving the suppression chamber-SAFETY to-drywell vacuum breakers are presented in Reference 1 as part of the ANALYSES accident response of the primary containment systems. Internal (suppression chamber-to-drywell) and external (reactor building- to-suppression chamber) vacuum breakers are provided as part of the primary containment to limit the negative differential pressure across the drywell and suppression chamber walls that form part of the primary containment boundary. The safety analyses assume that the internal vacuum breakers are closed are fully open at a differential pressure ofA0.fljpsid (Ref. 1). 0 zof the internal vacuum breakers are assumed to fail in a p o (e 1). The results of the analyses show that the design 0 pressure is not exceeded even under the worst case accident scenario. The vcuum breaker opening differential pressure setpoint and the eiht requirement thaI o0 o1 tfaced vacuum breakers be OPERABLE are a result on the vacuum breakers to limit the vent 0
- system waterleg height. IThe total cross sectional area of the main ve based on the Bodega Bay s sysem eeen terywell and suppre:ion chamber needed to fuil Pressure Suppression System tests. These tests were this requirement has been establishe as a minimum of [51.5] times the total break areaZRef. 1). In turn, t vacuum relief capacity be een the 0D break LOCAwhich tends to cause downcomer water level drywell and supression chambe should be [1/16] of the tot main vent cross secti al area, with the v yes set to operate at [0.5] sid asme 0
variations, as a preliminary step differenti Dressure. Design Basis Accident (DBA) analyses the in the large rupture test sequence. The vacuum breaker capacity selected ismore than vacuum breakers [RH the suppression to be closed initially and to remain closed and leak
- o-at a positive pressure relative to the drywell.
tightm 0 adequate to limit the pressure ram- e.r7, differential between the suppression chamber and The suppression chamber-to-drywell vacuum breakers satisfy Criterion 3 rywell. post LOCA. v of 10 CFR 50.36(c)(2)(ii). seven LCO tJD e On' of th vacuum breakers must b OPERABLE for opening. lhtsuppression chamber-to-drywell vacuu reakers, however, are 0 or when a vacuum required to be closed (except during testing when the vacuum breakers breaker isopen during are performing their intended design function). The vacuum breaker Inettingor de4nertrng OPERABILITY requirement provides assurance that the drywell-to-operations suppression chamber negative differential pressure remains below the design value. The requirement that the vacuum breakers be closed ensures that there is no excessive bypass leakage should a LOCA occur. APPLICABILITY In MOD ,2, and 3, the SuppRr essra S stem is required to Ibe..OERABLE to mitigatet heffsof a DBA.IExcessive negative l pressure Inside the drywell could occur due to Inadvertent actuation of 7 this system. _He vacuum breakers, thert, arequired to be / / 9 OPERAB in MODES 1,2, and 3, wh he Suppression Pool ray Syst s rquired to be OPERABL miiaeth
~ effect
,dertent actuation of the Sup sion Pool Spray Sysn/
Move to B 3.6.1.8-3 as INSERT I BWR14 STS B 3.6.1.8-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 190 of 431
Attachment 1, Volume 11, Rev. 0, Page 191 of 431 Suppression Chamber-to-Drywell Vacuum Breakers
- ~~B 3.6.1 -
BASES APPLICABILITY (continued) E951)n MODES 1, 2, and 3, a DBA could result in excessive negative differential pressure across the drywell wall, caused by the rapid which, afterthe depressurization of the drywell. The event that results in the limiting rapid suppression chamber-to- depressurization of the drywell is the primary system rupture that purges open (due to excessive the drywell of air and fills the drywell free airspace with steam. betweentheisuppression Subsequent condensation of the steam would result in depressurization chamber and drywell), of the drywel. The limiting pressure and temperature of the primary INSERT 1from depressurlationofthe system prior to a DBA occur in MODES 1, 2, and 3.3 suppression chamberl In MODES 4 and 5, the probability and consequences of these events are reduced by the pressure and temperature limitations in these MODES; therefore, maintaining suppression chamber-to-drywell vacuum breakers OPERABLE is not required in MODE 4 or 5. ACTIONS A.1 With one of the required vacuum breakers inoperable for opening (e.g., the vacuum breaker is not open and may be stuck closed or not within its opening setpoint limit, so that it would not function as designed during an L= event that depressurized the drywell), the remainin gnOPERABLE () vacuum breakers are capable of providing the vacuum relief function. However, overall system reliability is reduced because a single failure in one of the remaining vacuum breakers could result in an excessive seven _suppression chamber-to-drywell differential pressure during a DBA. 0 Therefore, with one of th jey required vacuum breakers inoperable, 72 hours is allowed to restore at least one of the inoperable vacuum breakers to OPERABLE status so that plant conditions are consistent with those assumed for the design basis analysis. The 72 hour Completion Time is considered acceptable due to the low probability of an event in which the remaining vacuum breaker capability would not be adequate. B.1 An open vacuum breaker allows communication between the drywell and suppression chamber airspace, and, as a result, there is the potential for suppression chamber overpressurization due to this bypass leakage if a LOCA were to occur. Therefore, the open vacuum breaker must be closed. A short time is allowed to close the vacuum breaker due to the low probability of an event that would pressurize primary containment. If vacuum breaker position indication is not reliable, an alternate method o the ay g that the vacuum breakers are closed is to verify that fdifferential
' pressure f0psi between the suppression chamber and drywell is maintainec tor 1 hoursythUWi makeuDI The required hour Completion
/ Time is considered adequate to perform this test. 1 within the Allowable Region of Figure B 3.6.1.7-1. The Figure was originally developed from a test performed with a shim holding each vacuum breaker 1116 Inch open at the bottom. 1 BWR/4 STS B 3.6.1.8-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 191 of 431
Attachment 1, Volume 11, Rev. 0, Page 192 of 431 Suppression Chamber-to-Drywell Vacuum Breakers B 3.6. BASES ACTIONS (continued) C.1 and C.2 If the inoperable suppression chamber-to-drywell vacuum breaker cannot be closed or restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 0i) REQUIREMENTS Each vacuum breaker is verified closed to ensure that this potential large bypass leakage path is not present. This Surveillance is performed by th deca oif t vacuum breaker position indication or by verifying that e'-;' differential pressure sdi between the suppression chamber and ( ) within the Allowable drywell is maintainesor 1 hour ut makeup. The 14 day Frequency Figure Of3. is based on engineering judgment, is considered adequate in view of other indications of vacuum breaker status available to operations personnel, and has been shown to be acceptable thro operating 12 experience. This verification is also required within t2,ours a en discharge of o the suppression chamber the safety/relief valv ni operation that causes the drywell-to-suppression chamber differential pressure to be reduced by 2gp.5jpsid. o 'N~o~tgaddNo this SR R allows suppression chamber-to-drywell vacuum breakers opened in conjunction with the performance of a Surveillance to not be considered as failing this SR. These periods of opening vacuum breakers are controlled by plant procedures and do not represent inoperable vacuum breakers. The second Note Is included to clarify that vacuum breakers open due to an actual differenllal pressure are not considered as failing this SR. The third Note Is Included to 0 Dclarify that vacuum breakers open, one at a time, during primary containment Inerting or SR 3.6.1 t 2 de-lnerting operations are not considered as failing this SR. This allowance Is necessary
,to assist In purging air or nitrogen from the suppression chamber vent header. -)
Each required vacuum breaker must be cycled to ensure that it opens adequately to perform its design function and returns to the fully closed position. This ensures that the safety analysis assumptions are valid. The 31 day Frequency of this SR was developed, based on Inservice Testing Program requirements to perform valve testing at least once BWR14 STS B 3.6.1.8-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 192 of 431
Attachment 1, Volume 11, Rev. 0, Page 193 of 431 Suppression Chamber-to-Drywell Vacuum Breakers B 3.6.1.g 0 BASES SURVEILLANCE REQUIREMENTS (continued) every 92 days. A 31 day Frequency was chosen to provide additional assurance that the vacuum breakers are OPERABLE, since they are located in a harsh environment (the suppression chamber airspace). i addition, this fctional test is requiredw in 12 hours after eith 0 discharg steam to the suppr on chamber from the s y/relief valv or after an operati at causes any of the vac breakers to pen, SR 3.6.1f.3 -
,--F7-1 0
3 Verification of the vacuum breaker opening setpoint is necessary to (actng on the suppression ensure that the safety analysis assumption regarding vacuum breaker full ,fl chamber fce of the valve disc) open differential pressure ofRO.5 psia'is valid. The ft month Frequency () is based on the need to perform this Surveillance under the conditions that-apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. For this facility, the [ onth Frequency has been shown to be acceptable, based on operating experience, and is further justified because of other surveillances performed at shorter Frequencies that convey the proper functioning status of each vacuum breaker. REFERENCES .t 1. 'SAR, Section L[Iw 5.2.1.2.3 0 0
.4~ 0 BWR/4 STS B 3.6.1.8-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 193 of 431
Attachment 1, Volume 11, Rev. 0, Page 194 of 431 B 3.6.1.7 (0 INSERT 2 0.4 I I I I I I 0.3. I I NI C,, U) 0.2 llowable R IL I I N 11-1 I I 4-W N a) (1. tr_ 0.1 I 11N. im
-11 I
0.0 I i I 0 50 100 150 200 Time (sec) Figure B 3.6.1.7-1 (Page 1 of 1) Drywell-Suppression Chamber Differential Pressure Decay Insert Page B 3.6.1.8-6 Attachment 1, Volume 11, Rev. 0, Page 194 of 431
Attachment 1, Volume 11, Rev. 0, Page 195 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.1.7 BASES, SUPPRESSION CHAMBER-TO-DRYWELL VACUUM BREAKERS
- 1. Changes have been made to reflect changes made to the Specification.
- 2. The brackets have been removed and the proper plant specific information/value has been provided.
- 3. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the plant specific nomenclature, number, reference, system description, analysis description, or licensing basis description.
- 4. Inadvertent actuation of the suppression pool spray system is not the main concern for depressurizing the drywell, a LOCA inside the drywell is the main concern. In addition, actuation of suppression pool spray is not a major concern in determining the adequacy of the vacuum breakers at Monticello, the drywell spray is the more limiting system. Therefore, this section has been revised to place emphasis on the proper events.
- 5. Typographical error corrected.
- 6. Change made for clarity. The term "rapid" does not apply to cooling cycles, since the description of cooling cycles states that it occurs slowly.
Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 195 of 431
Attachment 1, Volume 11, Rev. 0, Page 196 of 431 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 11, Rev. 0, Page 196 of 431.
Attachment 1, Volume 11, Rev. 0, Page 197 of 431 DETERMINATION OF NO SIGNIFiCANT HAZARDS CONSIDERATIONS ITS 3.6.1.7, SUPPRESSION CHAMBER-TO-DRYWELL VACUUM BREAKERS There are no specific NSHC discussions for this Specification. Monticello Page 1 of I Attachment 1, Volume 11, Rev. 0, Page 197 of 431
Attachment 1, Volume 11, Rev. 0, Page 198 of 431 ATTACHMENT 8 ITS 3.6.1.8, Residual Heat Removal (RHR) Drywell Spray Attachment 1, Volume 11, Rev. 0, Page 198 of 431
Attachment 1,Volume 11, Rev. 0, Page 199 of 431 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 11, Rev. 0, Page 199 of 431
( C C ITS 3.6.1.8 0 ITS ITS 4.0 SURVEILLANCE REOUIREMENTS 3.6.1.8 System LCO 3.6.1.8 1. Demonstrate the operabiity of the drywell spray C) 3.6.1.8.2 headers and nozzles with an air test during each D 0 10 year period. Applicability {See IS 3.A.2.3 - CD 0
-[ See ITS 3.7.1 }
ZA
-.
- -1 0 0 ACTiON A 2. One Containment Sprayh~ioSubsystem may be *See ITS 3.6.2.3}
Inoperable for 7 days
-9,
-A 0)
CD~ ACTION C 3. If the requirements of 3.5.C.1 or 2 cannot be met, an orderly shutdown of the reactor wilt be Initiated -o and the reactor waler temperature shall be reduced to less than 212F within CR hours. C) I0
-4' (A) 3.514.5 104 08/01/01 Amendment No. 27 77, 7. 0, 102,122 Page 1 of 1
Attachment 1, Volume 11, Rev. 0, Page 201 of 431 DISCUSSION OF CHANGES ITS 3.6.1.8, RESIDUAL HEAT REMOVAL (RHR) DRYWELL SPRAY ADMINISTRATIVE CHANGES A.1 In the conversion of the Monticello Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1433, Rev. 3, "Standard Technical Specifications General Electric Plants, BWR/4" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 CTS 3.5.C.1 Footnote *, which states "For allowed out of service times for the RHR pumps see Section 3.5.A," is a cross reference to another Specification that provides additional requirements associated with the RHR pumps. This cross reference is not included in ITS 3.6.1.8. This changes the CTS by deleting the cross reference to other Specification requirements. The purpose of CTS 3.5.C is to ensure the OPERABILITY of the RHR drywell spray subsystems. ITS 3.6.1.8 does not include any cross references to other Specifications that govern the OPERABILITY requirements for the RHR pumps. This change is acceptable since the other Specifications prescribe the appropriate requirements for the RHR pumps and this cross reference is not necessary. This change is considered administrative because it does not result in any technical changes to the CTS. MORE RESTRICTIVE CHANGES M.1 CTS 3.5.C.1 is applicable when irradiated fuel is in the reactor vessel and reactor water temperature is greater than 212 0F. ITS LCO 3.6.1.8 is applicable in MODES 1, 2, and 3. This changes the CTS by requiring two RHR drywell spray subsystems to be OPERABLE in MODE 2 when reactor water temperature is less than or equal to 2120F. The purpose of CTS 3.5.C.1 is to ensure the RHR drywell spray subsystems are OPERABLE to mitigate the consequences of a design basis accident. The RHR drywell spray subsystems are required to be OPERABLE during MODES 1, 2, and 3 when there is considerable energy in the reactor core and a DBA could significantly increase the drywell temperature. In MODES 1 and 3, the reactor coolant temperature will always be above 212 0F. In MODE 2, the reactor coolant temperature may be less than or equal to 2120 F when the reactor is subcritical but control rods are withdrawn. Therefore, it is necessary and acceptable to require the RHR drywell spray subsystems to be OPERABLE. This change is designated as more restrictive because the LCO will be applicable under more reactor operating conditions than in the CTS. M.2 Currently, the CTS does not provide any specific Surveillance Requirement to verify the alignment of the RHR drywell spray subsystems. ITS SR 3.6.1.8.1 requires verification that each RHR drywell spray subsystem manual and power operated valve in the flow path that is not locked, sealed, or otherwise secured in Monticello Page 1 of 3 Attachment 1, Volume 11, Rev. 0, Page 201 of 431
Attachment 1, Volume 11, Rev. 0, Page 202 of 431 DISCUSSION OF CHANGES ITS 3.6.1.8, RESIDUAL HEAT REMOVAL (RHR) DRYWELL SPRAY position is in the correct position or can be aligned to the correct position every 31 days. This changes the CTS by adding this Surveillance Requirement to the Technical Specifications. The purpose of ITS SR 3.6.1.8.1 is to provide assurance that the proper flow paths will exist for drywell spray operation. This change is acceptable because it provides additional assurance that each RHR drywell spray subsystem w(ill be capable of performing its function. This change is designated as more restrictive because it adds Surveillance Requirements to the CTS. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA. 1 (Type I - Removing Details of System Design and System Description, Including Design Limits) CTS 3.5.C.1 states that a containment spray subsystem consists of the following equipment powered from one division: 1 RHR Heat Exchanger, 1 RHR Pump, and valves and piping necessary for drywell spray. ITS 3.6.1.8 requires two RHR drywell spray subsystems to be OPERABLE, but the details of what constitutes an OPERABLE subsystem are moved to the ITS Bases. This changes the CTS by moving the details of what constitutes an OPERABLE subsystem to the ITS Bases. The removal of these details, which are related to system design, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement for two RHR drywell spray subsystems to be OPERABLE. Also, this change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5 of the ITS. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because information. relating to system design is being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES L.I (Category 4 - Relaxation of Required Action) When two RHR containment spray (i.e., drywell spray) subsystems are inoperable, a plant shutdown is required by CTS 3.5.C.3; no time is provided to restore a subsystem. With two RHR drywell spray subsystems inoperable, ITS 3.6.1.8 ACTION B will allow 8 hours to restore one inoperable RHR drywell spray subsystem prior to requiring a plant shutdown. This changes the CTS by allowing 8 hours to restore one of two inoperable RHR drywell spray subsystems prior to requiring a plant shutdown. Monticello Page 2 of 3 Attachment 1, Volume 11, Rev. 0, Page 202 of 431
Attachment 1, Volume 11, Rev. 0, Page 203 of 431 DISCUSSION OF CHANGES ITS 3.6.1.8, RESIDUAL HEAT REMOVAL (RHR) DRYWELL SPRAY The purpose of CTS 3.5.C isto require sufficient drywell spray to ensure the primary containment conditions for the safety analyses are met. The proposed Completion Time of 8 hours is acceptable because it provides some time to restore one of the subsystems prior to requiring a shutdown, yet is short enough that it does not significantly increase the probability of an accident to occur during this additional time. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. L.2 (Category 3 - Relaxation of Completion Time) CTS 3.5.C.3 requires the plant to be shut down and reactor water temperature reduced to less than 2120 F within 24 hours if the requirements of CTS 3.5.C.1 or CTS 3.5.C.2 are not met. Under similar conditions (as modified by DOC L.1), ITS 3.6.1.8 ACTION C requires the reactor be in MODE 3 in 12 hours and in MODE 4 in 36 hours. This changes the CTS by requiring the plant to be in MODE 3 in 12 hours and by extending the time to reduce reactor water temperature to < 21 20F (i.e., MODE 4) from 24 hours to 36 hours. The purpose of CTS 3.5.C.3 is to place the plant outside the Applicability of the Specification within a reasonable amount of time. This change is acceptable because the Completion Time is consistent with the safe operation under the specified Condition, considering the OPERABLE status of the redundant systems or features. This includes the capacity and capability of remaining systems or features, a reasonable time for repairs or replacement, and the low probability of a DBA occurring during the allowed Completion Time. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. This change is also acceptable because it requires the unit to be in an intermediate condition (MODE 3) sooner than is currently required (12 hours versus 24 hours). This portion of the change reduces the amount of time the unit would be allowed to continue to operate in MODES 1 and 2 once the condition is identified. The consequences of a pressurization event are significantly reduced when the reactor is shutdown and a controlled cooldown is already in progress. This change is designated as less restrictive because additional time is allowed to place the unit outside the LCO Applicability than is allowed in the CTS. Monticello Page 3 of 3 Attachment 1, Volume 11, Rev. 0, Page 203 of 431
Attachment 1, Volume 11, Rev. 0, Page 204 of 431 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 204 of 431
Attachment 1, Volume 11, Rev. 0, Page 205 of 431 Spray( CTS 3.6 CONTAINMENT SYSTEMS 3-5.C 3.62, Residual Heat Removal (RHR)ISuppressr Pool Spray 0 3.5.C.1 LCO 3.6. Two RHRI suppr n pool spray subsystems shall be OPERABLE. 0D 3.5.C.1 APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 3.5.C.2 A. One RHR A.1 Restore RHR su ssion 7 days } 0D E fE~spray subsystem spray subsystem to inoperable. OPERABLE status. Restore one RHR 8 hours 0 DOC L. B. Isuppressio-5 poolspray subsystem to OPERABLE
}
status. 3.5.C.3 C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time not met. AND C.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
.el SR 3.6 .1 pooilspray subsystem 31 days DOC M2 Vreswn nua power operate and aomatic valve in the o0 flow path that is not locked, sealed, or otherwise secured in position is in the correct position or can be aligned to the correct position.
BWR/4 STS 3.6.2.4-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 205 of 431
Attachment 1, Volume 11, Rev. 0, Page 206 of 431 RHR[ PoolSpray
~3.6. (3 c SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.6¶%.2 [Veri7y each RHR p mp develops Aflow rate In acc'dac wih thd ice 0 2 [4pt0] gpm througj the heat exchfinger while rating in the sj pression pool pray mode. tin Pro ram 0 o 92 days 4 [Verify each drywell spray header and 4.5.C.1 Inozzle Is unobstructed. BWR/4 STS 3.6.2.4 -2 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 206 of 431
Attachment 1, Volume I1, Rev. 0, Page 207 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.1.8, RESIDUAL HEAT REMOVAL (RHR) DRYWELL SPRAY
- 1. The Monticello safety analyses takes credit for the operation of the drywell spray function, not the suppression pool spray function. This mode of operation is known as the Residual Heat Removal Drywell Spray mode. Therefore, the name of this Specification, as well as the LCO, ACTIONS, and Surveillances, have been modified to reflect this assumption. In addition, since the new Specification supports the drywell, not the suppression pool, the Specification number has been changed to ITS 3.6.1.8 (The ISTS 3.6.1 Specifications are drywell related while the ISTS 3.6.2 Specifications are suppression pool related).
- 2. The Monticello design does not include any automatically actuated RHR drywell spray valves. The containment spray mode is manually actuated. Therefore, the reference to "automatic" in ISTS SR 3.6.2.4.1 (ITS SR 3.6.1.8.1) has been deleted.
- 3. The bracketed requirement (ISTS SR 3.6.2.4.2) has been deleted. The current licensing basis for Monticello does not require operation of suppression pool spray, as discussed in JFD 1 above.
- 4. A new Surveillance (ITS SR 3.6.1.8.2) has been added that verifies each drywell spray header and nozzle is unobstructed every 10 years. This Surveillance is required to ensure that, when a drywell spray subsystem is required, it will perform as designed. If a header and associated spray nozzles are obstructed, then the design function may not be met.
Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 207 of 431
Attachment 1, Volume 11, Rev. 0, Page 208 of 431 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 208 of 431
Attachment 1, Volume I1, Rev. 0, Page 209 of 431 RHR g n PoollSpray B 3.6% 0 B 3.6 CONTAINMENT SYSTEMS B 3.64 Residual Heat Removal (RHR) jpi6ooI Spray 0 E1 FDsyWell BASES BACKGROUND Following a Design Basis Accident (DBA), the RI-R[Supprosslton Pooll 0 Spray Syste ]removes hetfo thepsuppression chm r airspace. The 7suppresso Ipo ned toAsorb the sudde iriu fharom is ds~ig the primary system roa BA or a rapid depressu iatn of the reactor pressure vessel PVt) trouugh safe y/relief valves. Te/heat addition toC [I C the suppression ool resuits in increased steam in th siprsion chamber, which jncreases primnary/containment pressute. Steam blowdown from paDBA can alsco b Ipas the suppression pool and end ups 0 in the suppres chamber airsplce.' Some means rut be provided tot in remove heat f jmthe suppressiqo chamber so that the pressure an-dl temperature i primary containment remain withyin analyzed desig
~ide p Thi fucinispoie by two redundant RHRlsunpr On spray subsystems. The purpose of this LCO is to ensure that bowhi s poolI 0
subsystems are OPERABLE in applicable MODES.I rell Each of the two RHtp ronool spray subsystems contains two (D pumps and one heat exchanger, which are manually initiated and independently controlled. The two subsystems perform the su ssion most of spray function by circulating water from the suppreson 'pool throug the RHR heat exchangers and returninlrit t-otheisuporeso olspaasaad e he I spar ers on ly accommdate a smal porti o he oa RHFXpunp fow;ta remainder of the/fowatreturns to thesprsinoo through th supesion pool cooigrtn lie. Thus, ~t upeso pool cooling and ppression pool s ray functio the Suppression Pol Spray Sytr isiiitd H ervice water, A circulating through the tube side of the heat exchangers, exchanges hea ultimate rdT; with the supression ool water and discharges this heat to the e na heat sink. Either RHR suppreon Poolpray subsystem is sufficient to cam~e aispac/dung te pstulated DBA. Suppessin APPLICABLE Reference 1 contains the results of analyses used to predictri vl SAFETY drsr-el3 temperature following parge ans-sn all brea ANALYSES 0s cooTiaccidentR. The intent of the analyses is to demonstrate vossesof that th re ur reduction capacity of the RHR pray 00 elI salne breaks System is adequate to maintain the primary containment conditions within i (temperature design limits. The time history for primary containment Trur calculated to demonstrate that the maximum pre-sureLemains below the design limit. temperature The RHRI sn'oollSpiy System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). 0D BWR/4 STS B 3.6.2.4-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 209 of 431
Attachment 1, Volume II, Rev. 0, Page 210 of 431 B 3.6.1.8 Q, INSERT I condenses any steam that may exist in the drywell thereby lowering drywell pressure and temperature. The RHR Drywell Spray mode of operation is not credited in the DBA loss of coolant accident (LOCA), however it is credited for the evaluation of steam line breaks inside the drywell. For these events, the RHR Drywell Spray System will ensure that the drywell air temperature is within the peak drywell air temperature limit of 335°F specified for the drywell temperature envelope for equipment qualification and will also ensure that the drywell wall temperature is within the design limit of 281 OF. Q3 that may exist in the drywell INSERT 2 Insert Page B 3.6.2.4-1 Attachment 1, Volume 11, Rev. 0, Page 210 of 431
Attachment 1, Volume 11, Rev. 0, Page 211 of 431 RH Su Pool Spray
'the consequences' B 3.6.Qo (
of steam line l: breaks In the BASES .drywell[rwl LCO In the event of a DBA, a minimu f one RHRsuppressron pool spray subsystem is required to mitigatelteta yn~ass-Fmakadept- n G maintain the primary containment peak ow the design limits (Ref. 1). To ensure that these requirements are met, two RHR - temperature} su pren poo spray subsystems must be OPERABLE with power from two safety related independent power supplies. Therefore, in the event of an accident, at least one subsyste is OPERABLE assuming the worst case single active failure. An RHRjsuppron ool spray subsystem is (inaderandnryowles)spray OPERABLE when one of the pumps, the heat exchanger, and associated
' piping, valves, instrumentation, and controls are OPERABLE.
APPLICABILITY In MODES 1, 2, and 3, a DBA could cause pressurization of primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining RHRosupp eison pool spray subsystems OPERABLE is not required in MODE 4 or 5. 3 0D ACTIONS A.1 [P l With one RHRsoupprgsston po0 spray subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this Condition, the remaining OPERABLE RHR su sszon (i @ spray subsystem is adequate to perform the primary containment 1yass leakage mitigation function. However, the overall reliability is drywell reduced because a single failure in the OPERABLE subsystem could srv result in reduced 1primary cog ent bypass mitigation capability. The IJ 7 day Completion Time was chosen in light of the redundant RHR
=wsr ospray capabilities afforded by the OPERABLE (i) subsystem and the low probability of a DBA occurring during this period.
B.1[ rwl] With both RHRrsuppressin pMspray subsystems inoperable, at least one subsystem must be restored to OPERABLE status within 8 hours. In CgdFywlI this #ondition, there is a substantial loss of the primary ainmentl (7)( mitigation function. The 8 hour Completion Time is go based on this loss of function and is considered acceptable due to the low probability of a DBA and because alternative methods to remove heat from primary containment are available. BWR/4 STS B 3.6.2.4-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 211 of 431
Attachment 1, Volume 11, Rev. 0, Page 212 of 431 RHR ~uoo Sprayooll ra BASES ACTIONS (continued) C.1 and C.2 If the inoperable RHR~suppTessRn poollspray subsystem cannot be restored to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE 0R3 REQUIREMENTS and deii Verifying the correct alignment for manual power operatedF[(i l e}tici~valves in the RH Isuppress1on pool spray mode flow path (E provides assurance that the proper flow paths will exist for system operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position provided it can be aligned to the drywell spray accident osition within the ime assumed in the accident analysis. This is acceptable since the RH isuppressi ool cooling mode is manually 0 initiated. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. The Frequency of 31 days is justified because the valves are operated under procedural control, improper valve position would affect only a single subsystem, the probability of an event requiring initiation of the system is low, and the subsystem is a manually initiated system. This Frequency has been shown to be acceptable based on operating experience. SR.6'0 Verifying each R pump develops a flow rate 2 [4 0] gpm while operating in the ppression pool spray mode wit flow through the heat { INSERT33 exchanger ens es that pump performance has t degraded during the cycle. Flow is normal test of centrifugal pump erformance required by Section Xl of he ASME Code (Ref. 2). This te t confirms one point on BWR/4 STS B 3.6.2.4-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 212 of 431
Attachment 1, Volume 11, Rev. 0, Page 213 of 431 B 3.6.1.8
,0 INSERT 3 This Surveillance is performed every 10 years to verify that the drywell spray nozzles are not obstructed and that spray flow will be provided when required. The 10 year Frequency is adequate to detect degradation in performance due to the passive nozzle design and has been shown to be acceptable through operating experience.
Insert Page B 3.6.2.4-3 Attachment 1, Volume 11, Rev. 0, Page 213 of 431
Attachment 1, Volume II, Rev. 0, Page 214 of 431 RHRI fioPool Spray B6 0 BASES SURVEILLANCE REQUIREMENTS (continued)
-
. .. - of overall.. pertormance. Such . I the pump design curve and. .is indicative inservice inspectio s confirm component OPERABJLITY, trend I-performance, ancydetect incipient failures by indi ting abnormal performance. Tfe Frequency of this SR is [in acordance with the Inservice Testi A Proaram. but the Freauencv ust not exceed 92 davsl.
REFERENCES CDThSAR, Section .2.3. 00
- 12. ASME.-Biier and Pressure VesseltCode. Section XI. I 0D BWR/4 STS B 3.6.2.4-4 Rev. 3.0, 03/31/04 Attachment 1,Volume 11, Rev. 0, Page 214 of 431
Attachment 1, Volume II, Rev. 0, Page 215 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.1.8 BASES, RESIDUAL HEAT REMOVAL (RHR) DRYWELL SPRAY
- 1. Changes have been made to reflect changes made to the Specification.
- 2. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the plant specific nomenclature, number, reference, system description, analysis description, or licensing basis description.
- 3. Typographical/grammatical error corrected.
- 4. The brackets have been removed and the proper plant specific information/value has been provided.
Monticello Page 1 of I Attachment 1, Volume 11, Rev. 0, Page 215 of 431
Attachment 1, Volume 11, Rev. 0, Page 216 of 431 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 11, Rev. 0, Page 216 of 431
Attachment 1, Volume 11, Rev. 0, Page 217 of 431 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.6.1.8, RESIDUAL HEAT REMOVAL (RHR) DRYWELL SPRAY There are no specific NSHC discussions for this Specification. Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 217 of 431
Attachment 1, Volume 11, Rev. 0, Page 218 of 431 ATTACHMENT 9 ITS 3.6.2.1, Suppression Pool Average Temperature Attachment 1, Volume 11, Rev. 0, Page 218 of 431
Attachment 1, Volume 11, Rev. 0, Page 219 of 431 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 11, Rev. 0, Page 219 of 431
C C C 0 ITS 3.6.2.1 ITS ITS M S Applies to the primary and secondaty contl t Integrity. CD W 0 ObloctW: 0 To verify the Integrily of th ry and secondary containmentL CD Specificaltion -A A. Primary Containment 3.6.2.1 ) 1. Suppression Pool Volume and Temperature Applicability a. The suppressIon chamber water temperature CD 0 shall be checked once per day. 0 0 -9, LCO 3.6.2.1.a 5.2.1.1
)operalio
- b. Whenevertherelsindicattono trefiee te p
i adds heat the pool tern e o the suppresslon ture shall M -9, C-, CD o obsed als A.3v 1 5 minutes until the heat additIon I3 terminated. LCO 3.6.2.1.b 0 I . A visual inspection of the suppresslon chamber 0 interlor Including water lne regions and the See ITS 3.6.1.1 Required Action A.2 InterIor painted surfaces above the water One
- shall be made at each refueling outage.
LCO 3.6.2.1.c ACTION D - LCO 3.6.2.1.a - 3.7/4.7 156 01/28105 4 j Add proposed ACTIONAA Amendment No. W3,3 141 t - lAdd proposed ACTION C L3 Page 1 of 2
( C ITS 3.6.2.1 C ITS
- d. W'Vhenev' there is ind o o rellef valte
- 0) ACTION E pressure vessel shail be depressured opera n with a supp -slonpool tempera r s 200 psig oolmeradtur ex the af 860'F and the rimary coolant systK 03 CD suppressiron pool temperature exceed 120°E7 pesure >200 pg, an extended vis xamination of fe suppresslon dia rler shall 3(C The suppression pool water level shaUl be /be conducte ~efore resuming poer 0 2 -4.0 and - +3.0 Inches. With suppression nnooration.
0 pool water level not within limits, restore water level to within limits within the succeeding 0, See ITS 3.6.2.2 } 0 21 ACTIONS B, D. and E 1. 3 0
-U tD 0 :) -U (0 0
to to 0 0CD K) 0 IN 3.714.7 157 09123/02 Amendment No. 30, 62, 66,117 130 Page 2 of 2
Attachment 1, Volume 11, Rev. 0, Page 222 of 431 DISCUSSION OF CHANGES ITS 3.6.2.1, SUPPRESSION POOL AVERAGE TEMPERATURE ADMINISTRATIVE CHANGES A.1 In the conversion of the Monticello Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1433, Rev. 3, "Standard Technical Specifications General Electric Plants, BWR/4" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 CTS 3.7.A.1 .a requires water temperature to be < 900 F during the condition of normal operation. ITS 3.6.2.1 restates this condition as THERMAL POWER
> 1% RTP. This changes the CTS by restating thie term normal operation in a more specific form used in the ITS.
The purpose of the CTS 3.7.A.1.a term is to state when the temperature limit is applicable. CTS 1.0.0 definition of Power Operation, which is analogous to the term "normal operating," is defined as the reactor critical above 1% RTP. Therefore, this change is presentation preference only and is acceptable. This change is designated as an administrative change because it does not result in any technical changes to the CTS. A.3 Whenever there is indication of relief valve operation that adds heat to the suppression pool, CTS 4.7.A.1.b requires the suppression pool temperature be continuously monitored and also observed and logged every 5 minutes until the heat addition is terminated. Under similar conditions (as modified by DOC M.5), ITS SR 3.6.2.1.1 requires the suppression pool temperature to be verified (which is analogous to observed) to be within the applicable limit once per 5 minutes. This changes the CTS by deleting the requirements to continuously monitor and log the suppression pool temperature. The purpose of CTS 4.7.A.1.b is to ensure the suppression pool temperature is being closely monitored when there is a higher potential that the suppression pool temperature limit could be exceeded. The logging requirement duplicates the requirements of 10 CFR 50 Appendix B, Section XVII (Quality Assurance records to maintain records of activities affecting quality, including the results of tests (i.e., Technical Specification Surveillances). Compliance with 10 CFR 50 Appendix B is required by the Monticello Operating License, which is adequate to ensure appropriate data is taken and maintained. The details of the regulations within the Technical Specifications are repetitious and unnecessary. In addition, the design of the instrumentation is such that the suppression pool temperature is always continuously monitored. Changes to the plant design are controlled by 10 CFR 50.59. Thus, stating the design in the Technical Specifications is not necessary. Therefore, retaining the requirement to perform the associated Surveillance and eliminating the details from Technical Specifications that are found in 10 CFR 50 Appendix B or are part of the plant design is considered a presentation preference and is acceptable. As such, this change is considered an administrative change. Monticello Page 1 of 7 Attachment 1, Volume 11, Rev. 0, Page 222 of 431
Attachment 1, Volume 11, Rev. 0, Page 223 of 431 DISCUSSION OF CHANGES ITS 3.6.2.1, SUPPRESSION POOL AVERAGE TEMPERATURE MORE RESTRICTIVE CHANGES M.1 CTS 3.7.A.1 is applicable, in part, when irradiated fuel is in the reactor vessel and reactor water temperature is above 212 0F. ITS LCO 3.6.2.1 is applicable in MODES 1, 2, and 3. This changes the CTS by requiring the suppression pool water temperature to be within limits in MODE 2 when reactor water temperature is less than or equal to 212 0F. The purpose of CTS 3.7.A.1 is to ensure the suppression pool water temperature is within limits to mitigate the consequences of a design basis accident. The suppression pool water temperature is required to be within limits during MODES 1, 2, and 3 when there isconsiderable energy in the reactor core and a DBA could cause a significant heat up of the suppression pool. In MODES 1 0 and 3, the reactor coolant temperature will always be above 212 F. In MODE 2, the reactor coolant temperature may be less than or equal to 212 0 F when the reactor is subcritical but control rods are withdrawn. Therefore, it is necessary and acceptable to require the suppression pool water temperature to be within limits. This change is designated as more restrictive because the LCO will be applicable under more reactor operating conditions than in the CTS. M.2 CTS 3.7.A.1.c requires the reactor to be scrammed immediately if suppression pool water temperature is > 1100F. ITS 3.6.2.1 ACTION D requires a similar action, and also requires verifying the suppression pool temperature is < 120OF once per 30 minutes and to be in MODE 4 in 36 hours. This changes the CTS by requiring increased monitoring of the suppression pool average temperature and requiring the plant be placed in a MODE outside the applicability of the LCO when the suppression pool water temperature is > 11 0F. The purpose of CTS 3.7.A.1.c is to ensure the unit is placed in a condition that will not require the suppression pool function to mitigate the consequences of a 0 design basis accident, and to reduce the possibility of exceeding the 170 F end of bulk blowdown temperature resulting from a DBA LOCA. This change is designated as more restrictive because it adds new Required Actions not previously required in the CTS. M.3 CTS 3.7.A.1.d requires the reactor vessel to be depressurized to < 200 psig if the suppression pool temperature exceeds 1200F. However, this action is only required "during reactor isolation conditions" (i.e., main steam isolation valves (MSIVs) closed). ITS 3.6.2.1 ACTION E requires a similar action (as modified by DOC M.4), but it is required at all times when in MODE 3, not just during reactor isolation conditions. This changes the CTS by requiring a reactor vessel depressurization to < 200 psig at all times when in MODE 3 and suppression pool temperature exceeds 120'F. The purpose of CTS 3.7.A.1.d is to help ensure the maximum bulk suppression pool temperature limit is not exceeded, by quickly depressurizing the reactor pressure vessel, thus limiting the amount of heat that can be added to the suppression pool if a DBA were to occur. This change is acceptable because, if the suppression pool temperature exceeds 120'F, significant heat could still be added to the suppression pool regardless of MSIV position. Thus, the action to Monticello Page 2 of 7 Attachment 1, Volume 11, Rev. 0, Page 223 of 431
Attachment 1, Volume 11, Rev. 0, Page 224 of 431 DISCUSSION OF CHANGES ITS 3.6.2.1, SUPPRESSION POOL AVERAGE TEMPERATURE depressurize the reactor vessel is appropriate. This change is designated as more restrictive because it requires an action to be taken in more operating conditions than previously required in the CTS. M.4 CTS 3.7.A.1.d requires the reactor to be depressurized to < 200 psig "at normal cooldown rates" if suppression pool temperature exceeds 120 0F. ITS 3.6.2.1 ACTION E requires depressurizing the reactor to < 200 psig in 12 hours. This changes the CTS by specifying a Completion Time to depressurize the reactor vessel. The purpose of CTS 3.7.A.1.d is to help ensure the maximum bulk suppression pool temperature limit is not exceeded, by quickly depressurizing the reactor pressure vessel, thus limiting the amount of heat that can be added to the suppression pool if a DBA were to occur. This change is acceptable since it places an appropriate time to complete the depressurization. This change is designated as more restrictive because it reduces the time allowed by the CTS to complete a reactor depressurization to < 200 psig. M.5 CTS 4.7.A.1.b requires the suppression pool temperature to be checked every, 5 minutes whenever there is indication of "relief valve operation that adds heat to the suppression pool." ITS SR 3.6.2.1.1 requires similar suppression pool temperature verification every 5 minutes, but requires the verification to be performed anytime there istesting that adds heat to the suppression pool. This changes the CTS by requiring the every 5 minute suppression pool temperature verification anytime there is testing that adds heat to the suppression pool, not just when there is indication of "relief valve operation which adds heat to the suppression pool." The purpose of CTS 4.7.A.1.b is to ensure the suppression pool temperature is being closely monitored when there is a higher potential that the suppression pool temperature limit could be exceeded. When any testing is being performed that adds heat to the suppression pool, the suppression pool temperature should be more closely monitored since changes in suppression pool temperature are likely. Therefore, this change is acceptable. This change is designated as more restrictive because it requires a Surveillance Requirement to be performed more frequently in the ITS than is required in the CTS. M.6 CTS 3.7.A.1 .f, in part, requires the unit to be placed in a cold shutdown condition within 24 hours if the requirements of 3.7.A.1.a (i.e., suppression pool temperature during normal operation shall be <c90'F) or 3.7.A.1.b (i.e., suppression pool temperature not reduced to < 900F within 24 hours after suspension of testing that adds heat to the suppression pool) cannot be met. However, as described in DOC A.2, the condition of normal operation is when the reactor is critical and > 1% RTP. Thus, since CTS 3.7.A.1.a is only applicable when > 1% RTP, once reactor power is reduced to < 1%RTP, the CTS 3.7.A.1.a requirement is no longer applicable, and continuation to cold shutdown is not required. ITS 3.6.2.1 ACTION B requires a reduction in THERMAL POWER to
< 1% RTP in 12 hours. This changes the CTS by reducing the time allowed to be
< 1% RTP from 24 hours to 12 hours.
Monticello Page 3 of 7 Attachment 1, Volume 11, Rev. 0, Page 224 of 431
Attachment 1, Volume 11, Rev. 0, Page 225 of 431 DISCUSSION OF CHANGES ITS 3.6.2.1, SUPPRESSION POOL AVERAGE TEMPERATURE The purpose of CTS 3.7.A.1.f is to place the unit in a MODE in which the LCO is no longer applicable within a reasonable amount of time. This change is acceptable because the Completion Time is consistent with the safe operation under the specified Condition, considering the OPERABLE status of the redundant systems or features. This includes the capacity and capability of remaining systems or features, a reasonable time for repairs or replacement, and the low probability of a DBA occurring during the allowed Completion Time. The allowed Completion Time is reasonable, based on operating experience, to reach the required unit condition from full power conditions in an orderly manner and without challenging unit systems. This change is designated as more restrictive because less time is allowed to place the unit outside the LCO Applicability than is allowed in the CTS. M.7 After the completion of testing that adds heat to the suppression pool, 0 CTS 3.7.A.1.b allows 24 hours to restore suppression pool temperature to < 90 F if testing that adds heat~to the suppression pool is the cause for suppression pool temperature exceeding 90*F. ITS 3.6.2.1 ACTION A provides a similar 24 hour restoration time, however, an additional requirement (ITS 3.6.2.1 Required 0 Action A.1) to verify suppression pool temperature is < 110 F once per hour is also required. This changes the CTS by adding a requirement to verify suppression pool temperature is < 11 0F once per hour after the completion of testing that adds heat to the suppression pool, if the testing resulted in suppression pool temperature exceeding 90'F. The purpose of the new Required Action is to periodically monitor suppression pool temperature to ensure it does not exceed 1100 F without being quickly noticed by the operation staff. This helps ensure that proper actions are taken if the suppression pool temperature does exceed 110F (i.e., the reactor is scrammed). Therefore, this new periodic monitoring requirement is acceptable. This change is designated as more restrictive because a new requirement to periodically monitor suppression pool temperature under certain conditions is added to the CTS. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L.1 (Category 2 - Relaxation of Applicability) CTS 3.7.A.1 is applicable, in part, when irradiated fuel is in the reactor vessel and work is being done which has the potential to drain the vessel. As a consequence of this Applicability, CTS 3.7.A.1.f requires suspending all activities with the potential for draining the reactor vessel if the requirements of CTS 3.7.A.1 cannot be met. ITS 3.6.2.1 is Monticello Page 4 of 7 Attachment 1, Volume 11, Rev. 0, Page 225 of 431
Attachment 1, Volume 11, Rev. 0, Page 226 of 431 DISCUSSION OF CHANGES ITS 3.6.2.1, SUPPRESSION POOL AVERAGE TEMPERATURE applicable only in MODES 1, 2, and 3. This changes the CTS by deleting the requirement for the suppression pool water temperature to be within limits when irradiated fuel is in the reactor vessel and work is being done that has the potential to drain the reactor vessel and the requirement to suspend those operations when the LCO is not met. The purpose of CTS 3.7.A.1 is to ensure the suppression pool is capable of absorbing the decay heat and sensible energy released during a reactor blowdown from safety/relief valve discharges or from a Design Basis Accident (LOCA). This change is acceptable because the ITS 3.6.2.1 requirements continue to ensure that the suppression pool temperature is maintained in the MODES and other specified conditions assumed in the safety analyses and licensing basis. The ITS continues to require the suppression pool temperature limits to be met in MODES 1, 2, and 3. Thus, deleting the Applicability requirement for suppression pool temperature during operations with a potential for draining the reactor vessel does not affect the capability of the suppression pool to perform its safety function since the LCO still requires the limits to be met when a DBA could cause significant heatup of the suppression pool. This change is only allowing the temperature limits to not be met in MODES 4 and 5. The only remaining conditions in which irradiated fuel is in the vessel, and in these MODES significant heatup of the suppression pool following a DBA LOCA is not possible. This change is designated as less restrictive because the LCO requirements are applicable in fewer operating conditions than in the CTS. L.2 (Category 4 - Relaxation of Required Action) CTS 3.7.A.1 .f requires the reactor be placed in a cold shutdown condition within 24 hours if the suppression pool temperature is > 900F and testing that adds heat to the suppression pool is not the cause of exceeding 90'F. ITS 3.6.2.1 ACTION A provides 24 hours to restore the suppression pool temperature to < 900F, provided the suppression pool temperature is verified to be < 11 0F once per hour, prior to requiring the unit to exit the Applicability of the LCO. This changes the0CTS by allowing 24 hours to restore suppression pool temperature to < 90 F when testing that added heat was not the cause of exceeding 90'F, provided suppression pool temperature is verified < 11 0IF once per hour. The purpose of ITS 3.6.2.1 ACTION A is to provide a short period of time to restore the inoperable suppression pool average temperature prior to requiring exiting the Applicability of the LCO. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing time to repair inoperable features. The Required Actions are consistent with safe operation under the specified Condition, considering the OPERABILITY status of required features, the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of a DBA occurring during the repair period. ITS 3.6.2.1 ACTION A allows the suppression pool average temperature to be not within limits for a time period similar to that allowed when testing that adds heat to the suppression pool is the cause of exceeding 900F. This inoperability is acceptable because the primary containment analysis shows that the suppression pool can still mitigate the consequences of a design basis event as long as the suppression pool Monticello Page 5 of 7 Attachment 1, Volume 11, Rev. 0, Page 226 of 431
Attachment 1, Volume 11, Rev. 0, Page 227 of 431 DISCUSSION OF CHANGES ITS 3.6.2.1, SUPPRESSION POOL AVERAGE TEMPERATURE temperature remains below 11 0F. In addition, a new 1 hour periodic verification is required to ensure the suppression pool temperature is < 11 0F. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. L.3 (Category 4 - Relaxation of Required Action) CTS 3.7.A.1.f requires the reactor be placed in a cold shutdown condition within 24 hours if the suppression pool temperature exceeds 100OF during test operation which adds heat to the suppression pool. ITS 3.6.2.1 ACTION C requires suspending all testing that adds heat to the suppression pool immediately. Once testing is suspended, ITS 3.6.2.1 ACTION A would allow 24 hours to restore temperature to < 900F, consistent with the time allowed in CTS 3.7.A.1.b. This changes the CTS by providing an allowance to suspend testing that adds heat to the suppression pool if suppression pool temperature exceeds 1000F, in lieu of requiring a unit shutdown. The purpose of CTS 3.7.A.1.f is to provide actions that result in exiting the Applicability of the LCO if suppression pool temperature exceeds 100 0F during testing operations that add heat to the suppression pool. ITS 3.6.2.1 ACTION C requirement to immediately suspend all testing will preserve the heat absorption capability of the suppression pool, and with testing suspended require restoration of the suppression pool temperature to < 900F within 24 hours. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing time to repair inoperable features. The Required Actions are consistent with safe operation under the specified Condition, considering the OPERABILITY status of required features, the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of a DBA occurring during the restoration period. Furthermore, if testing were suspended just prior to exceeding 100F, CTS 3.7.A.1.b would allow up to 24 hours to restore suppression pool temperature to < 900F. The allowance to continue to operate with a suppression pool temperature > 100IF for up to 24 hours is also acceptable because the primary containment analysis shows that the suppression pool can still mitigate the consequences of a design basis event as long as the suppression pool temperature remains below 11 0F. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. L.4 (Category 3 - Relaxation of Completion Time) CTS 3.7.A.1.f requires the unit to be placed in a cold shutdown condition within 24 hours if the requirements of CTS 3.7.A.1 cannot be met. ITS 3.6.2.1 Required Actions D.3 and E.2 require the reactor be in MODE 4 in 36 hours. This changes the CTS by extending the time allowed to be in cold shutdown (i.e., MODE 4) from 24 hours to 36 hours. The purpose of CTS 3.7.A.1.f is to place the unit In a MODE in which the LCO is no longer applicable within a reasonable amount of time. This change is acceptable because the Completion Time is consistent with the safe operation under the specified Condition, considering the OPERABLE status of the redundant systems or features. This includes the capacity and capability of Monticello Page 6 of 7 Attachment 1, Volume 11, Rev. 0, Page 227 of 431
Attachment 1, Volume 11, Rev. 0, Page 228 of 431 DISCUSSION OF CHANGES ITS 3.6.2.1, SUPPRESSION POOL AVERAGE TEMPERATURE remaining systems or features, a reasonable time for repairs or replacement, and the low probability of a DBA occurring during the allowed Completion Time. The allowed Completion Time is reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. This change is designated as less restrictive because additional time is allowed to place the unit outside the LCO Applicability than is allowed in the CTS. L.5 (Category 5- Deletion of Surveillance Requirement) CTS 4.7.A.1.d requires an extended visual examination of the suppression chamber before resuming power operation whenever there is indication of relief valve operation with a suppression pool temperature of > 1600 F and the primary coolant system pressure > 200 psig. The ITS does not include this Surveillance Requirement. This changes the CTS by deleting the Surveillance Requirement to perform a suppression chamber inspection based on special temperature and pressure conditions. The purpose of CTS 4.7.A.1.d is to verify the structural integrity of both the internal and external torus areas after being subjected to an elevated temperature and pressure. This Surveillance is being deleted in accordance with NEDO-30832, "Elimination of Limit on BWR Suppression Pool Temperature for SRV Discharge with Quenchers," dated December 1984. The basis for deleting this Surveillance is that testing has demonstrated that there are no undue loads on the suppression pool or its components at elevated temperatures and pressures when SRVs discharge through "quenchers" (spargers). At Monticello, each relief valve discharge line terminates in a T-quencher (sparger). Therefore, deleting this Surveillance is acceptable. This change is designated as less restrictive because Surveillances which are required in the CTS will not be required in the ITS. Monticello Page 7 of 7 Attachment 1, Volume 11, Rev. 0, Page 228 of 431
Attachment 1, Volume 11, Rev. 0, Page 229 of 431 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 229 of 431
Attachment 1, Volume 11, Rev. 0, Page 230 of 431 Suppression Pool Average Temperature 3.6.2.1 CTS 3.6 CONTAINMENT SYSTEMS 3.7.A.i 3.6.2.1 Suppression Pool Average Temperature LCO 3.6.2.1 Suppression pool average temperature shall be: 3.7Ai.a a. cha Fe[when ar 40 divisions of full scale a (I 71J~ithhane 0 THERMAL POWER > 1% RTFJ and no testing that adds heat to the suppression pool is being performed, b b: sI o Fl[whenay Ef IRMch divisions
..
o F1on Range 7 fwith THERMAL POWER > 1% RTPQ and (0 testing that adds heat to the suppression pool is being performed, and VI11 CgrF [when all I RM Qhannelsnr25/4011 3.7A1.c c. diiisiou scale on Range 71flwith THERMAL POWER 0D
- 1% RTF1.
3.7A.1 APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 3.7A.1.b, A. Suppression pool A.1 Verify suppression pool Once per hour DOC L.2 average temperature average temperature I°F but5 [I 011 CgF.
- 11C°OF. 0 AND AND
[Any OPERABLEIRM A.2 Restore suppression pool 24 hours channel> [254 divis full scale on average temperature to WOF. 0 POWER > 1%RTF¶ AND Not performing testing that adds heat to the suppression pool. BWR/4 STS 3.6.2.1-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 230 of 431
Attachment 1, Volume 11, Rev. 0, Page 231 of 431 Suppression Pool Average Temperature CTS 3.6.2.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME 3.7.A.1.f B. Required Action and B.1 Reduce HERMAL 12 hours associated Completion POWERI[until all Time of Condition A not met. OPERABLEI
* [2514 cannels vsions of full 0
eagon Rangee7to < 1% RTP.
-_4 DOC L.3 C. Suppression pool C.1 Suspend all testing that Immediately average temperature adds heat to the
> AND suppression pool. .0 AND 29&ge 74THERMAL 0
POWER > 1% RTFg. AND Performing testing that adds heat to the suppression pool. 4_ 3.7A1.c, D. Suppression pool D.1 Place the reactor mode Immediately 3.7A1.f average temperature switch in the shutdown
>011 C5°Fjbut 2M~0]FI. position. 00 AND
__ De4nineE D.2 lVisuppression pool average temperature Once per 30 minutes 0 0D AND D.3 Be in MODE 4. 36 hours
.
BWRI4 STS 3.6.2.1-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 231 of 431
Attachment 1, Volume 11, Rev. 0, Page 232 of 431 Suppression Pool Average Temperature CTS 3.6.2.1 ACTIONS (continued) CONDITION REQUIRED ACTION j COMPLETION TIME E. Suppression pool EA Depressurize the reactor 12 hours 3.7A.1.d, 3.7.A.1.f average temperature vessel to <1200 psig. 0
> T1 2Ca°F.
AND E.2 e in MOD / [36 hours] 0 SURVEILLANCE REQUIREMENTS Y SURVEILLANCE FREQUENCY
*1*
4.7.A.1.a, SR 3.6.2.1.1 Verify suppression pool average temperature is 24 hours 4.7.A.I.b within the applicable limits. AND 5 minutes when performing testing that adds heat to the suppression pool
.1.
BWR/4 STS 3.6.2.1-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 232 of 431
Attachment 1, Volume 11, Rev. 0, Page 233 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.2.1, SUPPRESSION POOL AVERAGE TEMPERATURE
- 1. The brackets have been removed and the proper plant specific information/value has been provided.
- 2. ISTS 3.6.2.1, ACTIONS D and E currently allow a resetting of the shutdown requirement (be in MODE 4) Completion Time when the suppression pool average temperature increases above 120OF or decreases to 120 0F. If temperature is initially
> 110F but < 1200F, ISTS 3.6.2.1 Condition D is entered and the 36 hour Completion Time clock of ISTS 3.6.2.1 Required Action D.3 (be in MODE 4) starts.
However, if temperature increases above 120 0F, ISTS 3.6.2.1 Condition D is not longer applicable, and ISTS 3.6.2.1 Condition E is entered. Upon entry into ISTS 3.6.2.1 Condition E, Required Action E.2 allows a new 36 hour Completion Time clock to be in MODE 4. Furthermore, if the plant is already in ISTS 3.6.2.1 Condition E (i.e., temperature is > 1200F), and temperature decreases to 120OF or less, then ISTS 3.6.2.1 Condition E is no longer applicable, and ISTS 3.6.2.1 Condition D is entered, starting a new 36 hour Completion Time clock to be in MODE 4. Therefore, to ensure the plant is shut down within 36 hours after entering a requirement to be in MODE 4, ISTS 3.6.2.1 Condition D has been modified to delete the upper temperature limit, ISTS 3.6.2.1 Required Action D.2 has been modified to require determining suppression pool temperature (in lieu of a specific limit), and ISTS 3.6.2.1 Required Action E.2 has been deleted. Thus, when temperature exceeds 110F, ITS 3.6.2.1 Condition D is entered and a 36 hour Completion Time clock is started to be in MODE 4. If temperature increases above 1200F, ITS 3.6.2.1 Condition E is also entered and Required Action E.1 taken. Furthermore, the Completion Time clock of ITS 3.6.2.1 Required Action D.3 continues to run, and the plant must be in at least MODE 4 within 36 hours after temperature exceeds 1100F. This change is also consistent with proposed TSTF-458. Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 233 of 431
Attachment 1, Volume 11, Rev. 0, Page 234 of 431 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1,Volume 11, Rev. 0, Page 234 of 431
Attachment 1,Volume 11, Rev. 0, Page 235 of 431 Suppression Pool Average Temperature B 3.6.2.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.2.1 Suppression Pool Average Temperature BASES BACKGROUND The suppression chamber is a toroidal shaped, steel pressure vessel containing a volume of water called the suppression pool. The suppression pool is designed to absorb the decay heat and sensible energy released during a reactor blowdown from safety/relief valve discharges or from Design Basis Accidents (DBAs). The suppression pool must quench all the steam released through the downcomer lines during a loss of coolant accident (LOCA). This is the essential mitigative feature of a pressure suppression containment that ensures that the peak containment pressure is maintained below the maximum allowable pressure for DBAs I[62]g psig). The suppression pool must also condense steam from steam exhaust lines in the turbine driven systems (i.e., the 0D High Pressure Coolant Injection System and Reactor Core Isolation Cooling System). Suppression pool average temperature (along with LCO 3.6.2.2, "Suppression Pool Water Level") is a key indication of the capacity of the suppression pool to fulfill these requirements. The technical concerns that lead to the development of suppression pool average temperature limits are as follows:
- a. Complete steam condensation -ithe original limit for the end of a LOCA blowdown was 170'F, based on the Bodega Bay and (i)
Humboldt Bay Testsr
- b. Primary containment peak pressure and temperature -design -
pressure isj psig'and design temperature is 1 F (Ref. 1)3 i) 0
- c. Condensation oscillation loads -lmaximum allowable initial temperature isiM11q°Fi and 0 0
- d. Chugging loads -theseonly occur at <Ml35rF; therefore, there is no initial temperature limit because of chuggings. 0 APPLICABLE The postulated DBA against which the primary containment performance SAFETY is evaluated is the entire spectrum of postulated pipe breaks within the ANALYSES primary containment. Inputs to the safety analyses include initial suppression pool water volume and suppression pool temperature [0 14J (Reference Jfor LOCAs and Reference r the pool temperature analyses required by Reference An initia pool temperature of~m°F
\isassumed for the Referen~eiand Referenc analyses. Reactor BWR/4 STS B 3.6.2.1-1 Rev. 3.0, 03/31/04 Attachment 1,Volume 11, Rev. 0, Page 235 of 431
Attachment 1, Volume 11, Rev. O, Page 236 of 431 Suppression Pool Average Temperature B 3.6.2.1 BASES APPLICABLE SAFETY ANALYSES (continued) shutdown at a pool temperature of llQj0 F and vessel depressurization at a ool temperature of [1 2Oj0 F are assumed for the Reference Mnalyses. 0F, at which testing is terminated, is not used in th The limit o safety analyses because DBAs are assumed to not initiate during unit testing. 3 Suppression pool average temperature satisfies Criteria 2 and 3 of 10 CFR 50.36(c)(2)(ii). LCO A limitation on the suppression pool average temperature is required to provide assurance that the containment conditions assumed for the safety analyses are met. This limitation subsequently ensures that peak primary containment pressures and temperatures do not exceed maximum allowable values during a postulated DBA or any transient resulting in heatup of the suppression pool. The LCO requirements are:
- a. Average temperature [F w en ERABLE inter range r can i /40 divisions o cal n 1~nOe7 Rwith THERMAL POWER > 1% RATED THERMAL X POWER (RTPU and no testing that adds heat to the suppression pool is being performed. This requirement ensures that licensing bases initial conditions are met.
- b. Average temperature s SlFffwhe-n any-wERABLE IRM channelo ji_]~n ttl~c~~-~g fw-ith THERMAL )
POWER > 1% RTPW and testing that adds heat to the suppression pool is being performed. This required value ensures that the unit has testing flexibility, and was selected to provide margin below the Ti1I0'F limit at which reactor shutdown is required. When testing ends, temperature must be restored to °1OF within 24 hours 0 accordigo Required Action A.2. Therefore, the time period that the 0 F is short enough not to cause a significant temperature is > increase in unit risk.
- c. Average temperature s(11Q MF w en a
*Iare U--WUfdym~s of full scalgii~qe 71 Wwith THERMAL POWER 5 1% RTPM. This requirement ensures that the unit will be shut down at > 11q 0F. The pool is designed to absorb decay heat and sensible heat but could be heated beyond design limits by the steam generated if the reactor is not shut down.
BWR/4 STS B 3.6.2.1-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 236 of 431
Attachment 1,Volume 11, Rev. 0, Page 237 of 431 Suppression Pool Average Temperature B 3.6.2.1 BASES LCO (continued) [Note that [2 divisions of full sca ilRM Range 7 is a . enient measul fwhen the reactor i oducing power essen eequivalent tol ffTP At r eve l1%Mis RTFP, heat input is approximately equal to normal system heat losses. APPLICABILITY In MODES 1, 2, and 3, a DBA could cause significant heatup of the suppression pool. In MODES 4 and 5, the probability and consequences .i of these events are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining suppression pool average temperature within limits is not required in MODE 4 or 5. 00 ACTIONS A.1 and A.2 With the suppression pool average temperature above the specified limit when not performing testing that adds heat to the suppression pool and when above the specified powe'rindftior1, the initial conditions exceed the conditions assumed for the Reference. 3, and MtnalysesC I i2 However, primary containment cooling capability still exists, and the primary containment pressure suppression function will occur at temperatures well above those assumed for safety analyses. Therefore, continued operation is allowed for a limited time. The 24 hour Completion Time is adequate to allow the suppression pool average temperature to be restored below the limit. Additionally, when suppression pool temperature is >E1oF, increased monitoring of the suppression pool temperature is required to ensure that it remains s11qlF. The once per hour Completion Time is adequate based on past experience, which has 0 shown that pool temperature increases relatively slowly except when testing that adds heat to the suppression pool is being performed. Furthermore, the once per hour Completion Time is considered adequate in view of other indications in the control room, including alarms, to alert the operator to an abnormal suppression pool average temperature condition. B.1 If the suppression pool average temperature cannot be restored to within limits within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the power must be reduced tol [< [25/401 divisioons sn Range67Mfl ga PtBL- slS 1% RTPawithin 12 hours. The 12 hour 0 Completion Time is reasonable, based on operating experience, to reduce power from full power conditions in an orderly manner and without challenging plant systems. BWR/4 STS B 3.6.2.1-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 237 of 431
Attachment 1, Volume 11, Rev. 0, Page 238 of 431 Suppression Pool Average Temperature B 3.6.2.1 BASES ACTIONS (continued) C.1 2
-Suppression pool average temperature is aloeo be> ° F Hi F any OP ABLE IRM-ha nel is > I`251Mdivisions of full c on I (i e 7 @Dith THERMAL POWER > 1% RTFU, and when testing that adds heat to the suppression pool is being performed. However, if 100 temperature is > all testing must be immediately suspended to O1°F, preserve the heat absorption capability of the suppression pool. With the testing suspended, Condition A is entered and the Required Actions and associated Completion Times are applicable.
D.1 and D.2 Suppression pool average temperature > 111COF requires that the reactor (i MODE be shut down immediately. This is accomplished by placing the reactor hf36~ mode switch in the shutdown position. Further cooldown t 4, is @ (i) I 36 required at normal cooldown rates (provided pool temperature remains sl12cJ2°F). Additionally, when suppression pool temperature is > 11OPF, (i) increased monitoring of pool temperature is required Ro ensurd- it A} remair-120]°Fl The once per 30 minute Completion Time is adequate, based on operating experience. Given the high suppression IRquied__ A.pool average temperature in thisoondition, the monitoring Frequency is increased to twice that = Furthermore, the 30 minute Completion Time is considered adequate in view of other indications available in the control room, including alarms, to alert the operator to an abnormal suppression pool average temperature condition. NER 1 G If suppression pool average temperature cannot be maintained at 5 [ q t12 pant epFt st be brought to a MO iwhich the LCO doesl Inot apply. To aelieve this statusFtFe reacor pressure must be reduced 1200fpsig within 12 hours and the lant rought to at least 0r) [MUDE 4wh 36 hours The allowed Completion Timem RLreasonable, based on operating experience, to reach the required plant con itionas ' from full power conditions in an orderly manner and without challenging plant systems. BWR/4 STS B 3.6.2.1-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 238 of 431
Attachment 1, Volume 11, Rev. 0, Page 239 of 431 B 3.6.2.1 (D) INSERT I Additionally, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 4 within 36 hours. The allowed Completion Time is reasonable, based on operating experience, to reach the required plant condition from full power conditions in an orderly manner and without challenging plant systems. Insert Page B 3.6.2.1-4 Attachment 1, Volume 11, Rev. 0, Page 239 of 431
Attachment 1, Volume 11, Rev. 0, Page 240 of 431 Suppression Pool Average Temperature B 3.6.2.1 BASES . S ACTIONS (continued) Continued addition of heat to the suppression pool with suppression pool temperature > V12 OF could result in exceeding the design basis maximum allowable values for primary containment temperature or 0 pressure. Furthermore, if a blowdown were to occur when the' temperature was > T12CgoF, the maximum allowable bulk and local temperatures could be exceeded very quickly. 0D SURVEILLANCE SR 3.6.2.1.1 REQUIREMENTS The suppression pool average temperature is regularly monitored to ensure that the required limits are satisfied. The average temperature is determined by taking an arithmetic average of OPERABLE suppression pool water temperature channels. The 24 hour Frequency has been shown, based on operating experience, to be acceptable. When heat is being added to the suppression pool by testing, however, it is necessary to monitor suppression pool temperature more frequently. The 5 minute Frequency during testing is justified by the rates at which tests will heat up the suppression pool, has been shown to be acceptable based on operating experience, and provides assurance that allowable pool temperatures are not exceeded. The Frequencies are further justified in view of other indications available in the control room, including alarms, to alert the operator to an abnormal suppression pool average temperature condition. J 0 REFERENCES SAR, Section[F _NEDC-23487-P.
- p. USAR, Section 5.2.1.1.
'Monticello Nudear Generating Plant 00 IFSAR, ion 15.1 . Suppresslon Pool Temperature Response.' December 1981 0
- 4. NUREG-0783. rkIContainmentProgram,'April1979. 0 E [Mark I Con knt Program.o (00 BWR/4 STS B 3.6.2.1-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 240 of 431
Attachment 1, Volume 11, Rev. 0, Page 241 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.2.1 BASES, SUPPRESSION POOL AVERAGE TEMPERATURE
- 1. The brackets have been removed and the proper plant specific information/value has been provided.
- 2. Changes have been made to reflect changes made to the Specification.
- 3. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the plant specific nomenclature, number, reference, system description, analysis description, or licensing basis description.
- 4. Typographical error corrected.
- 5. Editorial change made for enhanced clarity.
- 6. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, NEI 01-03, Section 5.1.3.
Monticello Page 1 of I Attachment 1,Volume 11, Rev. 0, Page 241 of 431
Attachment 1, Volume 11, Rev. 0, Page 242 of 431 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 11, Rev. 0, Page 242 of 431
Attachment 1, Volume 11, Rev. 0, Page 243 of 431 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.6.2.1, SUPPRESSION POOL AVERAGE TEMPERATURE There are no specific NSHC discussions for this Specification. Monticello Page 1 of I Attachment 1, Volume 11, Rev. 0, Page 243 of 431
, Volume 11, Rev. 0, Page 244 of 431 ATTACHMENT 10 ITS 3.6.2.2, Suppression Pool Water Level , Volume 11, Rev. 0, Page 244 of 431
Attachment 1, Volume 11, Rev. 0 Page 245 of 431 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1,Volume 11, Rev. 0, Page 245 of 431
C C ITS 3.6.2.2 C ITS 1 3.0 LUMITING CONDIONS FOR OPERATION 4.0 SURVEILLANCE REOUIREMENTS 3.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS IN CD Aoolicabirtd /roFbilfty: Applies to thi operating st o the primary and secondary to the primary and secondauy containm gAppies egrity. CD containment systems CD Oblecthre //Oblective: 0 To he integrity of the primary and a ary To verify the Integrity of the and secondary c ainment systems. crontainment. 0 Specification: ni See ITS 3.6.2.11 CD A. Primary ContainmenL A. Primary Containment 0 3.6.2.2 7,
- 1. SuppressIon Pool Volume andt 1. SuppressIon Pool Volume and em eratur Applicability When Irradiated fuel Is in the reactor vessel and l a. The suppressIon chamber water temperature CD CD either or water temrture ITS 3.5.2 > shall be checked once per day.
N) rwor is a done ch I diTS 3tenti61 LCO 3.6.22 r~~~~ ri hevsele th tllowing-re4ireens he
-{ See ITS 3.62.1 }
ITS 3.5.21 b. Whenever there Is indication of relief vate 135E2: I
-Ebemetlexcpt s pematd DVSpeSee r I operation which adds heat to the suppression to pool, the pool temperature shea! be continually 0)
- a. Water temperature during normal operating 03 shallbe s90F. monitored and also observed end logged every I CD6 5 minutes until the heat addition Is terminated. CY)
- b. Water temperature during test operation which adds heat to the suppression pool thall be C. A visual inspection of the suppression chamber See iTS 3.6.1.1 } 0 s I00 F and shall not be >901F for more than Interior Including water line regions and the -P.
24 hours. Interior painted surfaces above the water line Ca)
- a. If the suppression chamber water temperature shall be made at each refueling outage.
Is >110F. the reactor shall be scrammed Immediately. Power operation shall not be resumed until the pool temperature Is :5 9091F { See ITS 3.6.2.11 3.7/4.7 1se 01/28105 Amendment No. 63.23,141 Page 1 of 2
C C C ITS ITS 3.0 LtMITING CONDmONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS d During reactor Isolation conditions the reactor Whenever there is Indication of relief valve 01 CD 3 LCO 3.6.2.2 pressure vessel shall be depressurized to
< 200 paig at normal cooldovn rates if the suppression pool temperature exceed 120-F.
- e. IiiTe suppression pool water level shalt be
-t_-4.0 and s +3.0 Inches. With suppression
_ I See ITS 3.6.2.' 1 operation with a suppression pool temperature of 2t 160F and the primary coolant system pressure > 200 psig. an extended visual examination of the suppression chamber shall be conducted before resuming power operation. See ITS 3.6.2.1 } a) 0 3 (D 0 poo water level not within limits, restore water ACTION A level to within limits within the succeeding 2 hours. SR 3.6.2.2.1 e. The suppression pool water level shall be I - [MOOS 3 In checked once per day. 0 9.) I. It the requirements of 3.7.A.1 cannot be met. - C ACTION B the reactor shall be laced InSa Cold Shutdow tD
- condion within E1ours, n suspen al 3
0
;U aiities with tor ratreslL draining the
-t See ITS 3.5.2 } -
;t1 X
tD 0
-o -4 A) t 0t tD K) to
-4
.9h 0 CA) rs
-3 CA) 3.714.7 157 09/23102 Amendment No. 30f2 f5i6 t7, 130 Page 2 of 2
Attachment 1, Volume 11, Rev. 0, Page 248 of 431 DISCUSSION OF CHANGES ITS 3.6.2.2, SUPPRESSION POOL WATER LEVEL ADMINISTRATIVE CHANGES A.1 In the conversion of the Monticello Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1433, Rev. 3, "Standard Technical Specifications General Electric Plants, BWR/4" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. MORE RESTRICTIVE CHANGES M.1 CTS 3.7.A.1 is applicable, in part, when irradiated fuel is in the reactor vessel and reactor water temperature is above 212 0F. ITS LCO 3.6.2.2 is applicable in MODES 1, 2, and 3. This changes the CTS by requiring the suppression pool water level to be within limits in MODE 2, when reactor water temperature is less than or equal to 212 0F. The purpose of CTS 3.7.A.1 is to ensure the suppression pool water level is within limits to mitigate the consequences of a design basis accident. The suppression pool water level is required to be within limits during MODES 1, 2, and 3 when there is considerable energy in the reactor core and a DBA could cause a significant heat up of the suppression pool. In MODES 1 and 3, the reactor coolant temperature will always be above 212 0F. In MODE 2, the reactor coolant temperature may be less than or equal to 212 0F when the reactor is subcritical but control rods are withdrawn. Therefore, it is necessary and acceptable to require the suppression pool water level to be within limits. This change is designated as more restrictive because the LCO will be applicable under more reactor operating conditions than in the CTS. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L.1 (Category 3 - Relaxation of Completion Time) CTS 3.7.A.1.f requires the unit to be placed in the cold shutdown condition within 24 hours if the suppression pool water level requirements of CTS 3.7.A.1.e are not met. ITS 3.6.2.2 ACTION B requires the unit to be in MODE 3 in 12 hours and in MODE 4 in 36 hours. This changes the CTS by requiring the unit to be in MODE 3 in 12 hours and by Monticello Page 1 of 2 Attachment 1, Volume 11, Rev. 0, Page 248 of 431
Attachment 1, Volume 11, Rev. 0, Page 249 of 431 DISCUSSION OF CHANGES ITS 3.6.2.2, SUPPRESSION POOL WATER LEVEL extending the time allowed to be in cold shutdown (i.e., MODE 4) from 24 hours to 36 hours. The purpose of CTS 3.7.A.1.f is to place the unit outside the Applicability of the Specification within a reasonable amount of time. This change is acceptable because the Completion Time is consistent with the safe operation under the specified Condition, considering the OPERABLE status of the redundant systems or features. This includes the capacity and capability of remaining systems or features, a reasonable time for repairs or replacement, and the low probability of a DBA occurring during the allowed Completion Time. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. This change is also acceptable because it requires the unit to be in an intermediate condition (MODE 3) sooner than is currently required (12 hours versus 24 hours). This portion of the change reduces the amount of time the unit would be allowed to continue to operate in MODES I and 2 once the condition is identified. The consequences of a pressurization event are significantly reduced when the reactor is shutdown and a controlled cooldown is already in progress. This change is designated as less restrictive because additional time is allowed to place the unit outside the LCO Applicability than is allowed in the CTS. Monticello Page 2 of 2 Attachment 1, Volume 11, Rev. 0, Page 249 of 431
Attachment 1, Volume 11, Rev. 0, Page 250 of 431 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 250 of 431
Attachment 1, Volume 11, Rev. 0, Page 251 of 431
- Suppression Pool Water Level 3.6.2.2 K2J CTS 3.6 CONTAINMENT SYSTEMS 3.7A1 3.6.2.2 Suppression Pool Water Level 3.7.Al.e LCO 3.6.2.2 Suppression pool water level shall be 1 inches 2 inches] and < 0O 3.7.A.1 APPLICABILITY: MODES 1, 2, and 3.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 3.7Ai.e A. Suppression pool water A.1 Restore suppression pool 2 hours level not within limits. water level to within limits. 3.7.A.1 .f B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met. AND B.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.7A1.e SR 3.6.2.2.1 Verify suppression pool water level is within limits. 24 hours BWR/4 STS 3.6.2.2-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev.O, Page 251 of 431
Attachment 1, Volume 11, Rev. 0, Page 252 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.2.2, SUPPRESSION POOL WATER LEVEL
- 1. The brackets have been removed and the proper plant specific information/value has been provided.
Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 252 of 431
Attachment 1, Volume 11, Rev. 0, Page 253 of 431 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 253 of 431
Attachment 1, Volume 11, Rev. 0, Page 254 of 431. Suppression Pool Water Level B 3.6.2.2 B 3.6 CONTAINMENT SYSTEMS B 3.6.2.2 Suppression Pool Water Level BASES BACKGROUND The suppression chamber is a toroidal shaped, steel pressure vessel containing a volume of water called the suppression pool. The suppression pool is designed to absorb the energy associated with decay heat and sensible heat released during a reactor blowdown from safety/relief valve (S/RV) discharges or from a Design Basis Accident (DBA). The suppression pool must quench all the steam released through the downcomer lines during a loss of coolant accident (LOCA). This is the essential mitigative feature of a pressure suppression containment, which ensures that the peak containment pressure is maintained below the maximum allowable pressure for DBAs Q620 psig). The suppression pool must also condense steam from the steam exhaust lines in the turbine driven systems (i.e., High Pressure Coolant Injection (HPCI) System and Reactor Core Isolation Cooling (RCIC) System) and provides the main emergency water supply source for the reactor vessel. [68,i The suppression pool volume ranges betwee nJ ft3 at the low water level limit of limit of inche L. incheda andl ft at the high water level (0 If the suppression pool water level is too low, an insufficient amount of water would be available to adequately condense the steam from the S/RV quenchers, main vents, or HPCI and RCIC turbine exhaust lines. Low suppression pool water level could also result in an inadequate emergency makeup water source to the Emergency Core Cooling System. The lower volume would also absorb less steam energy before heating up excessively. Therefore, a minimum suppression pool water level is specified. If the suppression pool water level is too high, it could result in excessive clearing loads from S/RV discharges and excessive pool swell loads during a DBA LOCA. Therefore, a maximum pool water level is specified. This LCO specifies an acceptable range to prevent the suppression pool water level from being either too high or too low. APPLICABLE Initial suppression pool water level affects suppression pool temperature SAFETY response calculations, calculated drywell pressure during vent clearing ANALYSES for a DBA, calculated pool swell loads for a DBA LOCA, and calculated loads due to S/RV discharges. Suppression pool water level must be maintained within the limits specified so that the safety analysis of Reference I remains valid. Suppression pool water level satisfies Criteria 2 and 3 of 10 CFR 50.36(c)(2)(ii). BWR/4 STS B 3.6.2.2-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 254 of 431
Attachment 1, Volume 11, Rev. 0, Page 255 of 431 Suppression Pool Water Level B 3.6.2.2 BASES LCO 1 limit that suppression pool water level be 2 N 2ft 21inchesM and inchesM is required to ensure that the primary containment 0 conditions assumed for the safety analyses are met. Either the high or low water level limits were used in the safety analyses, depending upon which is more conservative for a particular calculation. APPLICABILITY In MODES 1, 2, and 3, a DBA would cause significant loads on the primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. The requirements for maintaining suppression pool water level within limits in MODE 4 or 5 is addressed in LCO 3.5.2, "ECCS-Shutdown." ACTIONS A.1 With suppression pool water level outside the limits, the conditions assumed for the safety analyses are not met. If water level is below the minimum level, the pressure suppression function still exists as long as 8nmll m e are covered, HPCI and RCIC turbine exhausts are covered, and S/RV quenchers are covered. If suppression pool water level is above the maximum level, protection against overpressurization still exists due to the margin in the peak containment pressure analysis and the capability of the Drywell Spray System. Therefore, continued operation for a limited time is allowed. The 2 hour Completion Time is sufficient to restore suppression pool water level to within limits. Also, it takes into account the low probability of an event impacting the suppression pool water level occurring during this interval. B.1 and B.2 If suppression pool water level cannot be restored to within limits within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. BWR14 STS B 3.6.2.2-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 255 of 431
Attachment 1, Volume 11, Rev. 0, Page 256 of 431 Suppression Pool Water Level B 3.6.2.2 BASES SURVEILLANCE SR 3.6.2.2.1 REQUIREMENTS Verification of the suppression pool water level is to ensure that the
'has been shown to required limits are satisfied. The 24 hour Frequency Ff this SR was be acceptable based develope efsdrng operating experiS expeperating in suppsinol water level aO ie-mld to trendingvcto r level instrument dr ring the 0D ap~al MOE nd to Kising the proximity to th deified LCO Furthermore, the 24 hour Frequency is considered adequate in view of other indications available in the control room, including alarms, to alert the operator to an abnormal suppression pool water level condition.
REFERENCES 1. -7'SAR, Section 00 BWR/4 STS B 3.6.2.2-3 Rev. 3.0, 03/31/04 Attachment 1,Volume 11, Rev. 0, Page 256 of 431
Attachment 1, Volume II, Rev. 0, Page 257 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.2.2 BASES, SUPPRESSION POOL WATER LEVEL
- 1. The brackets have been removed and the proper plant specific information/value has been provided.
- 2. Changes have been made to reflect changes made to the Specification.
- 3. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the plant specific nomenclature, number, reference, system description, analysis description, or licensing basis description.
Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 257 of 431
Attachment 1, Volume 11, Rev. 0, Page 258 of 431 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 11, Rev. 0, Page 258 of 431
Attachment 1, Volume 11, Rev. 0, Page 259 of 431 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.6.2.2, SUPPRESSION POOL WATER LEVEL There are no specific NSHC discussions for this Specification. Monticello Page 1 of I Attachment 1, Volume 11, Rev. 0, Page 259 of 431
Attachment 1, Volume 11, Rev. 0, Page 260 of 431 ATTACHMENT 11 ITS 3.6.2.3, Residual Heat Removal (RHR) Suppression Pool Cooling Attachment 1, Volume 11, Rev. 0, Page 260 of 431
Attachment 1, Volume 11, Rev. 0, Page 261 of 431 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 11, Rev. 0, Page 261 of 431
C C C 0 ITS 3.6.2.3 ITS IN
.See ITS 3.6.1.8}
0) 0 0 0 CD ITS 3.7.1 ) Ad sed Survedtances SR 3.6.23.1 and SR 3.6.2.3.2 M. 0 0 a) (0 (0 One ronWment§pLahoolng Subsystem may be Inoperable for ? days If the requirements of 3.5.C.1 or 2 cannot be met, an orderly shutdown of the reactor wil be InMItWad -4' and the reactor water temperature shall be reduced MI -0 to less than 212F within orabowediut of servxw.ttres for the RH tpi 3.514.5 104 08/01/01 Amendment No. 27,77, 70, 95,102,122 Page 1 of I
Attachment 1, Volume 11, Rev. 0, Page 263 of 431 DISCUSSION OF CHANGES ITS 3.6.2.3, RESIDUAL HEAT REMOVAL (RHR) SUPPRESSION POOL COOLING ADMINISTRATIVE CHANGES A.1 In the conversion of the Monticello Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1433, Rev. 3, "Standard Technical Specifications General Electric Plants, BWR/4" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 CTS 3.5.C.1 Footnote *, which states "For allowed out of service times for the RHR pumps see Section 3.5.A," is a cross reference to another Specification that provides additional requirements associated with the RHR pumps. This cross reference is not included in ITS 3.6.2.3. This changes the CTS by deleting the cross reference to other Specification requirements. The purpose of CTS 3.5.C is to ensure the OPERABILITY of the RHR suppression pool cooling subsystems. ITS 3.6.2.3 does not include any cross references to other Specifications that govern the OPERABILITY requirements for the RHR pumps. This change is acceptable since the other Specifications prescribe the appropriate requirements for the RHR pumps and this cross reference is not necessary. This change is considered administrative because it does not result in any technical changes to the CTS. MORE RESTRICTIVE CHANGES M.1 CTS 3.5.C.1 is applicable when irradiated fuel is in the reactor vessel and reactor water temperature is greater than 21 20F. ITS LCO 3.6.2.3 is applicable in MODES 1, 2, and 3. This changes the CTS by requiring two RHR suppression pool cooling subsystems to be OPERABLE in MODE 2 when reactor water temperature is less than or equal to 212 0F. The purpose of CTS 3.5.C.1 is to ensure the RHR suppression pool cooling subsystems are OPERABLE to mitigate the consequences of a design basis accident. The RHR suppression pool cooling subsystems are required to be OPERABLE during MODES 1,2, and 3 when there is considerable energy in the reactor core and a DBA could cause a significant heat up of the suppression pool. In MODES 1 and 3, the reactor coolant temperature will always be above 2120F. In MODE 2, the reactor coolant temperature may be less than or equal to 2120F when the reactor is subcritical but control rods are withdrawn. Therefore, it is necessary and acceptable to require the RHR suppression pool cooling subsystems to be OPERABLE. This change is designated as more restrictive because the LCO will be applicable under more reactor operating conditions than in the CTS. M.2 Currently, the CTS does not provide any specific Surveillances to verify OPERABILITY of the RHR suppression pool cooling subsystems. ITS SR 3.6.2.3.1 requires verification that each RHR suppression pool cooling Monticello Page 1 of 3 Attachment 1, Volume 11, Rev. 0, Page 263 of 431
Attachment 1, Volume 11, Rev. 0, Page 264 of 431 DISCUSSION OF CHANGES ITS 3.6.2.3, RESIDUAL HEAT REMOVAL (RHR) SUPPRESSION POOL COOLING subsystem manual and power operated valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position or can be aligned to the correct position every 31 days. ITS SR 3.6.2.3.2 requires verification that each required RHR pump develops a flow rate greater than 3780 gpm through the associated heat exchanger while operating in the suppression pool cooling mode, in accordance with the Inservice Testing Program. This changes the CTS by adding these Surveillance Requirements to the Technical Specifications. The purpose of ITS SR 3.6.2.3.1 is to provide assurance that the proper flow paths will exist for suppression pool cooling operation. The purpose of ITS SR 3.6.2.3.2 is to ensure the pumps can meet the flow rate requirements assumed in the accident analysis. This change is acceptable because it provides additional assurance that the RHR suppression pool cooling will be capable of performing its function. This change is designated as more restrictive because it adds Surveillance Requirements to the CTS. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA.1 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 3.5.C.1 states that a containment cooling subsystem consists of the following equipment powered from one division: 1 RHR Heat Exchanger, 1 RHR Pump, and valves and piping necessary for torus cooling. ITS 3.6.2.3 requires two RHR suppression pool cooling subsystems to be OPERABLE, but the details of what constitutes an OPERABLE subsystem are moved to the ITS Bases. This changes the CTS by moving the details of what constitutes an OPERABLE subsystem to the ITS Bases. The removal of these details, which are related to system design, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement for two RHR suppression pool cooling subsystems to be OPERABLE. Also, this change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program InChapter 5 of the ITS. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications. Monticello Page 2 of 3 Attachment 1, Volume 11, Rev. 0, Page 264 of 431
Attachment 1, Volume 11, Rev. 0, Page 265 of 431 DISCUSSION OF CHANGES ITS 3.6.2.3, RESIDUAL HEAT REMOVAL (RHR) SUPPRESSION POOL COOLING LESS RESTRICTIVE CHANGES L.1 (Category 4 - Relaxation of Required Action) When two RHR suppression pool cooling subsystems are inoperable, a unit shutdown is required by CTS 3.5.C.3; no time is provided to restore a subsystem. With two RHR suppression pool cooling subsystems inoperable, ITS 3.6.2.3 ACTION B will allow 8 hours to restore one inoperable RHR suppression pool cooling subsystem prior to requiring a unit shutdown. This changes the CTS by allowing 8 hours to restore one of two inoperable RHR suppression pool cooling subsystems prior to requiring a unit shutdown. The purpose of CTS 3.5.C is to require sufficient containment cooling to ensure the primary containment conditions for the safety analyses are met. The proposed 8 hour Completion Time is acceptable since an immediate shutdown has the potential to result in a unit scram and discharge of steam to the suppression pool, when both suppression pool cooling subsystems are inoperable and incapable of removing the generated heat. The 8 hours provides some time to restore one of the subsystems prior to requiring a shutdown (thus precluding the potential problem described above), yet is short enough that it does not significantly increase the probability of an accident to occur during this additional time. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. L.2 (Category 3 - Relaxation of Completion Time) CTS 3.5.C.3 requires the unit to be shut down and reactor water temperature reduced to less than 21 20F within 24 hours if the requirements of CTS 3.5.C.1 or CTS 3.5.C.2 are not met. Under similar conditions (as modified by DOC L.1), ITS 3.6.2.3 ACTION C requires the reactor be in MODE 3 in 12 hours and in MODE 4 in 36 hours. This changes the CTS by requiring the unit to be in MODE 3 in 12 hours and by extending the time to reduce reactor water temperature to < 212 0 F (i.e., MODE 4) from 24 hours to 36 hours. The purpose of CTS 3.5.C.3 is to place the unit outside the Applicability of the Specification within a reasonable amount of time. This change is acceptable because the Completion Time is consistent with the safe operation under the specified Condition, considering the OPERABLE status of the redundant systems or features. This includes the capacity and capability of remaining systems or features, a reasonable time for repairs or replacement, and the low probability of a DBA occurring during the allowed Completion Time. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. This change is also acceptable because it requires the unit to be in an intermediate condition (MODE 3) sooner than is currently required (12 hours versus 24 hours). This portion of the change reduces the amount of time the unit would be allowed to continue to operate in MODES I and 2 once the condition is identified. The consequences of a pressurization event are significantly reduced when the reactor is shutdown and a controlled cooldown is already in progress. This change is designated as less restrictive because additional time is allowed to place the unit outside the LCO Applicability than is allowed in the CTS. Monticello Page 3 of 3 Attachment 1, Volume 11, Rev. 0, Page 265 of 431
Attachment 1, Volume 11, Rev. 0, Page 266 of 431 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 266 of 431
Attachment 1, Volume 11, Rev. 0, Page 267 of 431 RHR Suppression Pool Cooling 3.6.2.3 CTS 3.6 CONTAINMENT SYSTEMS 3.5.C 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling 3.5.C.1 LCO 3.6.2.3 Two RHR suppression pool cooling subsystems shall be OPERABLE. 3.5.C.1 APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 3.5.C.2 A. One RHR suppression A.1 Restore RHR suppression 7 days pool cooling subsystem pool cooling subsystem to inoperable. OPERABLE status. DOCL.1 B. Two RHR suppression B.1 Restore one RHR 8 hours pool cooling subsystems suppression pool cooling inoperable. subsystem to OPERABLE status. 3.5.C.3 C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time not met. AND C.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE J FREQUENCY DOC M-2 SR 3.6.2.3.1 eVerifv each RHR suppressionpoolcooling 31 days subsystem manua l power operated and maic 0 valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position or can be aligned to the correct position. BWR/4 STS 3.6.2.3-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 267 of 431
Attachment 1, Volume 11, Rev. 0, Page 268 of 431 RHR Suppression Pool Cooling 3.6.2.3 Qt CTS SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY r DOC M.2 SR 3.6.2.3.2 Vify eacO'RHR pump develops gpm through a flow the associated rate heat Eln accordance 0 with the Inservice exchanger while operating in the suppression pool Testing Program 0 cooling mode. cr9 Raysl BWR/4 STS 3.6.2.3-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 268 of 431
Attachment 1, Volume 11, Rev. 0, Page 269 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.2.3, RESIDUAL HEAT REMOVAL (RHR) SUPPRESSION POOL COOLING
- 1. The Monticello design does not include any automatically actuated suppression pool cooling valves. The suppression pool cooling mode is manually actuated.
Therefore, the reference to "automatic" in ISTS SR 3.6.2.3.1 has been deleted.
- 2. The Monticello CTS and safety analysis only require one of the two RHR pumps in a suppression pool cooling subsystem to be OPERABLE. Therefore, ITS SR 3.6.2.3.2 has been modified to only require the "required" RHR pumps to be tested. This change is consistent with the use of the word required in the ITS.
- 3. The brackets have been removed and the proper plant specific information/value has been provided.
Monticello Page 1 of I Attachment 1, Volume 11, Rev. 0, Page 269 of 431
Attachment 1, Volume 11, Rev. 0, Page 270 of 431 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 270 of 431
Attachment 1, Volume 11, Rev. 0, Page 271 of 431 RHR Suppression Pool Cooling B 3.6.2.3 B 3.6 CONTAINMENT SYSTEMS B 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling BASES BACKGROUND Following a Design Basis Accident (DBA), the RHR Suppression Pool Cooling System removes heat from the suppression pool. The suppression pool is designed to absorb the sudden input of heat from the primary system. In the long term, the pool continues to absorb residual heat generated by fuel in the reactor core. Some means must be provided to remove heat from the suppression pool so that the temperature inside the primary containment remains within design limits. This function is provided by two redundant RHR suppression pool cooling subsystems. The purpose of this LCO is to ensure that both subsystems are OPERABLE in applicable MODES. Each RHR subsystem contains two pumps and one heat exchanger and is manually initiated and independently controlled. The two subsystems perform the suppression pool cooling function by circulating water from the suppression pool through the RHR heat exchangers and returning it to the suppression pool. RHR service water, circulating through the tube side of the heat exchangers, exchanges heat with the suppression pool water and discharges this heat to thel heat sinkate The heat removal capability of one RHR pump in one subsystem is sufficient to meet the overall DBA pool cooling requirement for loss of coolant accidents (LOCAs) and transient events such as a turbine trip or stuck open safety/relief valve (S/RV). S/RV leakage and high pressure core injection and Reactor Core Isolation Cooling System testing increase suppression pool temperature more slowly. The RHR Suppression Pool Cooling System is also used to lower the suppression pool water bulk temperature following such events. APPLICABLE Reference 1 contains the results of analyses used to predict primary SAFETY containment pressure and temperature following large and small break ANALYSES LOCAs. The intent of the analyses is to demonstrate that the heat removal capacity of the RHR Suppression Pool Cooling System is adequate to maintain the primary containment conditions within design limits. The suppression pool temperature is calculated to remain below the design limit. The RHR Suppression Pool Cooling System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). BWR/4 STS B 3.6.2.3-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 271 of 431
Attachment 1, Volume 11, Rev. 0, Page 272 of 431 RHR Suppression Pool Cooling B 3.6.2.3 BASES LCO During a DBA, a minimum of one RHR suppression pool cooling subsystem is required to maintain the primary containment peak pressure and temperature below design limits (Ref. 1). To ensure that these requirements are met, two RHR suppression pool cooling subsystems must be OPERABLE with power from two safety related independent power supplies. Therefore, in the event of an accident, at least one subsystem is OPERABLE assuming the worst case single active failure. An RHR suppression pool cooling subsystem is OPERABLE when one of the pumps, the heat exchanger, and associated piping, valves, instrumentation, and controls are OPERABLE. APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment and cause a heatup and pressurization of primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Therefore,-the RHR Suppression Pool Cooling System is not required to be OPERABLE in MODE 4 or 5. ACTIONS A.1 With one RHR suppression pool cooling subsystem inoperable, the inoperable subs ystem must be restored to OPERABLE status within O 7 days. In this sondition, the remainingRHR suppression pool cooling Esu system is adequate to perform the primary containment cooling 00 function. However, the overall reliability is reduced because a single failure in the OPERABLE subsystem could result in reduced primary containment cooling capability. The 7 day Completion Time is acceptable in light of the redundant RHR suppression pool cooling capabilities afforded by the OPERABLE subsystem and the low probability of a DBA occurring during this period. B.1 With two RHR suppression pool cooling subsystems inoperable, one subsystem must be restored to OPERABLE status within 8 hours. In this condition, there is a substantial loss of the primary containment pressure and temperature mitigation function. The 8 hour Completion Time is based on this loss of function and is considered acceptable due to the low probability of a DBA and the potential avoidance of a plant shutdown transient that could result in the need for the RHR suppression pool cooling subsystems to operate. BWR/4 STS B 3.6.2.3-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 272 of 431
Attachment 1, Volume 11, Rev. 0, Page 273 of 431 RHR Suppression Pool Cooling B 3.6.2.3 BASES ACTIONS (continued) C.1 and C.2 If the Required Action and associated Completion Time cannot be met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems: SURVEILLANCE SR 3.6.2.3.1 REQUIREMENTS Verifying the correct alignment for manual--power operated 0 autoffaticvalves in the RHR suppression pool cooling mode flow path provides assurance that the proper flow path exists for system operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position within the time assumed in the accident analysis. This is acceptable since the RHR suppression pool cooling mode is manually initiated. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. The Frequency of 31 days is justified because the valves are operated under procedural control, improper valve position would affect only a single subsystem, the probability of an event requiring initiation of the system is low, and the subsystem is a manually initiated system. This Frequency has been shown to be acceptable based on operating experience. SR 3.6.2.3.2 Verifying that each RHR pump develops a flow rate 2a gpm while operating in the suppression pool cooling mode with flow t rough the 0D associated heat exchanger ensures that pump performance has not degraded during the cycle. Flow is a normal test of centrifugal pump performance required by ASMEnCodeiTS nrXi (Ref. 2). This testI confirms one point on the pump design curve, and the results are 0 indicative of overall performance. Such inservice iinspetions confirm e component OPERABILITY, trend performance, and detect incipient 0D failures by indicating abnormal performance. The Frequency of this SR is Oin accordance with the Inservice Testing Program r a s. 0 BWR/4 STS B 3.6.2.3-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 273 of 431
Attachment 1, Volume 11, Rev. 0, Page 274 of 431 RHR Suppression Pool Cooling B 3.6.2.3 BASES REFERENCES 17SAR, Sectionm? 2.and 5.2.3 AM Boiler PressienCod n (OM) Code. l 00
- 2. ASMEI, Boiler and Press~ o e Section XlL (
BWR/4 STS B 3.6.2.3-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 274 of 431
Attachment 1, Volume 11, Rev. 0, Page 275 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.2.3 BASES, RESIDUAL HEAT REMOVAL (RHR) SUPPRESSION POOL COOLING
- 1. Typographical/grammatical error corrected.
- 2. Changes have been made for consistency with similar phrases in other parts of the Bases.
- 3. Changes have been made to reflect changes made to the Specification.
- 4. The brackets have been removed and the proper plant specific information/value has been provided.
- 5. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the plant specific nomenclature, number, reference, system description, analysis description, or licensing basis description.
Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 275 of 431
Attachment 1, Volume 11, Rev. 0, Page 276 of 431 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 11, Rev. 0, Page 276 of 431
Attachment 1, Volume 11, Rev. 0, Page 277 of 431 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.6.2.3, RESIDUAL HEAT REMOVAL (RHR) SUPPRESSION POOL COOLING There are no specific NSHC discussions for this Specification. Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 277 of 431
Attachment 1,Volume 11, Rev. 0, Page 278 of 431 ATTACHMENT 12 ITS 3.6.3.1, Primary Containment Oxygen Concentration Attachment 1, Volume 11, Rev. O,Page 278 of 431
Attachment 1, Volume 11, Rev. 0, Page 279 of 431 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 11, Rev. 0, Page 279 of 431
( C C 3 ITS 3.6.3.1 ITS ITS 3.0 LIMITING CONDITIONS FOR OPERATION. 4.0 SURVEILLANCE REQUIREMENTS I I.
- e. One position alarm circuit can be inoperable b. When the position of any drywell-suppression C,
0 providing that the redundant position alarm chamber vacuum breaker valve IsIndicated to C, 0) 0 circuit is operable. Both position alarm circuits be not fully closed at a time when such closure 0 may be inoperable for a period not to exceed Is required, the drywell to suppression chamber to seven days provided that all vacuum breakers differential pressure decay shall be =1 are operable. demonstrated to be less than that shown on Figure 3.7.1 Immediately and following any evidence of subsequent operation of the 0 0 f. If requirements of 3.7.A.4 cannot be met, the See ITS 3.6.1.7} reactor shall be placed in a Cold Shutdown Inoperable valve until the Inoperable valve is a I condition within 24 hours. restored to a normal condition. S 5
- c. When both position alarm circuits are made or a 0.
found to be Inoperable, the control panel indicator light status shall be recorded daily to -4 0 -o detect changes In the vacuum breaker position.
-i At 3.6.3.1 1 5. Primary Containment Oxygen Concentration 5. Primary Containment Oxygen Concentration I CAD CD
-0 SR 3.6.3.1.1 Whenever inerting is required, the primary ;)
- i a. Fie primary containment atmosphere shall be containment oxygen concea shall be tos LCO 3.6.3.1 educed to less than 4% oxygen by volume@3]
At measured irod on a weekly basis.
- I senever the reactor Is In the run>
0, co e, except as specified In 3.7.A.5.b. - 4' CDt
- b. Nithin the 24-hour period after Thermal Power Appricabiity is > 15% Rated Thermnal Power following co 00 startup, to 24 hours prior to reducing Thermal 4T6 Power to < 15% Rated Thermal Power prior to Wo tho next scheduled reactor shutdownTjho _l containment atmosphfere oxygen concentration LCO 3.6.3.1 shall be reduced to less than 4% by volume, Land maintained In this condition.
3.714.7 165 09/23/02 Amendment No. 494, 130 Page 1 of 2
c C ITS 3.6.3.1 C., ITS 0 3.0 LIMITNG CONDmONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS ACTION A c. Whenever primary containment oxygen 1W concentration Is equal to or exceeds 4% by C) C, volume, except as permitted by 3.7.A.5.b above, within the subsequent 24 hour period 2 CD return the oxygen concentration to less than 4% CD by volume. ACTION B d. If the requirements of 3.7A.5 cannot be met, reduce Thermal Power to s 15% Rated 0 Thermal Power, within 8 hours. 0 B. Standby Gas Treatment System
- 1. Two separate and Independent standby gas CD treatment system circuits shall be operable at all B. Standby Gas Treatment System CD times when secondary containment Integrity Is required, except as specified In sections 3.7.B.1.(a) 1. Once per month, operate each train of the standby and (b). gas treatment system for 10 continuous hours with the inrine heaters operating. CD
- a. After one of the standby gas treatment system
.71 circuits is made or found to be Inoperable for See ITS 3.6.4.3} any reason, reactor operation and fuel handling (0 Is permissible only during the succeeding seven CD days, provided that all active components in the other standby gas treatment system are operable. Within 36 hours following the 7 days. the reactor shall be placed in a condition for 0
-A' which the standby gas treatment system is not required in accordance with Specification 3.7.C.2.(a) through (d).
3.7/4.7 166 09/23/02 Amendment No. 6 6, 77, 04
,130 Page 2 of 2
Attachment 1, Volume 11, Rev. 0, Page 282 of 431 DISCUSSION OF CHANGES ITS 3.6.3.1, PRIMARY CONTAINMENT OXYGEN CONCENTRATION ADMINISTRATIVE CHANGES A.1 In the conversion of the Monticello Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1433, Rev. 3, "Standard Technical Specifications General Electric Plants, BWR/4" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 CTS 4.7.A.5 states that whenever inerting is required, the primary containment oxygen concentration shall be measured and recorded on a weekly basis. Under similar conditions, ITS SR 3.6.3.1.1 requires a verification that the primary containment oxygen concentration is within limits, but does not include this requirement to record the primary containment oxygen concentration. This changes the CTS by deleting the explicit requirement to record the primary containment oxygen concentration. The purpose of CTS 4.7.A.5 is to verify the primary containment oxygen concentration is within limits. This change is acceptable because the requirement to record oxygen concentration duplicates the requirements of 10 CFR 50 Appendix B, Section XVII (Quality Assurance Records) to maintain records of activities affecting quality, including the results of tests (i.e., Technical Specification Surveillances). Compliance with 10 CFR 50 Appendix B is required by the Monticello Operating License, which is adequate to ensure appropriate data is taken and maintained. The details of the regulations within the Technical Specifications are repetitious and unnecessary. Therefore, retaining the requirement to perform the associated Surveillance and eliminating the detail from Technical Specifications that is found in 10 CFR 50 Appendix B is considered a presentation preference. As such, this change is considered an administrative change. MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA.1 (Type 3 - Removing ProceduralDetails for Meeting TS Requirements or Reporting Requirements) CTS 3.7.A.5.a states that the primary containment atmosphere shall be reduced to less than 4% oxygen by volume "with nitrogen gas." ITS LCO 3.6.3.1 states that primary containment oxygen concentration shall be < 4 volume percent, but does not state that it is to be done using Monticello Page 1 of 2 Attachment 1, Volume 11, Rev. 0, Page 282 of 431
Attachment 1, Volume 11, Rev. 0, Page 283 of 431 DISCUSSION OF CHANGES ITS 3.6.3.1, PRIMARY CONTAINMENT OXYGEN CONCENTRATION nitrogen gas. This changes the CTS by relocating the details on how to reduce oxygen concentration (i.e., "with nitrogen gas") to the ITS Bases. The removal of these details for maintaining the LCO requirements from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement that the primary containment oxygen concentration shall be < 4 volume percent. Also, this change is acceptable because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the CTS. LESS RESTRICTIVE CHANGES None Monticello Page 2 of 2 Attachment 1, Volume 11, Rev. 0, Page 283 of 431
Attachment 1, Volume 11, Rev. 0, Page 284 of 431 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1,Volume 11, Rev. 0, Page 284 of 431
Attachment 1, Volume 11, Rev. 0, Page 285 of 431 Primary Containment Oxygen Concentration 3.6.3 CTS 3.6 CONTAINMENT SYSTEMS 3.7A.5 3.6.3LL Primary Containment Oxygen Concentration 0D 3.7A.5.a, 3.7-A.5.b LCO 3.6.3 The primary containment oxygen concentration shall be < 4.0 volume percent. 0D APPLICABILITY: MODE I during the time period: 3.7A.5.a, 3.7A.5.b a. From [24hours after THERMAL POWER is > [1 5P/o RTP following ( startup, to
- b. @ hours prior to reducing THERMAL POWER to < 15p/o RTP prior 0 to the next scheduled reactor shutdown.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME KI3.7.A.5.c A. Primary containment A.1 Restore oxygen 24 hours oxygen concentration concentration to within limit. not within limit. 3.7A5.d B. Required Action and B.1 Reduce THERMAL 8 hours associated Completion Time not met. POWER to
- V15r/o RTP.
0 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4J7A5 SR 3 .6 .3 .F.1 Verify primary containment oxygen concentration is within limits. 7 days 0 BWR/4 STS 3.6.3.2-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 285 of 431
Attachment 1, Volume 11, Rev. 0, Page 286 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.3.1, PRIMARY CONTAINMENT OXYGEN CONCENTRATION
- 1. ISTS 3.6.3.2 is renumbered as ITS 3.6.3.1 as a result of the deletion of ISTS 3.6.3.1, "Drywell Cooling System Fans."
- 2. The brackets have been removed and the proper plant specific information/value has been provided.
Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 286 of 431
Attachment 1, Volume 11, Rev. 0, Page 287 of 431 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 287 of 431
Attachment 1, Volume 11, Rev. 0, Page 288 of 431 Primary Containment Oxygen Concentration B 3.6.3 B 3.6 CONTAINMENT SYSTEMS B 3.6.3Pmary Containment Oxygen Concentration 0 BASES BACKGR DUND .AlInuc eaceors must designed to withstand events that generate hydrogen either due to the zirconium metal water reaction in the core or ed ndue to radiolsis. The primary method to control hydrogen is to inert the (I) gas primary containmen. With the primary containment inert, that is, oxygen concentration < 4.0 volume percent (v/o), a combustible mixture cannot be present in the primary containment for any hydrogen concentration. An event that rapidly generates hydrogen from zirconium metal water reaction will result in excessive hydrogen in primary containment, but oxygen concentration will remain < 4.0 v/o and no combustion can occur. This LCO ensures that oxygen concentration does not exceed 4.0 v/o during operation'in the applicable conditions. APPLICABLE The Reference 1 calculations assume that the primary containment is SAFETY inerted when a Design Basis Accident loss of coolant accident occurs. ANALYSES Thus, the hydrogen assumed to be released to the primary containment as a result of metal water reaction in the reactor core will not produce combustible gas mixtures in the primary containment. K) Primary containment oxygen concentration satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO The primary containment oxygen concentration is maintained < 4.0 v/o to ensure that an event that produces any amount of hydrogen does not result in a combustible mixture inside primary containment. n e 0 APPLICABILITY The primary containment oxygen concentration must be within the specified limit when primary containment is inerted, except as allowed by the relaxations during startup and shutdown addressed below. The primary containment must be inert in MODE 1, since this is the condition and oxygen with the highest probability of an event that could produce hydrogen* 0 Inerting the primary containment is an operational problem because it prevents containment access without an appropriate breathing apparatus. Therefore, the primary containment is inerted as late as possible in the an startup and de-inerted as soon as possible in the plant shutdown. As Io a r15% RTP, the potential for an event that generates significant hydrogeels low and the primary containment need not be inert. Furthermore, the probability of an event that generates 0 hydrogen occurring within the first j24J hours of a startup, or within the lastR24j hours before a shutdown, Is low enough that these 'windows," when the primary containment Is not inerted, are also justified. The 0 0241 hour time period is a reasonable amount of time to allow plant personnel to perform inerting or de-inerting. K) BWR/4 STS B 3.6.3.2-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 288 of 431
Attachment 1, Volume 11, Rev. 0, Page 289 of 431 Primary Containment Oxygen Concentrationy( B 3.6.3 BASES ACTIONS A.1 If oxygen concentration is 2 4.0 v/o at any time while operating in MODE 1, with the exception of the relaxations allowed during startup and shutdown, oxygen concentration must be restored to < 4.0 v/o within 24 hours. The 24 hour Completion Time is allowed when oxygen concentration-is 2 4.0 v/o because of the low probability and long duration of an event that would generate significant amounts of hydrogen occurring during this period. land 0D B.1 If oxygen concentration cannot be restored to within limits within the potential foran event that generates significant required Completion Time, the plant must be brought to la MO t e not app . To achieve this status, power must be reduced whic 0 hydrogen and oxygen Is low to s I % RTP within 8 hours. The 8 hour Completion Time is reasonable, based on operating experience, to reduce reactor power from 0 full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE REQUIREMENTS The primary containment must be determined to be iner by verifying that ed, oxygen concentration is < 4.0 v/o. The 7 day Frequency is based on the ) slow rate at which oxygen concentration can changed on othercod; indications of abnormal conditions (whichg tead to more frequen checking by operators in accordance with plant procedures). Also, this Frequency has been shown to be acceptable through operating experience. REFERENCES E9-ISAR, Section 00( BWR/4 STS B 3.6.3.2-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 289 of 431
Attachment 1, Volume 11, Rev. 0, Page 290 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.3.1 BASES, PRIMARY CONTAINMENT OXYGEN CONCENTRATION
- 1. The Bases has been renumbered due to the deletion of ISTS Bases 3.6.3.1.
- 2. Editorial change made for enhanced clarity.
- 3. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the plant specific nomenclature, number, reference, system description, analysis description, or licensing basis description.
- 4. The brackets have been removed and the proper plant-specific information/value has been provided.
- 5. Typographical/grammatical error corrected. I
- 6. The Applicability of this Specification is MODE 1, however it is modified by providing an allowance to not meet the LCO requirements during the specified time period for startups and scheduled reactor shutdowns. The Required Action B.1 Bases states that reducing power to < 15% RTP places the plant in a MODE in which the LCO does not apply. This statement is not consistent with the Applicability; therefore this statement has been modified to be consistent with the description in the Bases Applicability. In addition, the reactor power level of "< 15% RTP" specified in the Bases Applicability has been changed to "< 15% RTP" to be consistent with the power level specified in Required Action B.1.
Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 290 of 431
Attachment 1, Volume 11, Rev. 0, Page 291 of 431 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 11, Rev. 0, Page 291 of 431
Attachment 1, Volume II, Rev. 0, Page 292 of 431 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.6.3.1, PRIMARY CONTAINMENT OXYGEN CONCENTRATION There are no specific NSHC discussions for this Specification. Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 292 of 431
, Volume 11, Rev. 0, Page 293 of 431 ATTACHMENT 13 ITS 3.6.4.1, Secondary Containment , Volume 11, Rev. 0, Page 293 of 431
Attachment 1, Volume 11, Rev. 0, Page 294 of 431 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 11, Rev. 0, Page 294 of 431
C C ITS 3.6.4.1 ITS ITS 3.0 UMING CONDmONS FOR OPERATION 4.0 SURVEILLANCE REQUtREMENTS C. Secondary Containment C. Secondary Containment
- 0) Secondary containment surveillance shall be LCO 3.6.4.1 performed as Indicated bebw Add proposed SR 3.6.4.1.1
- a. Secondary contairnment capablity to maintain/
0 at least a 114 Inch of water vacuu 0 LCO 3.6.4.1 hAnda(y containm nt cap ailter train _ CD CD Applicability CD -o1 Verification that each automatic damper -o actuates to Rs Isolation position shall be 0) 0) Eo performed: Co (D (1) Each refuellng outage. CD (2) After maintenance, repair or reptacement work Is performed on the damper or Its I CD associated actuator, control circuit, or power circuit. (0 C' CD) 0
-P.
I 169 1012195 3.714.7 Amendment No. 3, 63, 76, 94 Page 1 of 4
Attachment 1, Volume 11, Rev. 0, Page 296 of 431 0 ITS 3.6.4.1 ITS INSERT A
- d. Recently irradiated fuel is not being moved within secondary Applicability containment.
- e. Operations with the potential for draining the reactor vessel are not being performed.
INSERTB
/--,X
<lAdd proposed ACTIN A- l I L.
During Run, Startup, or Hot Shutdown if Specifications 3.7.C.1 through 3.7.C.3 cannot be ACTION B met, initiate a normal orderly shutdown and have the reactor, in the Cold Shutdown condition within 36 hour And 4.b If Specifications 3.7.C.1 through 3.7.C.3 cannot be met, immediately suspend: Add proposed Required Action C.1 Note a ) ACTION C- 1. Operations with a potential for draining the reactor vessel.
- 2. Handling of recently irradiated fuel in the secondary containment.
- 13. Moe a fuel cask in or building.
Insert Page 169 Page 2 of 4 Attachment 1, Volume 11, Rev. 0, Page 296 of 431
C C, ITS 3.6.4.1 C 0 3.0 UMmNG CONDmOONS FOR OPERATION 4.0 SURVEILLANCE REOUIREMENTS I _ f [ reactor core, a eratlons with a potenjial for reducing the shutdo aiergln below that spqclfied In 03 specification 8.3.A. and handiing opadiated fuel or the huel cas Inthe secondary cordelnment are to A.6e CD be Immedieley suspended if secfndazy contaimnm kitesiwv is not man ained. 0 D. Primary ContaInment IsolatIon Valves (PCIVs) D. Primery Containrment Isolation Valves (PCIVs)
- 1. During reactor power operating conditIons, all 1. The primary containment automatic Isolation valve (D Primary Containment automatic Isolation valves and surveillance shall be performed as follows:
0 all primary system Instrument lne flow check valves 0 shall be operable except as specified In 3.7.D.2 and a. At least once per operating cycle the operable F I 3.7.D.3. _ Isolation valves that are power operated and automatIcally Initiated shall be tested for i sImulated automatic InitiatIon and closure times. 0) 03 b. At least once per operaing cycle the primary CD system Instrument lne flow check valves shall be tested for proper operation. CD
- -1
- c. A normally open power-operated Isolation
\{SSee ITS 3.6.1.3 } -A 0
valves shall be tested In accordance with the -CD Inservice Testing Program. Main Steam -0 isolation valves shall be tested (one at a time) with the reactor power less than 75% of rated.
- d. At least once per week the main steam-line power-operated Isolation valves shall be exercised by partial closure and subsequent reopening.
3.714.7 170 01/28/05 Amendment No. 1471. 77. 122. t20, 141 Page 3 of 4
C 0 C ITS 3.6.4.1 CI. ITS
- 4. Protective Function - A system protective action which results from the protective action of the channets monhoring a particular plant condition.
thermal megawatts. C) R. Rated Neutron Flux - Rated flux is the neutron flux that corresponds to a steady-state power level of 1775 3 S. Rated Thermal Power - Rated thermal power means a steady-state power level of 1775 thermal megawatts. CD 0 listed In T. Reactor Coolant System Pressure or Reactor Vessel Pressure - Unless otherwise Indicated, reactor vessel pressures See ITS Chapter 1.0} 3 the Technical Specifications are those existing in the vessel steam space. CD temperature Is less U. Refueling Operation and Refuelina Outage - Refueling Operation Is any operation when the reactor water 0 2 0 F and movement of fuel or core components Is In progress. For the purpose of designating frequency of testing and than 21 a regularly scheduled refueling outage; however, where such outages occur within 8 Z., surveillance, a refueling outage shall mean outage, the required surveillance testing need not be performed until the next 3 months of the completion of the previous refueling CD n. regularly scheduled outage. 0 I are assured. V. Saftyitmf - The safety limits are limits below which the maintenance of the cladding and primary system integrity Co plant shutdown and review by the Commission before resuvnption of plant operation. Exceeding such a limit Is cause for Operation beyond such a limit may not in Itself result In serious consequences but Rindicates an operational deficiency subject -o I )dIl 05 L LU IV W (D SR 3.6.4.1.2 W. - Secondary Containmen means that the reactr building is dosed n _oni ona are B sup. w~ co CD -0 SR 3.6.4.1.3 1. At least one door In each access opening Is dosed. CD See TS 3 6.4 .
- 2. The standby gas treatment system Is oprable 0
(See ITS 3.6.4.2}
- 3. All reactor building ventilation system automatic isolation valves are operable or are secured in th ces posionS This determination X. Sensor Check - A qualitative determination of operability by observation of sensor behavior during operation.
possible, comparison with other Independent sensors measuring the same variable. See ITS Chapter 1.0 shelf Inude, where 4 9116198 1.0 Amendment No. 47,102 Page 4 of 4
Attachment 1, Volume 11, Rev. 0, Page 299 of 431 DISCUSSION OF CHANGES ITS 3.6.4.1, SECONDARY CONTAINMENT ADMINISTRATIVE CHANGES A.1 Inthe conversion of the Monticello Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1433, Rev. 3, "Standard Technical Specifications General Electric Plants, BWR/4" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 CTS 3.7.C.1 requires the Secondary Containment Integrity to be maintained and CTS 1.0.W and CTS 3.7.C.2 use the term Secondary Containment Integrity. ITS LCO 3.6.4.1 requires the secondary containment to be OPERABLE. This changes the CTS by deleting the specific Secondary Containment Integrity term and replacing it with a requirement for the secondary containment to be OPERABLE. The purpose of CTS 3.7.C.1 isto provide requirements pertaining to secondary containment OPERABILITY. This change is acceptable since all requirements of the CTS definition of Secondary Containment Integrity in CTS 1.0.W have been incorporated into ITS 3.6.4.1, ITS 3.6.4.2, or ITS 3.6.4.3. ITS 3.6.4.1 requires that the secondary containment to be OPERABLE. The definition of OPERABLE and the subsequent ITS 3.6.4.1 LCO, ACTIONS, and Surveillance Requirements are sufficient to encompass the applicable requirements of the CTS definition. This change is designated as administrative because it does not result in technical changes to the CTS. A.3 CTS 3.7.C.1 specifies requirements for the secondary containment during "all modes of plant operation." However, CTS 3.7.C.2 states that secondary containment is not required "when all of the listed conditions are satisfied," and provides a list of six conditions (CTS 3.7.C.2.a through f). ITS LCO 3.6.4.1 specifies requirements for the secondary containment in the positive sense (when the secondary containment is required to be OPERABLE). This changes the CTS by specifying the requirements for the secondary containment when it is required to be OPERABLE instead of when it is not required to be OPERABLE. Changes to the list of six conditions is discussed In DOCs M.1 and L.1. The purpose of CTS 3.7.C.1 and CTS 3.7.C.2 is to state when the secondary containment requirements are applicable. CTS 3.7.C.1 specifies requirements for the secondary containment during "all modes of plant operation." However, CTS 3.7.C.2 states that secondary containment is not required "when all of the listed conditions are satisfied." ITS LCO 3.6.4.1 specifies requirements for the secondary containment in the positive sense (when the secondary containment is required to be OPERABLE). This change is acceptable because it clearly defines the conditions for when the secondary containment requirements must be met and is consistent with the format specified for other systems and parameters. This change is designated as administrative since it does not result in a technical change to the CTS. Monticello Page 1 of 7 Attachment 1, Volume 11, Rev. 0, Page 299 of 431
Attachment 1, Volume 11, Rev. 0, Page 300 of 431 DISCUSSION OF CHANGES ITS 3.6.4.1, SECONDARY CONTAINMENT A.4 CTS 4.7.C.1.a requires the secondary containment capability test to be performed at "each refueling interval." ITS SR 3.6.4.1.4 requires this same test, however it is required to be performed every "24 months." This changes the CTS by changing the Frequency from "each refueling interval" to "24 months." This change is acceptable because the current "refueling interval" is "24 months." In letter L-MT-04-036, from Thomas J. Palmisano (NMC) to the USNRC, dated June 30, 2004, NMC has proposed to extend the fuel cycle from 18 months to 24 months and the same time has performed an evaluation in accordance with Generic Letter 91-04 to extend the unit Surveillance Requirements from 18 months to 24 months. CTS 4.7.C.1.a was included in this evaluation. This change is designated as administrative because it does not result in any technical changes to the CTS. A.5 CTS 3.7.C.4.b requires the unit to suspend handling of recently irradiated fuel. ITS 3.6.4.1 ACTION C includes the same requirement, however, ITS 3.6.4.1 Required Action C.1 includes a Note that states that LCO 3.0.3 is not applicable. This changes the CTS by adding this Note. The purpose of CTS 3.7.C.4.b is to provide the appropriate actions when the secondary containment is inoperable and the unit is moving recently irradiated fuel in the secondary containment. This change adds a Note that states LCO 3.0.3 is not applicable. This Note has been added because ITS LCO 3.0.3 has been added to ITS Section 3.0 in accordance with DOC M.1. This Note is necessary because if moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of recently irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown. Since ITS LCO 3.0.3 is not currently included in the CTS this change is considered administrative. This change is designated as administrative because it does not result in technical changes to the CTS. A.6 These changes to CTS 3.7.C.2.c, CTS 3.7.C.2.d, and CTS 3.7.C.4, and the addition of CTS 3.7.C.2.d and e are provided in the Monticello ITS consistent with the Technical Specifications Change Request submitted to the USNRC for approval in NMC letter L-MT-05-013, from Thomas J. Palmisano (NMC) to USNRC, dated April 12, 2005. As such, these changes are administrative. A.7 This change to CTS 4.7.C.1.a is provided in the Monticello ITS consistent with the Technical Specifications Change Request submitted to the USNRC for approval in NMC letter L-MT-04-036, from Thomas J. Palmisano (NMC) to USNRC, dated June 30, 2004. As such, this change is administrative. MORE RESTRICTIVE CHANGES M.1 CTS 3.7.C.2.a and 3.7.C.2.b state that the secondary containment requirements are not required when both the reactor is subcritical and Specification 3.3.A is met, and reactor water temperature is below 212oF, respectively. ITS 3.6.4.1 requires the secondary containment to be OPERABLE in MODES 1, 2, and 3. Monticello Page 2 of 7 Attachment 1, Volume 11, Rev. 0, Page 300 of 431
Attachment 1, Volume 11, Rev. 0, Page 301 of 431 DISCUSSION OF CHANGES ITS 3.6.4.1, SECONDARY CONTAINMENT This changes the CTS by requiring the secondary containment to be OPERABLE in MODE 2 when the reactor water temperature is less than or equal to 21 20F. The purpose of CTS 3.7.C.2, in part, is to ensure the secondary containment is OPERABLE to mitigate the consequences of a loss of coolant accident (LOCA). Secondary Containment is required to be OPERABLE in MODES 1, 2, and 3 when a design basis LOCA could cause a release of radioactive material. In MODES I and 3, the reactor coolant temperature will always be above 212 0F. In MODE 2, the reactor coolant temperature may be less than or equal to 2120 F when the reactor is subcritical but control rods are being withdrawn. Therefore, it is necessary and acceptable to require the secondary containment to be OPERABLE. This change is designated as more restrictive because the LCO will be applicable under more reactor operating conditions than in the CTS. M.2 CTS 4.7.C does not contain an explicit verification that secondary containment vacuum is maintained. ITS SR 3.6.4.1.1 requires verification that secondary containment vacuum is > 0.25 inch of vacuum water gauge. This changes the CTS by adding a Surveillance to verify secondary containment vacuum within limits. The purpose of ITS SR 3.6.4.1.1 is to ensure the secondary containment is at a condition that is assumed in the accident analysis. This change is acceptable because it provides additional assurance that the status of the secondary containment is consistent with the safety analysis. This change is designated as more restrictive because it adds a Surveillance Requirement to the CTS. M.3 CTS 4.7.C.1 .a requires the secondary containment capability test to be performed with a standby gas treatment (SGT) filter train every 24 months. ITS SR 3.6.4.1.4 requires this same test, however it is required to be performed every 24 months "on a STAGGERED TEST BASIS for each SGT subsystem." This changes the CTS by requiring the test to be performed using each SGT subsystem at least once per 48 months. The purpose of CTS 4.7.C.1.a is to ensure the Integrity of the secondary containment boundary. The change is acceptable since the proposed Surveillance Frequency will continue to require performance of the test every 24 months. The secondary purpose of this test will now also serve to ensure that each SGT subsystem is being tested every 48 months (based on the definition of STAGGERED TEST BASIS). This change is designated as more restrictive since the ITS will require the test to be performed with a different SGT System train each Surveillance interval. M.4 CTS 4.7.C.1.a requires the secondary containment capability test to be performed; however the test does not include a test duration. ITS SR 3.6.4.1.4 requires this same test, however it must now be performed for a "1 hour" period. This changes the CTS by requiring the secondary-containment capability test to be performed for a 1 hour period. The purpose of CTS 4.7.C.1.a is to ensure the integrity of the secondary containment boundary. The change is acceptable since the proposed test period Monticello Page 3 of 7 Attachment 1, Volume 11, Rev. 0, Page 301 of 431
Attachment 1, Volume 11, Rev. 0, Page 302 of 431 DISCUSSION OF CHANGES ITS 3.6.4.1, SECONDARY CONTAINMENT will help to ensure the secondary containment can maintain the required vacuum over a reasonable period of time. This change is designated as more restrictive since the ITS includes a test duration period of "1hour" while the CTS does not. M.5 CTS 3.7.C.4.a states if CTS 3.7.C.1 through CTS 3.7.C.3 cannot be met during Run, Startup, or Hot Shutdown, to initiate a normal orderly shutdown and have the reactor in Cold Shutdown condition within 36 hours. ITS 3.6.4.1 ACTION B requires the unit to be in MODE 3 in 12 hours and MODE 4 in 36 hours under the same conditions. This changes the CTS by requiring the unit be in an intermediate condition (MODE 3) within 12 hours. The purpose of the 3.7.C.4.a is to place the unit in a MODE in which the LCO is no longer applicable in a reasonable amount of time. This change will now require the unit to be at an intermediate condition sooner than is currently required. The proposed Completion Time of 12 hours to be in MODE 3 is acceptable because it is consistent with the time proposed in other Specifications that requires the unit to be in MODE 3. This change is designated as more restrictive since in the ITS the unit will be required to be in an intermediate condition sooner than is currently required by the CTS when the secondary containment is inoperable. M.6 CTS 1.0.W.1 states that the Secondary Containment Integrity includes the condition that at least one door in each access opening is closed. However, CTS 4.7.C does not contain an explicit verification of the status of the access openings. ITS SR 3.6.4.1.3 requires the verification, every 31 days, that at least one door in each access opening is closed. This changes the CTS by adding a periodic Surveillance Requirement to the CTS to confirm the condition of the access openings. The purpose of CTS 1.0.W is to state the conditions of OPERABILITY for the secondary containment. The CTS 1.0.W.1 requirement that at least one door in each secondary containment access opening is closed has been included in a Surveillance Requirement associated with the Secondary Containment Specification (ITS 3.6.4.1). The verification is required every 31 days. This change is acceptable because it continues to periodically ensure that at least one secondary containment access opening is closed. This change is designated as more restrictive since a new Surveillance Requirement has been added. M.7 CTS 1.0.W states, in part, that Secondary Containment Integrity means that the reactor building is closed. However, CTS 4.7.C does not contain an explicit verification of the status of the secondary containment equipment hatches. ITS SR 3.6.4.1.2 requires verification, every 31 days, that all secondary containment equipment hatches are closed and sealed. This changes the CTS by adding the requirement that each secondary containment equipment hatch is "sealed" and it also adds a new Surveillance Requirement. The purpose of CTS 1.0.W is to state the conditions of OPERABILITY for the secondary containment. The CTS 1.0.W definition of Secondary Containment Integrity includes the requirement that the reactor building is closed. CTS 1.0.W.1 covers the requirements for access openings while CTS I .O.W.3 covers secondary containment automatic isolation penetrations. The secondary Monticello Page 4 of 7 Attachment 1, Volume 11, Rev. 0, Page 302 of 431
Attachment 1, Volume 11, Rev. 0, Page 303 of 431 DISCUSSION OF CHANGES ITS 3.6.4.1, SECONDARY CONTAINMENT containment includes two other types of penetrations that have the potential of being opened. One of these types of penetrations includes hatches and the other type involves manual containment isolation penetrations (manual valves or blind flanges). The addition of the secondary containment manual penetrations to the Technical Specification has been added in accordance with the Discussion of Changes in ITS 3.6.4.2. The addition of the requirement to ensure the secondary containment hatches are "closed and sealed" every 31 days is acceptable because it provides additional assurance that the status of the secondary containment is consistent with the safety analyses. This change is designated as more restrictive, because it adds Surveillance Requirements to the CTS. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA.1 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 4.7.C.1.a requires the secondary containment capability test to maintain at least 1/4 inch of water vacuum "under calm wind (u < 5 mph) conditions." CTS 4.7.C.1.a also states "If calm wind conditions do not exist during this testing, the test data is to be corrected to calm wind conditions." ITS SR 3.6.4.1.4 includes the same test, however the wind conditions are not specified. This changes CTS by moving the details of the wind conditions to the ITS Bases. The removal of these details for performing Surveillance Requirements from the Technical Specifications isacceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement to verify the secondary containment capability to maintain the specified vacuum conditions. Also, this change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5 of the ITS. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES L.1 (Category I - Relaxation of LCO Requirements) CTS 3.7.C.2.c requires the secondary containment to be OPERABLE when the fuel cask is being moved within the reactor building and CTS 3.7.C.4.b.3 provides actions when this is not met. ITS 3.6.4.1 does not include this requirement. This changes the CTS by deleting the requirement to maintain the secondary containment OPERABLE when the fuel cask is being moved within the reactor building. Monticello Page 5 of 7 Attachment 1, Volume 11, Rev. 0, Page 303 of 431
Attachment 1, Volume 11, Rev. 0, Page 304 of 431 DISCUSSION OF CHANGES ITS 3.6.4.1, SECONDARY CONTAINMENT The purpose of the CTS 3.7.C.2.c is to ensure the secondary containment is OPERABLE when the fuel cask is being moved in the reactor building. This change is acceptable because the LCO requirements continue to ensure that the structures, systems, and components are maintained consistent with the safety analyses. This change is also acceptable because the USAR contains restrictions on the movement of heavy loads based on the heavy loads analysis. The bounding design basis fuel handling accident assumes an irradiated fuel assembly is dropped onto an array of irradiated fuel assemblies seated within the reactor pressure vessel. The movement of other loads is administratively controlled based on available analysis for the individual load. USAR Section 12.2.5 includes a description of the manner in which Monticello controls heavy loads. This USAR Section provides a description of safe load paths and requires the establishment of load handling procedures in accordance with NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants." The removal of fuel cask controls from the Technical Specifications has been approved by the NRC in the FitzPatrick ITS. Therefore, these activities are not restricted in the Technical Specifications. This change is designated as less restrictive because less stringent LCO requirements are being applied in the ITS than were applied in the CTS. L.2 (Category 4-Relaxation of Required Action) CTS 3.7.C.4 does not provide any explicit time to restore the secondary containment to OPERABLE status when it is found inoperable prior to requiring a unit shutdown. Under similar conditions, ITS 3.6.4.1 ACTION A provides 4 hours to restore the secondary containment to OPERABLE status in MODE 1, 2, and 3 prior to requiring a unit shutdown. This changes the CTS by providing an explicit ACTION to allow time to restore an inoperable secondary containment to OPERABLE status prior to requiring a unit shutdown. The purpose of CTS 3.7.C.4 is to provide appropriate compensatory actions for an inoperable secondary containment. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing time to repair inoperable features. The Required Actions are consistent with safe operation under the specified Condition, considering the operability status of the redundant systems of required features, the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of a DBA occurring during the repair period. This change provides an ACTION that allows 4 hours to restore an inoperable secondary containment to OPERABLE status in MODES 1, 2, and 3. The 4 hour Completion Time provides a period of time to correct the problem that is commensurate with the importance of maintaining secondary containment during MODES 1, 2, and 3. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. L.3 (Category 7 - Relaxation Of Surveillance Frequency, Non-24 Month Type Change) CTS 4.7.C.1 .a requires the secondary containment capability test to be performed "prior to refueling." ITS SR 3.6.4.1.4 includes a similar test, but does not include the mode restrictions for performing the required test. This changes Monticello Page 6 of 7 Attachment 1, Volume 11, Rev. 0, Page 304 of 431
Attachment 1, Volume 11, Rev. 0, Page 305 of 431 i DISCUSSION OF CHANGES ITS 3.6.4.1, SECONDARY CONTAINMENT the CTS by deleting the requirement to perform the Surveillance "prior to refueling." The purpose of the 4.7.C.1.a is to ensure the integrity of the secondary containment boundary. This change is acceptable because the new Surveillance Frequency has been evaluated to ensure that it provides an acceptable level of equipment reliability. The proposed Surveillance does not include the restriction on unit conditions ("prior to refueling"). The Surveillance can be adequately tested in the operating conditions without jeopardizing safe plant operations. Operating experience shows that this secondary containment capability test routinely passes when performed. Therefore, there is no need to perform the test just "prior to refueling" to ensure it can satisfy the test criteria for refueling activities. This change is designated as less restrictive because the Surveillance may be performed at plant conditions other than "prior to refueling." Monticello Page 7 of 7 Attachment 1, Volume 11, Rev. 0, Page 305 of 431
Attachment 1, Volume 11, Rev. 0, Page 306 of 431 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 306 of 431
Attachment 1, Volume 11, Rev. 0, Page 307 of 431 RSecondarA Containment 3.6.4.1 0D \K- CTS 3.6 CONTAINMENT SYSTEMS 3.6.4.1 [SecondaryM Containment 0 3.7.C. 3.7.C.2 LCO 3.6.4.1 The TsecondaryM containment shall be OPERABLE. (0 3.7.C.2 APPLICABILITY: MODES 1, 2, and 3, During movement ofjrecentlyJ irradiated fuel assemblies in the Psecondaryl containment, During operations with a potential for draining the reactor vessel
} 0D (OPDRVs).
ACTIONS CONDITION REQUIRED ACTION I COMPLETION TIME DOC L2 A. RSecondar4 containment A.1 Restore lsecondaryM 4 hours 0) inoperable in MODE 1, containment to OPERABLE 2, or 3. status. I4 3.7.CA.a B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not AND met. B.2 Be in MODE 4. 36 hours 3.7.C.4.b C. jSecondarA containment C.1 ----- NOTE inoperable during LCO 3.0.3 is not applicable. movement of IrecentlA irradiated fuel assemblies in the Suspend movement of Immediately 0D RsecondarA containment Irecentl9 irradiated fuel or during OPDRVs. assemblies in the
.econdarA containment.
AND C.2 Initiate action to suspend Immediately OPDRVs. BWR/4 STS 3.6.4.1-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 307 of 431
Attachment 1, Volume 11, Rev. 0, Page 308 of 431 Pecondaryf Containment 3.6.4.1 0 CTS SURVEILLANCE REQUIREMENTS I1 SURVEILLANCE FREQUENCY
+
DOCM.2 SR 3.6.4.1.1 0 VerifyosecondaryT containment vacuum is 24 hours ] 0 2j0.251 inch of vacuum water gauge. I2 1.O.W SR 3.6.4.1.2 Verify allisecondaryA containment equipment hatches are closed and sealed. 31 days 0
-r 1.o.w.0 SR 3.6.4.1.3 Verify one Rsecondary containment access door in each access opening is closed.
31 days 0 SR 3.6.4.1.4 [ Verify [seco containment can be drawn down months on a to [0. ch of vacuum water gauge in STAGGERED
< seconds using one standby gas trea t TEST BASIS for 0 SGT) subsystem. each subsystem]
. _
4.7.C.1.a SR 3.6.4.1 T- Verify thelsecondarA containment can be months on a 0D STAGGERED maintained 2 0.25Minch of vacuum water gauge for 1 hour using one SGT subsystem at a flow rate TEST BASIS for 0 5 M400CO cfm. each SGT subsystem BWR/4 STS 3.6.4.1-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 308 of 431
Attachment 1, Volume II, Rev. 0, Page 309 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.4.1, SECONDARY CONTAINMENT
- 1. The brackets have been removed and the proper plant specific information/value has been provided.
- 2. The bracketed Surveillance Requirement (ISTS SR 3.6.4.1.4), the drawdown test, has been deleted consistent with the current licensing basis. The Monticello safety analysis does not assume an explicit drawdown time, since it is assumed that the secondary containment is operating within the prescribed secondary containment vacuum limits of ITS SR 3.6.4.1.1 (L0.25 inch of vacuum water gauge). The subsequent SR has been renumbered to reflect the deletion.
Monticello Page 1 of I Attachment 1, Volume 11, Rev. 0, Page 309 of 431
Attachment 1, Volume 11, Rev. 0, Page 310 of 431 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 310 of 431
Attachment 1, Volume 11, Rev. 0, Page 311 of 431 [SecondaryM Containment B 3.6.4.1 0 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.1 ESecondaryl Containment 0 BASES BACKGROUND The function of the secondaryM containment is to contain, dilute, and hold up fission products that may leak from primary containment following a 0 Design Basis Accident (DBA). In conjunction with operation of the Standby Gas Treatment (SGT) System and closure of certain valves whose lines penetrate the RsecondaryM containment, the Msecondaryl containment is designed to reduce the activity level of the fission products 0 prior to release to the environment and to isolate and contain fission products that are released during certain operations that take place inside primary containment, when primary containment is not required to be OPERABLE, or that take place outside primary containment. The MsecondarA containment is a structure that completely encloses the primary containmentland those crrpannaiiway hp-Icontain priErsystem fluid. This structure forms a control volume that serves to hold up and dilute the fission products. It is possible for the CO pressure in the control volume to rise relative to the environmental pressure (e.g., due to pump and motor heat load additions). To prevent ground level exfiltration while allowing theasecondaryM containment to be designed as a conventional structure, the~secondaryA containment J requires support systems to maintain the control volume pressure at less than the external pressure. Requirements for these systems are specified separately in LCO 3.6.4.2, "Secondary Containment Isolation Valves (SCIVs)," and LCO 3.6.4.3, "Standby Gas Treatment (SGT) System." APPLICABLE SAFETY There are two principal accidents for which credit is taken for PecondaryM containment OPERABILITY. These are a loss of coolant accident 0D ANALYSES (LOCA) (Ref. 1)and a fuel handling accidentainvolving handling recently irradiated fuel i.e., fuel that has occupied part of a critical reactor core 0 hoswithin the previous a insidelsecondarA containment (Ref. 2). 0 The gsecondarl containment performs no active function in response to each of these limiting events; however, its leak tightness is required to ensure that the release of radioactive materials from the primary containment is restricted to those leakage paths and associated leakage rates assumed in the accident analysis and that fission products entrapped within the IsecondarA containment structure will be treated by the SGT System prior to discharge to the environment. 0D JSecondarh containment satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). 0 BWR/4 STS B 3.6.4.1-1 Rev. 3.0, 03/31/04 Attachment 1,Volume 11, Rev. 0, Page 311 of 431
Attachment 1, Volume 11, Rev. 0, Page 312 of 431 secondaryM Containment B 3.6.4.1 0D BASES LCO An OPERABLElsecondary4 containment provides a control volume into which fission products that bypass or leak from primary containment, or 0D are released from the reactor coolant pressure boundary components located inIsecondarAj containment, can be diluted and processed prior to release to the environment. For thelsecondaryg containment to be considered OPERABLE, it must have adequate leak tightness to ensure 0D that the required vacuum can be established and maintained; APPLICABILITY In MODES 1, 2, and 3, a LOCA could lead to a fission product release to primary containment that leaks to secondarj containment. Therefore, RsecondarA containment OPERABILITY is required during the same 0 operating conditions that require primary containment OPERABILITY. In MODES 4 and 5, the probability and consequences of the LOCA are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining IsecondarfA containment OPERABLE is 0 not required in MODE 4 or 5 to ensure a control volume, except for other situations for which significant releases of radioactive material can be postulated, such as during operations with a potential for draining the reactor vessel (OPDRVs) or during movement of Irecentlyj irradiated fuel assemblies in the Tsecondary containment. IDue to radioactive decay, secondary containment is only required to be OPERABLE during fuel (i4 l handling involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous [Xkday) J
------ REV WER'S NOTE------------i--
The addition of the term "rece tly" associated with handling irra iated fuel in all of the containment fun ion Technical Specification requi ements is only applicable to those lic sees who have demonstrated b analysis that after sufficient radioa ive decay has occurred, off-site oses resulting from a fuel han ling accident remain below the S andard Review Plan limits (well within CFR 100). Additionally, license s adding the term "recently" mus make the following 0 commitment which s consistent with NUMARC 93-0, Revision 4, Section 11.3.6.5,' afety Assessment for Removal f Equipment from Service During utdown Conditions," subheadin "Containment - Secondary (B )."
"The followi guidelines are included in the asessment of systems removed fr m service during movement of irr diated fuel:
BWR/4 STS B 3.6.4.1-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 312 of 431
Attachment 1, Volume 11, Rev. 0, Page 313 of 431 TSecondarA Containment B 3.6.4.1 0 BASES APPLICABILITY (continued)
-During fuel handling/core alter tions, ventilation system and rad tion monitor availability (as define in NUMARC 91-06) should be a sessed, with respect to filtration and onitoring of releases from the fu I.
Following shutdown, radioa tivity in the fuel decays away fair rapidly. The basis of the Technical pecification operability amend ntis the reduction in doses due to/uch decay. The goal of maintai ing ventilation system and radiation m itor availability is to reduce dos even further below that provided by he natural decay. 0
-A single normal or c ntingency method to promptly cIose primary or secondary contain ent penetrations should be devel ped. Such prompt methods need not ompletely block the penetration r be capable of resisting pressur/
The purpose of/he "prompt methods" mentione above are to enable ventilation sy ems to draw the release from a stulated fuel handling accident in t e proper direction such that it ca be treated and monitored.; ACTIONS A.1 If [secondaryM containment is inoperable, it must be restored to OPERABLE status within 4 hours. The 4 hour Completion Time provides 0 a period of time to correct the problem that is commensurate with the importance of maintainingasecondari containment during MODES 1, 2, and 3. This time period also ensures that the probability of an accident 0 (requiring secondaryJ containment OPERABILITY) occurring during periods where Tsecondarl containment is inoperable is minimal. J 0D B.1 and B.2 If [secondary containment cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a 0 MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. BWR/4 STS B 3.6.4.1-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 313 of 431
Attachment 1, Volume 11, Rev. 0, Page 314 of 431 EsecondarA Containment (E) B 3.6.4.1 BASES ACTIONS (continued) C.1 and C.2 Movement oflrecentlA irradiated fuel assemblies in thelsecondarA] containment and OPDRVs can be postulated to cause significant fission product release to thelsecondarg] containment. In such cases, the RsecondarAJ containment is the only barrier to release of fission products L to the environment. Therefore, movement offgrecentlA irradiated fuel r assemblies must be immediately suspended if thelsecondarA 9 containment is inoperable. Suspension of these activities shall not preclude completing an action that involves moving a component to a safe position. Also, action must be immediately initiated to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended. Required Action C.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If movingjrecentlij irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving arecentlAj irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel D movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of SIrecentl9j irradiated fuel (iD assemblies would not be a sufficient reason to require a reactor shutdown. SURVEILLANCE SR 3.6.4.1.1 0 REQUIREMENTS This SR ensures that the Csecondary containment boundary is sufficiently ( leak tight to preclude exfiltration under expected wind conditions. The 24 hour Frequency of this SR was developed based on operating experience related tolsecondarA containment vacuum variations during (i the applicable MODES and the low probability of a DBA occurring I betweeeveillanceR. Furthermore, the 24 hour Frequency is considered adequate in view of other indications available in the control room, including alarms, to alert the operator to an abnormal secondary containment vacuum condition.I (i BWRI4 STS B 3.6.4.1-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 314 of 431
Attachment 1, Volume 11, Rev. 0, Page 315 of 431 [Secondary] Containment B 3.6.4.1 BASES SURVEILLANCE REQUIREMENTS (continued) 0 SR 3.6.4.1.2 and SR 3.6.4.1.3 Verifying thatisecondarAj containment equipment hatches and one A) access door in each access opening are closed ensures that the infiltration of outside air of such a magnitude as to prevent maintaining the desired negative pressure does not occur. Verifying that all such openings are closed provides adequate assurance that exfiltration from thelsecondarA containment will not occur. In this application, the term "sealed" has no connotation of leak tightness. Maintaining Msecondary (J containment OPERABILITY requires verifying one door in the access opening is closed. RlAn access opening contains one inner and one outer door. In some cases, secondary containment access openings are shared such that alsecondar] containment barrier may have multiple (D inner or multiple outer doors. The intent is to not breach the secondary containment at any time whenasecondarA containment is required. This is achieved by maintaining the inner or outer portion of the barrier closed at all timesg However, all OsecondaryA containment access doors are normally kept closed, except when the access opening is being used for entry and exit or when maintenance is being performed on an access opening. The 31 day Frequency for these SRs has been shown to be adequate, based on operating experience, and is considered adequate in view of the other indications of door and hatch status that are available to the operator. ISR 3.16 .4and SR 3.6.4.1 1i) The SGT System exhausts thelsecondarA containment atmosphere to the environment through appropriate treatment equipmen subsys em is designed to draw down presure in the secondary conta' ment to 2[0.25] inches of vacuu water gauge in ! [1 20 seconds KY and aintain pressure in the [seconda ] containment at a [0. 6 inches of v cuum water gauge for 1 hour at flow rate s [4000] cfm. To ensure that all fission products released to the secondarA containment are 4 treated, [SR 3. . .4 and] SR 3.6.4.1 ri at a ressur ith 4 RsecondarA containment that is less than the owest postulated pressure external to the IsecondarA containment boundary can ra dl be 0 dmaintained. When the SGT System is operating as esigne , tees a ent an maintenance of IsecondarA containment pressure cannot be accomplished if thelsecondarMJ containment 0 boundary is not intact. s a ishment oure is confirme BWRI4 STS B 3.6.4.1-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 315 of 431
Attachment 1, Volume II, Rev. 0, Page 316 of 431 IsecondarA Containment B 3.6.4.1 03 BASES SURVEILLANCE REQUIREMENTS (continued) SR 3.%.4.1.4, which demonstrates that t [secondary] contain nt can be dr wn down to 2 [0.25] inches of va um water gaue in {2 2sec nds using one SGT subsystem.1 SR 3.6.4.1 emonstrates a e pressure in theWsecondary' containment can be maintained 2 020 Thentest isnormaly inches of vacuum water gauge for 1 hour using one SGT subsystem at a performed under calm wind (-c5 flow rate s4400CU cfm. The 1 hour test period allows[secondaryl ___________ ______ () mph) conditions. If s containment to be in thermal equilibrium at steady state conditions. The calm wind nt primary purpose of to~a SFNjf is to ensure secondar] containment (D4 existduringthis boundary integritv4The secondary purpose of tfo,$d SlfO is to ensure testing, the test that the SGT subsystem being tested functions as designed. There is a data Isto be cofrectedtocalm wind conditions, separate LCO with Surveillance Requirements which serves the primary purpose of ensuring OPERABILITY of the SGT System. T [ 0 (i) need not be performed with each SGT subsystem. The SGT subsystem used for to Surveillancep is staggered to ensure that in addition to E the requirements of LCO 3.6.4.3, either SGT subsystem will perform this 0 test. The inoperability of the SGT System does not necessarily constitute a failure of trjal SurveillancEM relative to thelsecondarA¶ containment 4) [T1 OPERABILITY. Operating experience has shown theasecondari (Di c con ent boundary usually passes Surveillance~when (i) performed at th 8 month Frequency. Therefore, the Frequency was ) concluded to be acceptable from a reliability standpoint. REFERENCES y SAR, Section15 RE(Ei) TSAR, Sectiotion a0 BWR/4 STS B 3.6.4.1-6 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 316 of 431
Attachment 1, Volume 11, Rev. 0, Page 317 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.4.1 BASES, SECONDARY CONTAINMENT
- 1. The brackets have been removed and the proper plant specific information/value has been provided.
- 2. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the plant specific nomenclature, number, reference, system description, analysis description, or licensing basis description.
- 3. The Reviewers Note is deleted as it is not part of the plant-specific ITS.
- 4. Changes have been made to reflect those changes made to the Specification.
Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 317 of 431
Attachment 1, Volume 11, Rev. 0, Page 318 of 431 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 11, Rev. 0, Page 318 of 431
Attachment 1, Volume 11, Rev. 0, Page 319 of 431 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.6.4.1, SECONDARY CONTAINMENT There are no specific NSHC discussions for this Specification. Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 319 of 431
Attachment 1, Volume 11, Rev. 0, Page 320 of 431 ATTACHMENT 14 ITS 3.6.4.2, Secondary Containment Isolation Valves (SCIVs) Attachment 1,Volume 11, Rev. 0, Page 320 of 431
Attachment 1, Volume 11, Rev. 0, Page 321 of 431 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 11, Rev. 0, Page 321 of 431
C C.' C ITS 3.6.4.2
\
ITS ITS 3.0 UMITNG CONDTONoS FOR OPERATION 4.0 SURVEILLANCE REQUiREMENTS C. Secor C o_ ___. P___X C. SeconayContalyinetlorVavs U. Secondary Containment t> lso2oVa!esFA.2J A) 0 ICO 3.6.4.2 1. Except as specified in 3.7.C.2 andf3.7.C.3. SecondaryContainme entnha1 be 0 O B a urinr all modes of plant opraio a P 3 LCO 3.6.4.2 2. Seodotanmn sotreu -.
- b~en all of the tollowina conditions are satisfied 0 Applcabiity - E The readtorIs subcritlcai and Spedificatlon - 0 Applicability 3.3A Is met.
3 b. The reactor water temperature is below 21 2°F 3 0 0 - X& 0 0 CD c CD
-A to 0 0 0)
CD CA)
-9' CD) 0
-tk CA) Page 1 of4
Attachment 1, Volume 11, Rev. 0, Page 323 of 431 0 ITS 3.6.4.2 ITS a INSERT A
- d. Recently irradiated fuel is not being moved within secondary Applicability containment.
- e. Operations with the potential for draining the reactor vessel are not being performed.
(i) INSERT B 4.a During Run, Startup, or Hot Shutdown if through 3.7.C.3 cannot be the Specifications 3.7.C.1 orderly ACTION C_ met, initiate a normal shutdown and have reactor n the Cold Shutdown condition within And 4.b If Specifications 3.7.C.1 through 3.7.C.3 cannot be met, immediately suspend: Add proposed Required Action D.1 Note I ) ACTION D - 1. Operations with a potential for draining the reactor vessel.
- 2. Handling of recently irradiated fuel in the secondary containment.
- 13. Maveent&f fuel cask in the-adinogr buildini.H- Q Insert Page 169 Page 2 of 4 Attachment 1, Volume 11, Rev. 0, Page 323 of 431
C C C ITS 3.6.4.2 J.O UMITING CONiDmONS FOR OPERATION 4.0 SURVEILLANCE REOUIREMENTS I reactor cores rations with a potenfial for reducing Su
- 0) the shutdown argln below that spqcified In C) speciication ~.3.A and handling o adiated fuel or 0)
Ihe fuel cesgi n the secondary cor-fainment are to ---- O (D be Immadately suspended it sec~ndary a 0 contsinm ht intesrity Isnct melrialned. 0 2 D. Primary Containment Isolation Valves (PCIVs) D. Primary Contalinment Isolation Vales (PCIVs) a 0
- 1. During reactor power operating conditions, all 1. The primary containment automatic Isolation valve 0 Primary Containment automatIc Isolation valves and surveillance shall be performed as follows:
all primary system Instniment lne flow check valves 2 0 shag be operable except as specified In 3.7.D.2 and a. At least once per operating cycle the operable 3 D C I 3.7.D.3. Isolation valves that are power operated and automatically Initiated shal be tested for 4 simulated automatic Initiation and closure times. -0
;U (a
0 b. At least once per operating cycle the primary (D tD syslem Instrument One flow check valves shall be tested for proper operation. -A) -h 0 0 c. Al normally open power-operated isolation valves shal be tested In accordance with the See ITS 3.6.1.3 } 0 Inservice Testing Program. Main Steam ID Isolation valves shall be tested (one at a time) 0 with the reactor power less than 75% of rated. CD) 0
- d. Al least once per week the main steam-line h power-operated Isolation valves shall be -L exercised by partial closure and subsequent reopening.
3.7/4.7 170 01/28/05 Amendment No. 3 71, 77, 122. 130,141 Page 3 of 4
c C ITS 3.6.4.2 ITS 0-
- 4. Protective Function -A system protective action which results from the protectve action of the channel monitoring a particular plant condition.
W C, S R Rated Neutron Flux- Rated flux Is the neutron flux that corresponds to a steady-state power level of 1775 thermal megawatts. 0 0 S. Rated Thermal Power - Rated thermal power means a steady-state power level of 1775 thermal megawatts. rD T. Reactor Coolant System Pressure or Reactor Vessel Pressure - Unless otherwise Indicated, reactor vessel pressures listed hI See ITS Chapter 1.01 0 the Technical Specifications are those existing In the vessel steam space. 0 U. Refueling Ooeration and Refueling Outage - Refueling Operation Is any operation when the reactor water temperature Is less 0 Co 0 than 21 2 F and movement of fuel or core components Is In progress. For the purpose of designating frequency of testing and -U 3 survelilance, a refueling outage shall mean a regularly scheduled refueling outage; however, where such outages occur within 8 o months of the completion of the previous refueling outage, the required surveillance testing need not be performed until the next Ct regularly scheduled outage. 0 0
-4
-n V. Snaet*LlmR - The safety imits are lmits below which the maintenance of the cladding and primary system integrty are assured. cD~ (D r Exceeding such a imit Is cause for plant shutdown and review by the Commission before resuenption of plant operation. Operation beyond such a limit may not In Itself result In serious consequences but I Indicates an operational deficiency subject fn ran..iltnno revaiew te reacor Wiin i 0~ LQn Inte meanst -U 0 con s
-
I . . . 7 .-9 ._ fl,_I' 'IS A Ii1 CD l 1. At least one ooor in eacn access opening Is closealj . *- ". 1 co en
- 2. The standb as treatment te is rable..3 K)
CO 3.6.42 CYI ACTION A Ar are secured In the clo - Ai) X. SensorCheck A qualitative determination of operability by observation of sensor behavior during operation. ims-eTermin aion l shall Include, where possible, comparison with other Independent sensors measuring the same variable. See ITS Chapter 1.0 1.0 4 9/16/98 Amendment No. 47,102 Page 4 of 4
Attachment 1, Volume 11, Rev. 0, Page 326 of 431 DISCUSSION OF CHANGES ITS 3.6.4.2, SECONDARY CONTAINMENT ISOLATION VALVES (SCIVs) ADMINISTRATIVE CHANGES A.1 In the conversion of the Monticello Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1 433, Rev. 3, "Standard Technical Specifications General Electric Plants, BWR/4" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 CTS 3.7.C.1 requires the Secondary Containment Integrity to be maintained and CTS 3.7.C.2 uses the term Secondary Containment Integrity. CTS 1.0.W, the Secondary Containment Integrity definition, in part, states that the reactor building is closed. This definition is interpreted to mean that all secondary containment penetrations are closed (i.e., penetrations including manual valves and are required to be closed during accident conditions). CTS 1.0.W.3 also requires all reactor building ventilation system automatic valves to be OPERABLE. ITS LCO 3.6.4.2 requires the Secondary Containment Isolation Valves (SCIVs) to be OPERABLE. This changes the CTS by including the requirements for SCIVs (i.e., manual valves, blind flanges, and reactor building automatic valves) in a separate Specification. The purpose of CTS 3.7.C.1 is to provide requirements pertaining to secondary containment OPERABILITY, including the valves that isolate secondary containment penetration flow paths. ITS LCO 3.6.4.2 requires each SCIV to be OPERABLE. This change is acceptable since the definition of OPERABLE and the subsequent ITS 3.6.4.2 LCO, ACTIONS, and Surveillance Requirements are sufficient to encompass the applicable requirements of the CTS definition, with respect to isolation valves. This change is designated as administrative because it does not result in technical changes to the CTS. A.3 CTS 3.7.C.1 specifies requirements for the secondary containment during "all modes of plant operation." However, CTS 3.7.C.2 states that secondary containment is not required "when all of the listed conditions are satisfied," and provides a list of six conditions (CTS 3.7.C.2.a through ). ITS LCO 3.6.4.2 specifies requirements for the secondary containment isolation valves in the positive sense (when the secondary containment isolation valves are required to be OPERABLE). This changes the CTS by specifying the requirements for the secondary containment isolation valves when they are required to be OPERABLE instead of when they are not required to be OPERABLE. The purpose of CTS 3.7.C.1 and CTS 3.7.C.2 is to state when the secondary containment isolation valve requirements are applicable. CTS 3.7.C.1 specifies requirements for the secondary containment isolation valves during "all modes of plant operation." However, CTS 3.7.C.2 states that secondary containment isolation valves are not required "when all of the listed conditions are satisfied." ITS LCO 3.6.4.2 specifies requirements for the secondary containment isolation valves in the positive sense (when the SCIVs are required to be OPERABLE). This change is acceptable because it clearly defines the conditions for when the secondary containment isolation valve requirements must be met and is Monticello Page 1 of 9 Attachment 1, Volume 11, Rev. 0, Page 326 of 431
Attachment 1, Volume 11, Rev. 0, Page 327 of 431 DISCUSSION OF CHANGES ITS 3.6.4.2, SECONDARY CONTAINMENT ISOLATION VALVES (SCIVs) consistent with the format specified for other systems and parameters. This change is designated as administrative since it does not result in a technical change to the CTS. A.4 CTS 3.7.C.3 provides requirements to be taken for one or more penetration flow paths with a SCIV inoperable. ITS 3.6.4.2 includes an explicit Note (ACTIONS Note 2) that provides instructions for the proper application of the ACTIONS for ITS compliance (i.e., Separate Condition entry is allowed for each penetration flow path). This changes the CTS by providing explicit direction as to how to utilize the ACTIONS when a SCIV is inoperable. This change is acceptable because the addition of the Note reflects the CTS allowance to take the appropriate Actions on a per "duct" basis. This change is designated as administrative since it does not result in a technical change to the CTS. A.5 CTS 3.7.C.3 does not specifically require Conditions to be entered for systems supported by inoperable secondary containment isolation valves. OPERABILITY of supported systems is addressed through the definition of OPERABILITY for each system, and appropriate LCO Actions are taken. ITS 3.6.4.2 ACTIONS Note 3 states "Enter applicable Conditions and Required Actions for systems made inoperable by SCIVs." ITS LCO 3.0.6 provides an exception to ITS LCO 3.0.2, stating "When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered." This changes the CTS by adding a specific statement to require supported system Conditions and Required Actions be entered, whereas in the CTS this would be done without the Note. This change is acceptable because the addition of the ITS Note reflects the CTS requirement to take applicable Actions for inoperable systems. The ITS Note is required because of the addition of ITS LCO 3.0.6, and because the requirement to declare supported systems inoperable is being retained. This change is designated as administrative because it does not result in any technical changes to the CTS. A.6 CTS 3.7.C.3 requires restoring the inoperable damper to OPERABLE status or isolating the affected duct by use of a closed damper or blind flange within eight hours. ITS 3.6.4.2 ACTIONS do not include the specific option to restore the valves to OPERABLE status, but includes other compensatory Required Actions to take within 8 hours. This changes the CTS by not explicitly stating the requirement to restore an inoperable valve to OPERABLE status. The purpose of CTS 3.7.C.3 is to provide appropriate compensatory actions for inoperable SCIVs. This change is acceptable because the technical requirements have not changed. Restoration of compliance with the LCO is always an available Required Action and it is the convention in the ITS to not state such "restore" options explicitly unless it is the only action or is required for clarity. This change is designated as administrative because it does not result in any technical changes to the CTS. Monticello Page 2 of 9 Attachment 1, Volume 11, Rev. 0, Page 327 of 431
Attachment 1, Volume 11, Rev. 0, Page 328 of 431 DISCUSSION OF CHANGES ITS 3.6.4.2, SECONDARY CONTAINMENT ISOLATION VALVES (SCIVs) A.7 CTS 3.7.C.3 states the actions that must be taken when the reactor building ventilation system automatic isolation dampers (valves) are not OPERABLE and requires the valves to be isolated by a closed damper or blind flange. In addition, CTS I.O.W.3 requires all reactor building ventilation system automatic isolation valves to be OPERABLE or "secured in the closed position." ITS 3.6.4.2 ACTION A covers inoperabilities associated with these penetrations and requires the affected penetration flow path to be isolated by use of at least one closed "and de-activated automatic valve," closed manual valve, or blind flange. This changes the CTS by incorporating the explicit CTS definition statement concerning the option to have the penetration "secured in the closed position" into ITS 3.6.4.2 ACTION A. The change that allows use of a manual valve is discussed in DOC L.4. The purpose of CTS 3.7.C.3 is to provide the appropriate compensatory actions for the reactor building ventilation system automatic valves. The requirements in CTS 3.7.C.3 include the requirements of CTS 1.0.W.3 to isolate the penetration when a reactor building ventilation system automatic valve is inoperable. The requirements of CTS 3.7.C.3 have been incorporated in ITS 3.6.4.2 ACTION A. Since the requirements prescribed in CTS 3.7.C.3 and are retained in ITS 3.6.4.3 ACTION A, the CTS definition for these requirements in the Secondary Containment Integrity definition is not needed. These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.8 CTS 4.7.C.1 .b.(1) requires verification that each automatic damper actuates to its isolation position "each refueling interval." ITS SR 3.6.4.2.3 requires a similar test every "24 months." This changes the CTS by changing the Frequency from "refueling interval" to "24 months." This change is acceptable because the current "refueling interval" is "24 months." In letter L-MT-04-036, from Thomas J. Palmisano (NMC) to the USNRC, dated June 30, 2004, NMC has proposed to extend the fuel cycle from 18 to 24 months and the same time has performed an evaluation in accordance with Generic Letter 91-04 to extend the unit Surveillance Requirements from 18 months to 24 months. CTS 4.7.C.1 .b.(1) was included in this evaluation. This change is designated as administrative because it does not result in any technical changes to the CTS. A.9 CTS 3.7.C.4.b.2 requires the unit to suspend handling of recently irradiated fuel. ITS 3.6.4.2 ACTION D includes the same requirement, however ITS 3.6.4.2 Required Action D.1 includes a Note that states that LCO 3.0.3 is not applicable. This changes the CTS by adding this Note. The purpose of CTS 3.7.C.4.b.2 is to provide the appropriate actions when the secondary containment is inoperable and the unit is moving recently irradiated fuel in the secondary containment. This change adds a Note that states LCO 3.0.3 is not applicable. This Note has been added because ITS LCO 3.0.3 has been added to ITS Section 3.0 in accordance with DOC M.1. This Note is necessary because if moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently Monticello Page 3 of 9 Attachment 1, Volume 11, Rev. 0, Page 328 of 431
Attachment 1, Volume 11, Rev. 0, Page 329 of 431 DISCUSSION OF CHANGES ITS 3.6.4.2, SECONDARY CONTAINMENT ISOLATION VALVES (SCIVs) irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of recently irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown. Since ITS LCO 3.0.3 is not currently included in the CTS this change is considered administrative. This change is designated as administrative because it does not result in technical changes to the CTS. A.10 These changes to CTS 3.7.C.2.c, CTS 3.7.C.2.d, and CTS 3.7.C.4, and the addition of CTS 3.7.C.2.d and e are provided in the Monticello ITS consistent with the Technical Specifications Change Request submitted to the USNRC for approval in NMC letter L-MT-05-013, from Thomas J. Palmisano (NMC) to USNRC, dated April 12, 2005. As such, these changes are administrative. A.11 This change to CTS 4.7.C.1.b.(1) is provided in the Monticello ITS consistent with the Technical Specifications Change Request submitted to the USNRC for approval in NMC letter L-MT-04-036, from Thomas J. Palmisano (NMC) to USNRC, dated June 30, 2004. As such, this change is administrative. MORE RESTRICTIVE CHANGES M.1 CTS 3.7.C.2.a and 3.7.C.2.b state that the secondary containment (i.e., SCIVs) requirements are not required when both the reactor is subcritical and Specification 3.3.A is met, and reactor water temperature is below 212 0F, respectively. ITS 3.6.4.2 requires the SCIVs to be OPERABLE in MODES 1, 2, and 3. This changes the CTS by requiring the SCIVs to be OPERABLE in MODE 2 when the reactor water temperature is less than or equal to 212 0F. The purpose of CTS 3.7.C.2, in part, is to ensure the SCIVs are OPERABLE to mitigate the consequences of a loss of coolant accident (LOCA). SCIVs are required to be OPERABLE in MODES 1, 2, and 3 when a design basis LOCA could cause a release of radioactive material. In MODES 1, and 3 the reactor coolant temperature will always be above 2120F. In MODE 2, the reactor coolant temperature may be less than or equal to 212 0F when the reactor is subcritical but control rods are being withdrawn. Therefore, it is necessary and acceptable to require the SCIVs to be OPERABLE. This change is designated as more restrictive because the LCO will be applicable under more reactor operating conditions than in the CTS. M.2 CTS 3.7.C.3 states that with an inoperable secondary containment isolation damper to isolate the affected duct by use of a closed damper or blind flange within eight hours. ITS 3.6.4.2 ACTION A includes a similar requirement, however it also includes an additional Required Action (ITS 3.6.4.2 Required Action A.2) to verify the affected penetration flow path is isolated once per 31 days. The Required Action also includes two Notes, one that states that isolation devices in high radiation areas may be verified by use of administrative means and a second that states that isolation devices that are locked, sealed, or otherwise secured may be verified by use of administrative means. This changes the CTS by adding this additional Required Action and associated Notes. Monticello Page 4 of 9 Attachment 1, Volume 11, Rev. 0, Page 329 of 431
Attachment 1, Volume 11, Rev. 0, Page 330 of 431 DISCUSSION OF CHANGES ITS 3.6.4.2, SECONDARY CONTAINMENT ISOLATION VALVES (SCIVs) The purpose of ITS 3.6.4.2 Required Action A.2 is to ensure that secondary containment penetrations required to be isolated following an accident, but no longer capable of being automatically isolated, will be in the isolation position should an event occur. The Completion Time of once per 31 days is appropriate because the valves are operated under administrative controls and the probability of their misalignment is low. This change is more restrictive since an additional verification will be required when operating within the Actions. M.3 CTS 1.0.W states, in part, that Secondary Containment Integrity means that the reactor building is closed, however CTS 4.7.C does not contain an explicit periodic verification of the status of the secondary containment isolation manual valves and blind flanges. ITS SR 3.6.4.2.1 requires a 31 day verification that each secondary containment isolation manual valve and blind flange that is not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed. In addition, this Surveillance is modified by two Notes that allow valves and blind flanges in high radiation areas to be verified by administrative means and allows SCIVs to be open under administrative controls. This changes the CTS by adding the Surveillance Requirement for verification of the status of the secondary containment isolation manual valves and blind flanges. The purpose of ITS SR 3.6.4.2.1 isto ensure that each secondary containment manual isolation valve and blind flange is closed. This change is acceptable because it provides additional assurance that the status of the secondary containment is consistent with the safety analysis. This change is designated as more restrictive because it adds a Surveillance Requirement to the CTS. M.4 CTS 4.7.C does not include any requirements to verify the isolation times of each power operated, automatic SCIV. ITS SR 3.6.4.2.2 requires this verification every 92 days. This changes the CTS by adding the additional Surveillance to verify the isolation time of automatic SCIVs is within limits. The purpose of ITS SR 3.6.4.2.2 is to provide additional assurance the SCIVs close within the required isolation times assumed in the safety analysis. This change is acceptable because it provides additional assurance that SCIVs will be capable of performing their function. This change is designated as more restrictive because it adds a Surveillance Requirement to the CTS. M.5 CTS 3.7.C.4.a states if CTS 3.7.C.1 through CTS 3.7.C.3 cannot be met during Run, Startup, or Hot Shutdown, to initiate a normal orderly shutdown and have the reactor in Cold Shutdown condition within 36 hours. ITS 3.6.4.2 ACTION C requires the unit to be in MODE 3 in 12 hours and MODE 4 in 36 hours under the same conditions. This changes the CTS by requiring the unit be in an intermediate condition (MODE 3) within 12 hours. The purpose of the 3.7.C.4.a is to place the unit in a MODE in which the LCO is no longer applicable in a reasonable amount of time. This change will now require the unit to be at an intermediate condition sooner than is currently required. The proposed Completion Time of 12 hours to be in MODE 3 is acceptable because it is consistent with the time proposed in other Specifications that requires the unit to be in MODE 3. This change is designated as more Monticello Page 5 of 9 Attachment 1, Volume 11, Rev. 0, Page 330 of 431
Attachment 1, Volume 11, Rev. 0, Page 331 of 431 DISCUSSION OF CHANGES ITS 3.6.4.2, SECONDARY CONTAINMENT ISOLATION VALVES (SCIVs) restrictive since in the ITS the unit will be required to be in an intermediate condition sooner than is currently required by the CTS when the a SCIV is inoperable and the affected penetration not isolated. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L.1 (Category I - Relaxation of LCO Requirements) CTS 3.7.C.2.c requires the SCIVs to be OPERABLE when the fuel cask is being moved within the reactor building and CTS 3.7.C.4.b.3 provides actions when this is not met. ITS LCO 3.6.4.2 does not include this requirement. This changes the CTS by deleting the requirement to maintain the SCIVs OPERABLE when the fuel cask is being moved within the reactor building. The purpose of the CTS 3.7.C.2.c is to ensure the SCIVs are OPERABLE when the fuel cask is being moved in the reactor building. This change is acceptable because the LCO requirements continue to ensure that the structures, systems, and components are maintained consistent with the safety analyses. This change is also acceptable because the USAR contains restrictions on the movement of heavy loads based on the heavy loads analysis. The bounding design basis fuel handling accident assumes an irradiated fuel assembly is dropped onto an array of irradiated fuel assemblies seated within the reactor pressure vessel. The movement of other loads is administratively controlled based on available analysis for the individual load. USAR Section 12.2.5 includes a description of the manner in which Monticello controls heavy loads. This USAR Section provides a description of safe load paths and requires the establishment of load handling procedures in accordance with NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants." The removal of fuel cask controls from the Technical Specifications has been approved by the NRC in the FitzPatrick ITS. Therefore, these activities are not restricted in the Technical Specifications. This change is designated as less restrictive because less stringent LCO requirements are being applied in the ITS than were applied in the CTS. Cr5.. L.2 (Category 4 - Relaxation of Required Action) CTS 3.7.C.3 requires the associated penetration flow path isolated when a secondary containment isolation valve is inoperable and not restored to OPERABLE status. However, this action does not include a provision that isolated valves closed to satisfy the requirements may be reopened on an intermittent basis under administrative controls. ITS 3.6.4.2 ACTIONS Note 1 allows any secondary containment penetration flow path, isolated due to an inoperable SCIV, to be unisolated Monticello Page 6 of 9 Attachment 1, Volume 11, Rev. 0, Page 331 of 431
Attachment 1, Volume 11, Rev. 0, Page 332 of 431 DISCUSSION OF CHANGES ITS 3.6.4.2, SECONDARY CONTAINMENT ISOLATION VALVES (SCIVs) intermittently under administrative controls. This changes the CTS by allowing the secondary containment penetrations to be opened under administrative controls when the associated penetration has been closed to satisfy the actions. The purpose of ITS 3.6.4.2 ACTIONS Note I is to allow the affected secondary containment penetration to be opened on an intermittent basis as required for evolutions such as performing Surveillances, repairs, and routine evolutions. This change is acceptable because the allowance requires administrative controls to be in place when the penetration is opened under administrative controls. The ITS 3.6.4.2 Bases states that these administrative controls consist of stationing a dedicated individual at the controls of the valve, who is in continuous communication with the control room. In this way, the penetration can be rapidly isolated when a need for secondary containment isolation is indicated. This allowance is also acceptable due to the low probability of an event that could pressurize the primary containment during the short time in which the SCIV is open and the administrative controls established to ensure the affected penetration can be isolated when a need for secondary containment isolation is indicated. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. L.3 (Category 3 - Relaxation of Completion Time) CTS 3.7.C.3 allows 8 hours to either restore or isolate an inoperable secondary containment isolation damper (reactor building automatic isolation dampers) if one damper associated with a penetration flow path is inoperable. If two dampers in a penetration are inoperable, CTS 3.7.C.4 must be entered immediately and an orderly shutdown is required. In addition, if a manual valve or blind flange is inoperable, CTS 3.7.C.4 must be entered immediately since CTS 3.7.C.3 only applies to the reactor building automatic dampers. ITS 3.6.4.2 ACTION A covers the condition of one or more penetration flow paths with one SCIV inoperable and it allows 8 hours to isolate the penetration flow path. However, it applies to all types of SCIVs, both automatic and manual. ITS 3.6.4.2 ACTION B covers the condition of one or more penetration flow paths with two SCIVs inoperable and it requires isolation of the penetration flow path within 4 hours. This changes the CTS by providing an 8 hour Completion Time for an inoperable non-automatic SCIV in a penetration flow path with one inoperable SCIV prior to requiring a unit shutdown and a 4 hour Completion Time if a penetration includes two SCIVs and both are inoperable prior to requiring a unit shutdown. The purpose of the CTS 3.7.C.3 isto provide a degree of assurance that the secondary containment penetration isolation boundary is maintained with an inoperable reactor building isolation damper. This change is acceptable because the Completion Time is consistent with safe operation under the specified Condition, the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of a DBA occurring during the allowed Completion Time. A penetration flow path with one inoperable SCIV must be isolated within an 8 hour Completion Time. The specified time period is reasonable considering the time required to isolate the penetration, and the probability of a design basis event that requires the SCIVs to be closed occurring during this short time is very low. With two SCIVs in one or more penetration flow paths inoperable, the affected penetration flow path must Monticello Page 7 of 9 Attachment 1, Volume 11, Rev. 0, Page 332 of 431
Attachment 1, Volume 11, Rev. 0, Page 333 of 431 DISCUSSION OF CHANGES ITS 3.6.4.2, SECONDARY CONTAINMENT ISOLATION VALVES (SCIVs) be isolated within 4 hours. The 4 hour Completion Time is also reasonable considering the time required to isolate the penetration and the probability of a design basis accident that requires the SCIVs to close or be closed occurring during this short time is very low. This change is designated as less restrictive because additional time is allowed to restore the components to within the LCO limits or to take compensatory measures than was allowed in the CTS. L.4 (Category 4 - Relaxation of Required Action) CTS 3.7.C.3 states that with an inoperable secondary containment isolation damper inoperable, isolate the affected duct by use of a closed damper or blind flange. ITS 3.6.4.2 Required Action A.1 requires that with one or more penetration flow paths with one SCIV inoperable, the affected penetration flow path be Isolated by use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange. This changes the CTS by allowing a manual valve as the means of isolating the penetration flow path. The change that requires de-activating the damper is discussed in DOC A.7. The purpose of CTS 4.7.C.3 is to provide assurance that the affected penetration flow path is isolated. This change is acceptable because the ITS Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing time to repair inoperable features. The ITS Required Actions are consistent with safe operation under the specified Condition, considering the operability status of the redundant systems of required features, the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of a design basis accident occurring during the repair period. This change allows the flow path to be isolated by a manual valve. The requirement to isolate the flow path is retained, and using a manual valve is an appropriate method of isolation. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. L.5 (Category 5 - Deletion of Surveillance Requirement) CTS 4.7.C.1 .b.(2) requires verification that each automatic damper actuates to its isolation position "After maintenance, repair or replacement work is performed on the damper or its associated actuator, control circuit, or power circuit." ITS SR 3.6.4.2.3 requires a similar test, however this specific Frequency is not required. This changes the CTS by deleting this post-maintenance Surveillance. The purpose of CTS 4.7.C.1.b.(2) is to verify OPERABILITY of secondary containment dampers following their maintenance, repair or replacement. This change is acceptable because the deleted Surveillance Requirement is not necessary to verify that the equipment used to meet the LCO can perform its required functions. Thus, appropriate equipment continues to be tested in a manner and at a Frequency necessary to give confidence that the equipment can perform its assumed safety function. Any time the OPERABILITY of a system or component has been affected by repair, maintenance, modification, or replacement of a component, post-maintenance testing is required to demonstrate the OPERABILITY of the system or component. This is described in the Bases for ITS SR 3.0.1 and required under SR 3.0.1. The OPERABILITY requirements for the secondary containment isolation valves are described in the Monticello Page 8 of 9 Attachment 1, Volume 11, Rev. 0, Page 333 of 431
Attachment 1, Volume 11, Rev. 0, Page 334 of 431 DISCUSSION OF CHANGES ITS 3.6.4.2, SECONDARY CONTAINMENT ISOLATION VALVES (SCIVs) Bases for ITS 3.6.4.2. In addition, the requirements of 10 CFR 50, Appendix B, Section Xi (Test Control), provide adequate controls for test programs to ensure that testing incorporates applicable acceptance criteria. Compliance with 10 CFR 50, Appendix B, is required under the unit operating license. As a result, post-maintenance testing will continue to be performed and an explicit requirement in the Technical Specifications is not necessary. This change is designated as less restrictive because Surveillances which are required in the CTS will not be required in the ITS. Monticello Page 9 of 9 Attachment 1, Volume 11, Rev. 0, Page 334 of 431
Attachment 1, Volume 11, Rev. 0, Page 335 of 431 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 335 of 431
Attachment 1, Volume 11, Rev. 0, Page 336 of 431 SCIVs 3.6.4.2 K) CTS 3.6 CONTAINMENT.SYSTEMS 3.6.4.2 Secondary Containment Isolation Valves (SCIVs) 3.7.C.l. LCO 3.6.4.2 Each SCIV shall be OPERABLE. 3.7.C.2 3.7.C-2 APPLICABILITY: MODES 1, 2, and 3, During movement of Rrecently irradiated fuel assemblies in the Psecondary4 containment, I (D During operations with a potential for draining the reactor vessel (OPDRVs). ACTIONS DOC L2 1. Penetration flow paths may be unisolated intermittently under administrative controls. DOC A4 2. Separate Condition entry is allowed for each penetration flow path. DOC A.5 3. Enter applicable Conditions and Required Actions for systems made inoperable by SCIVs. CONDITION REQUIRED ACTION COMPLETION TIME 3.7.C.3 A. One or more penetration A.1 Isolate the affected 8 hours flow paths with one penetration flow path by SCIV inoperable. use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange. AND BWR/4 STS 3.6.4.2-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 336 of 431
Attachment 1, Volume 11, Rev. 0, Page 337 of 431 SCIVs 3.6.4.2 CTS ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME 3.7.C.3 A.2 -----NOTES-----
- 1. Isolation devices in high radiation areas may be verified by use of administrative means.
- 2. Isolation devices that are locked, sealed, or otherwise secured may be verified by use of administrative means.
Verify the affected Once per 31 days penetration flow path is isolated. DOC L.3 B. ---- NOTE---- B.1 Isolate the affected 4 hours Only applicable to penetration flow path by penetration flow paths use of at least one closed with two isolation valves. and de-activated automatic valve, closed manual valve, or blind flange. One or more penetration flow paths with two SCIVs inoperable. 3.7.C.4.a C. Required Action and C.1 Be in MODE 3. 12 hours
'associated Completion Time of Condition A or IB AND not met in MODE 1, 2, or3. C.2 Be in MODE 4. 36 hours BWR/4 STS 3.6.4.2-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 337 of 431
Attachment 1, Volume 11, Rev. 0, Page 338 of 431 SCIVs 3.6.4.2 CTS ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME 3.7.C.4.b D. Required Action and D.1 LO-303 sNOTEa -- associated Completion LCO 3.0.3 is not applicable. Time of Condition A or B
}
not met during movement of jrecentlA Suspend movement of Immediately irradiated fuel assemblies in the
- IrecentlA irradiated fuel assemblies in the 0D asecondarA] containment TsecondarAj containment.
or during OPDRVs. AND D.2 Initiate action to suspend Immediately OPDRVs. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DOC M.3 SR 3.6.4.2.1 ----- h--NOTES----
- 1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.
- 2. Not required to be met for SCIVs that are open under administrative controls.
Verify each secondary containment isolation manual 31 days valve and blind flange that is not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed. DOC MA4 SR 3.6.4.2.2 Verify the isolation time of each power operated, In acc automatic SCIV is within limits. Ithe nservce 0 [92 daysM 4.7.C.l.b SR 3.6.4.2.3 Verify each automatic SCIV actuates to the isolation [I <months 0D position on an actual or simulated actuation signal. BWR/4 STS 3.6.4.2-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 338 of 431
Attachment 1, Volume 11, Rev. 0, Page 339 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.4.2, SECONDARY CONTAINMENT ISOLATION VALVES (SCIVs)
- 1. The brackets have been removed and the proper plant specific information/value has been provided.
Monticello Page 1 of I Attachment 1, Volume 11, Rev. 0, Page 339 of 431
Attachment 1, Volume 11, Rev. 0, Page 340 of 431 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 340 of 431
Attachment 1, Volume 11, Rev. 0, Page 341 of 431 SCIVs B 3.6.4.2 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.2 Secondary Containment Isolation Valves (SCIVs) BASES BACKGROUND The function of the SCIVs, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs) (Ref' A. Secondary containment isolation within the time limits specified for those isolation valves designed nd 2 0 to close automatically ensures that fission products that leak from primary containment following a DBA, or that are released during certain operations when primary containment is not required to be OPERABLE or take place outside primary containment, are maintained within the secondary containment boundary. The OPERABILITY requirements for SCIVs help ensure that an adequate Wsecondary containment boundary is maintained during and after an accident by minimizing potential paths to the environment. These 0 (which include isoiation aevices COnlSiSt O0eitner passive devices or active (automatic) plugs and caps devices. Manual valves, de-activated automatic valves secured in their as listed In closed position (including check valves with flow through the valve
.eec 3 secured), and blind flanges are considered passive devices.
ea Automatic SCIVsyclose on aasecondarA containment isolation signal to 1i i) establish a boundary for untreated radioactive material withinasecondaryM (I required to be containment following a DBA or other accidents. dosed during accident conditions Other penetration are isolated by the use of valves in the closed position (i or blind flanges. APPLICABLE The SCIVs must be OPERABLE to ensure theasecondarAj containment SAFETY barrier to fission product releases is established. The principal accidents ANALYSES for which theasecondarA containment boundary is required are a loss of coolant accident (Ref. 1) and a fuel handling accidentiinvolving handling irradiated fuel (i.e., fuel that has occupied part of a critical reactor lYcore within th rvosila ainsi e [seco containment ()i (Ref. 2). The secondary containment performs no active function in () response to either of these limiting events, but the boundary established by SCIVs is required to ensure that leakage from the primary containment is processed by the Standby Gas Treatment (SGT) System before being released to the environment. Maintaining SCIVs OPERABLE with isolation times within limits ensures that fission products will remain trapped insidelsecondary containment so that they can be treated by the SGT System prior to discharge to the environment. SCIVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). BWRI4 STS B 3.6.4.2-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 341 of 431
Attachment 1, Volume 11, Rev. 0, Page 342 of 431 SCIVs B 3.6.4.2 BASES LCO SCIVs form a part of thelsecondaryl containment boundary. The SCIV safety function is related to control of offsite radiation releases resulting from DBAs. The power operated, automatic isolation valves are considered OPERABLE when their isolation times are within limits and the valves actuate on an automatic isolation signal. The valves covered by this LCO, along with their associated stroke times, are listed in Reference 3. and blind flanges In pi;ace.7 the primary cn leastthandr e consideredn Eraloe thePER lTs are closeor opeere In accordne In MODS 4administrative control utomatic -nquncsfthesivaten are<) reucred duetoped ssurentemonrlant Mind flatiosare in Pthese These passive isolation valves or devices are listed in Reference 3. APPLICABILITY In MODES 1, 2, and 3, a DBA could lead to a fission product release to the primary containment that leaks to theRsecondary containment. Theorfore, the OPERABILITY of SCIVs is required. In MODES 4 and 5, the probability and consequences of these events are reduced due to pressure and temperature limitations in these MODES. Therefore, maintaining SCIVs OPERABLE is not required in MODE 4 or 5, except for other situations under which significant radioactive releases can be postulated, such as during operations with a potential for draining the reactor vessel (OPDRVs) or during movement ofarecentllow irradiated fuel assemblies in theisecondaryImcontyiundert aMoinita Irecentls irradiated fuel assemblies in theosecondard containment maya also occur in MODES 1, 2, and 3.tDue to radioactive decay, SCIVs ared only required to be OPERABLE during fuel handling involving handling n reent iradited uel(i~., uelthat has occupied part of a critical reactor ACTIONS The ACTIONS are modified by three Notes. The first Note allows penetration flow paths to be unisolated Inteamittently under flo atis controls. These controls consist -ofstationing a dedicated pvtr h is in continuous communication with the control room, at the controls of the isolation device. In this way, the penetration can be rapidly isolated when a need forisecondaryi containment isolation is indicated. e The second Note provides clarification tha~o th pupseo hi C separate Condition entry is allowed for each penetration flow path. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable SCIV. Complying with the Required Actions may allow for continued operation, and subsequent inoperable SCIVs are governed by subsequent Condition entry and application of associated Required Actions. BWRI4 STS B 3.6.4.2-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 342 of 431
Attachment 1, Volume 11, Rev. 0, Page 343 of 431 SCIVs B 3.6.4.2 BASES ACTIONS (continued) The third Note ensures appropriate remedial actions are taken, if necessary, if the affected system(s) are rendered inoperable by an inoperable SCIV. A.1 and A.2 In the event that there are one or more penetration flow paths with one SCIV inoperable, the affected penetration flow path(s) must be isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-activated automatic SCIV, a closed manual valve, and a blind flange. For penetrations isolated in accordance with Required Action A.1, the device used to isolate the penetration should be the closest available device to RsecondaryA containment. The Required Action must be completed within the 8 hour Completion Time. The specified time period is reasonable considering the time required to isolate the penetration, and the probability of a DBA, which requires the SCIVs to close, occurring during this short time is very low. For affected penetrations that have been isolated in accordance with Required Action A.1, the affected penetration must be verified to be isolated on a periodic basis. This is necessary to ensure thatlsecondarA (i containment penetrations required to be isolated following an accident, but no longer capable of being automatically isolated, will be in the isolation position should an event occur. The Completion Time of once per 31 days is appropriate because thelvaieslare operated under 0 administrative controls and the probability of their misalignment is low. This Required Action does not require any testing or device manipulation. Rather, it involves verification that the affected penetration remains isolated. Required Action A.2 is modified by two Notes. Note 1 applies to devices located in high radiation areas and allows them to be verified closed by use of administrative controls. Allowing verification by administrative controls is considered acceptable, since access to these areas is typically restricted. Note 2 applies to isolation devices that are locked, sealed, or otherwise secured in position and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since the function of locking, sealing, or securing components is to ensure that these devices are not inadvertently repositioned. Therefore, the probability of misalignment, once they have been verified to be in the proper position, is low. BWR/4 STS B 3.6.4.2-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 343 of 431
Attachment 1, Volume 11, Rev. 0, Page 344 of 431 SCIVs B 3.6.4.2 BASES ACTIONS (continued) B.1 With two SCIVs in one or more penetration flow paths inoperable, the affected penetration flow path must be isolated within 4 hours. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-activated automatic valve, a closed manual valve, and a blind flange. The 4 hour Completion Time is reasonable considering the time required to isolate the penetration and the probability of a DBA, which requires the SCIVs to close, occurring during this short time, is very low. The Condition has been modified by a Note stating that Condition B is only applicable to penetration flow paths with two isolation valves. This clarifies that only Condition A is entered if one SCIV is inoperable in each of two penetrations. C.1 and C.2 If any Required Action and associated Completion Time cannot be met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. D.1 and D.2 If any Required Action and associated Completion Time are not met, the plant must be placed in a condition in which the LCO does not apply. If applicable, the movement of recentlA irradiated fuel assemblies in the JO Isecondarbj containment must be immediately suspended. Suspension ofJv these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must be immediately initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and the subsequent potential for fission product release. Actions must continue until OPDRVs are suspended. BWR/4 STS B 3.6.4.2-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 344 of 431
Attachment 1, Volume 11, Rev. 0, Page 345 of 431 SCIVs B 3.6.4.2 BASES ACTIONS (continued) Required Action D.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If movingkrecentlJ irradiated fuel assemblies while in (j) MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving fuel while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of arecentlA irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown. SURVEILLANCE SR 3.6.4.2.1 REQUIREMENTS This SR verifies that each secondary containment manual isolation valve and blind flange that is not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside of the Isecondary containment boundary is within design limits. This SR does not require any testing or valve manipulation. Rather, it involves verification that those SCIVs in NsecondaryA containment that are capable of being mispositioned are in the correct position. Since these SCIVs are readily accessible to personnel during normal operation and verification of their position is relatively easy, the 31 day Frequency was chosen to provide added assurance that the SCIVs are in the correct positions. This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing. Two Notes have been added to this SR. The first Note applies to valves and blind flanges located in high radiation areas and allows them to be verified by use of administrative controls. Allowing verification by administrative controls is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, and 3 for ALARA reasons. Therefore, the probability of misalignment of these SCIVs, once they have been verified to be in the proper position, is low. A second Note has been included to clarify that SCIVs that are open under administrative controls are not required to meet the SR during the time the SCIVs are open. BWR/4 STS B 3.6.4.2-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 345 of 431
Attachment 1, Volume 11, Rev. 0, Page 346 of 431 SCIVs B 3.6.4.2 BASES SURVEILLANCE REQUIREMENTS (continued) SR 3.6.4.2.2 Verifying that the isolation time of each power operated, automatic SCIV is within limits is required to demonstrate OPERABILITY. The isolation time test ensures that the SCIV will isolate in a time period less than or equal to that assumed in the safety analyses. Thelsolatiopnad]( Fregu~enc of this SR I[in accordance nservice Testing] lProgr mo 92 days]. ' SR 3.6.4.2.3 Verifying that each automatic SCIV closes on a secondary containment isolation signal is required to prevent leakage of radioactive material from secondarj containment following a DBA or other accidents. This SR Q ensures that each automatic SCIV will actuate to the isolation position on LOo3.3.6.2,Secondary aecondar containment isolation signal. The LOGIC SYSTEM ( rnsbrmentation.. FUNCTIONAL TE ST ir R .6.2. overlaps this SR to provide complete 2 testing of the safety function. T [ ] month Frequency is based on the (D) need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at the [ I month Frequency. Therefore, the Frequency ( was concluded to be acceptable from a reliability standpoint. REFERENCES E SAR, Section[ 3 SAR, Section 1
- 3. jFSAR ction [ . 0 BWR/4 STS B 3.6.4.2-6 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 346 of 431
Attachment 1, Volume 11, Rev. 0, Page 347 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.4.2 BASES, SECONDARY CONTAINMENT ISOLATION VALVES (SCIVs)
- 1. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the plant specific nomenclature, number, reference, system description, analysis description, or licensing basis description.
- 2. The brackets have been removed and the proper plant specific information/value has been provided.
- 3. Typographical/grammatical error corrected.
- 4. These changes have been made for consistency with similar phrases in other places in the Bases and/or to be consistent with the Specifications.
- 5. The words in ISTS SR 3.6.4.2.2, stating that the isolation times are in the IST Program have been deleted. The IST Program does not include the times for the SCIVs. They are located in the Technical Requirements Manual.
- 6. This statement has been deleted since it is incorrect. Automatic SCIVs that are de-activated and secured in the closed position are not OPERABLE; they are inoperable.
- 7. The discussion in the LCO section about closed valves has been modified. This editorial preference is based on an incomplete and misleading discussion of the valves. This change does not modify the requirements or the interpretation of the requirements.
Monticello Page 1 of I Attachment 1, Volume 11, Rev. 0, Page 347 of 431
Attachment 1, Volume 11, Rev. 0, Page 348 of 431 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 11, Rev. 0, Page 348 of 431
Attachment 1, Volume 11, Rev. 0, Page 349 of 431 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.6.4.2, SECONDARY CONTAINMENT ISOLATION VALVES (SCIVs) There are no specific NSHC discussions for this Specification. Monticello Page 1 of I Attachment 1, Volume 11, Rev. 0, Page 349 of 431
c C C Monticello ITS Conversion Project Open Items By ITS Section/Speciflcation ITS/CTS 01 # Initiator Revision Description Status Assigned to: Due Date 3.6.4.2 151 Scott 1 The Bases discussion of ITS 3.6.4.2 ACTIONS Note 1discusses what Ismeant by penetration flow paths [may] be unisolated intermittently under administrative controls." Currently the Bases state, These controls consist of stationing a dedicated operator, who Is in continuous communication with the control room, at the controls of the Isolation device.' 0) Per Ben Krull, other administrative controls also need to be included, such as procedurally controlling the total number of square Inches of opening Inthe secondary containment to within 3 the capacity of the SGT System. 0 0 M a CD 0 0) Co D} CD
- U ;U 0 0 0
-o l*1 0 a) (4 2 Mo dy September 20 04Pae Sete be 20, Page 2 of 2 Monday, 2004
Attachment 1, Volume 11, Rev. 0, Page 351 of 431 ATTACHMENT 15 ITS 3.6.4.3, Standby Gas Treatment (SGT) System Attachment 1, Volume 11, Rev. 0, Page 351 of 431
Attachment 1, Volume 11, Rev. 0, Page 352 of 431 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1,Volume 11, Rev. 0, Page 352 of 431
C C C ITS 3.6.4.3 ITS ITS a) a) C, CD P0 0 0 0 0 -o A3.6A.3.1 1. Once per month, operate each train of the standby -o gas treatment system for t 10 continuous hours with the Inline heaters operating. a1) la 0 co CD) CD, C#) 0 0
-h CA) C,,
3.7/4.7 166 09/23/02 Amendment No. 66, 77, 0130 Page 1 of 5
C C ITS 3.6.4.3 0 ITS I l4.0 SURVEILLANCE REQUIREMENTS ( ) ACTION D EnterL( 3.0.3A immediate o s
- 0) 0)
C) 0 nrormance Requirement lasts CD
- a. Periodic Requirements a. At least once per 720 hours of system I
operation; or once per operating cyde, but not CD (1) The results of the in-place DOP tests at to exceed 18 months, whichever occurs first; or 3500 cfm (- 10%) on HEPA filters shall following painting, fire, or chemical release In show sI%DOP penetration. any ventilation zone communicating with the system while the system is operating that could 0 contaminate the HEPA filers or charcoal 0 (2) The results of In-place halogenated -o hydrocarbon tests at 3500 cfm (+/-+10%) on adsorbers, perform the following: l *0
- 0) a) charcoal banks shall show <1% Co CD penetration. (1) In-place DOP test the HEPA filter baanls.
CD, (3) The results of laboratory carbon sample (2) In-place test the charcoal adsorber banks See ITS 5.5 I analysis shall show 55% methy iodide with halogenated hydrocarbon tracer. ID U' -0 I penetration when tested In accordance with ASTM D3803-1989 at 30-C, 95% relative (3) Remove one carbon test sample from the CD, I humidity. charcoal edsorber In accordance with Regulatory Position C.B.b of Regulatory U' Guide 1.52, Revision 2. March 1978. C,) Subject this sample to a laboratory analysts 0 to verify methyl iodide removal effidency. -D' I- C, 3.7,4.7 167 8/1B/00 Amendment No. 60, tO 94, 112 Page 2 of 5
Attachment 1, Volume 11, Rev. 0, Page 355 of 431 IITS 3.6.4.3 ITS INSERT A
- c. With one standby gas treatment system train inoperable,
- 1) The following activities may continue for up to 7 days:
ACTION A (a) Movement of recently irradiated fuel assemblies in secondary containment; I (b) movement-oflttfuel cask in thejea~tertbuifding, and I (c) Operations with the potential to drain the reactor vessel. 2)Afi:er 7 days: Add proposed Required Action C Note (a) Immediatelyuspe1Zl movement of the fuel c e reactor < Ibu~il and l (b) Immediately place the operable standby gas treatment system ACTION C - train in operation, or (1) Immediately suspend movement of recently irradiated fuel assemblies in secondary containment; and (2) Immediaty suspend operations with the potential to drain 9 the reactor vessel. Add proposed Required Acton E.1 Note I I
- d. With both standby gas treatment trains inoperable immediately suspend:
ACTION E - Movement of recently irradiated fuel assemblies in secondary containment; [Movemszfficfiuel cask in the-reebs ildingj Operations with the I) potential for draining the reactor vessel. Insert Page 167 Page 3 of 5 Attachment 1, Volume 11, Rev. 0, Page 355 of 431
c C C ITS 3.6.4.3 ITS ITS rSee TS 5.5} 3.0 LIMING CONDmONS FOR OPERATION l 4.0 SURVEILLANCE REQUIREMENTS
........... ~~~~~~~~ _ ... I-
- b. At least once per operating cyde, but not to
- b. The system shall be shown to be operable with:
exceed 18 months, the following conditions (1) Combined filter pressure drop 86Inches shall be demonstrated for each standby gas water. treatment systern
- 0) - { See ITS 5.5 } 0)
(2) Inine heater power output k18kW. a (1) Pressure drop across the combined filters of each standby gas treatment system 0 drcult shall be measured at 3500 cfm 0 SR 3.6.4.3.3 c. The system shall be shown to be operable with automatic Initiation upon receipt of the following A (+/-10%) flow rate. Inputs: 3 CD (2) OperabilIty of Intine heater at nominal rated I (a) Low Low Reactor Water Level, ok/ d . power shall be verified. II /11 0 (b) Hlgh lpressure, or At least once perR n omtic SR 3.6.4.3.3 I1] -A CD (c) Rea or buiding vent lat enum hig Initiation of each standby gas treatment system circuit shall be demonstrated. 0 raton, or/ (di) ofefueling floor high dilalon Post Maintenance Testing CD CD
- 3. Post Maintenance Requirements a. After any maintenance or testing that could
- A) affect the leak tight integrity of the HEPA filters,
- a. After any maintenance or testing that could perform In-place DOP tests on the HEPA filters.
affect the HEPA fter or HEPA filter mounting -0 frame leak tight Integrlty, the results of the b. After any maintenance or testing that could In-place DOP tests at 3500 cfm (+/-10%) on affect the leak tight integrity of the charcoal See ITS .5 5 CD, HEPAfilters shad show s1% DOP penetration. adsorber banks, perform halogenated l CA) hydrocarbon tests on the charcoal / 0)
- b. After any maintenance or testing that could absorbers.
affect the charcoal adsorber leak tight Integrity, the results of In-place halogenated hydrocarbon C.1 tests at 3500 cfm(10%) on charcoal adsorber banks shall show I1% penetration. 3.7/4.7 168 10x95 Amendment No. 94 Page 4 of 5
c C ITS 3.6.4.3 C3\ ITS 0
- 4. Protective Function - A system protective action which results from the protective action of the channel monitoring a 10 particular plant condition.
0 R. Rated Neutron Flux Rated flux Is the neutron flux that corresponds to a steady-state power level of 1775 thermal megawatts. CD CD S. Rated Thermal Power - Rated thermal power means a steady-state power level of 1775 thermal megawatts. T. Reactor Coolant System Pressure or Reactor Vessel Pressure - Unless otherwise indicated, reactor vessel pressures listed In See ITS Chapter 1.01 '.7I the Technical Specifications are those existing In the vessel steam space. U. Refueling Operation and Refueling Outage - Refueling Operation Is any operation when the reactor water temperature Is less 0 0 than 21 2 F and movement of fuel or core components Is In progress. For the purpose of designating frequency of testing and 0 surveillance, a refueling outage shall mean a regularly scheduled refueling outage; however, where such outages occur within 8 0 3 months of the completion of the previous refueling outage, the required surveillance testing need not be performed until the next regularly scheduled outage. -o V. Safety Lmit - The safety Omits are limits below which the maintenance of the cladding and primary system integrity are assured. Exceeding such a limit is cause for plant shutdown and review by the Commission before resuiption of plant operation. Operation beyond such a Oimit may not In Itself result In serious consequences but it Indicates an operational deficiency subject to reaulatorv review. CD -U NP Sr,-ndanrv Czntinmnt lnteritvllSecndarv Containment InteoritV means that the reactor building is closeU
........................
1 conasonace e -0 CD CD) I . At least one door In each access opening Isdosed See ITS 3.6.4.1} (n Ul LCO 3.6.4.3 4'
- 2. The standby gas treatment system Is operable.
0
-h l 3. Al reactor building ventilation system automatic Isolation valves are operable or are secured in the closed positin. lSee ITS 3.6.4.2}
CA) Ca) X. Sensor Check - A qualitative determination of operability by observation of sensor behavior during operation. I his deermina ton I shall Include, where possible, comparison with other Independent sensors measuring the same variable. See ITS Chapter 1.01 1.0 4 9116/98 Amendment No. 4-., 102 Page 5 of 5
Attachment 1, Volume 11, Rev. 0, Page 358 of 431 DISCUSSION OF CHANGES ITS 3.6.4.3, STANDBY GAS TREATMENT (SGT) SYSTEM ADMINISTRATIVE CHANGES A.1 Inthe conversion of the Monticello Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1433, Rev. 3, "Standard Technical Specifications General Electric Plants, BWR/4" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 CTS 3.7.B.1 requires the SGT System to be OPERABLE whenever the secondary containment integrity is required and CTS 3.7.B.1.a references the conditions of CTS 3.7.C.2.(a) through (f). ITS LCO 3.6.4.3 requires the SGT System to be OPERABLE during MODES 1, 2, and 3, during movement of recently irradiated fuel assemblies in the secondary containment, and during operations with a potential for draining the reactor vessel (OPDRVs). This changes the CTS by deleting a cross reference to the secondary containment Applicability and replacing it with the specific Applicability for the SGT System. The purpose of CTS 3.7.B.1 is to ensure the SGT System is OPERABLE when necessary to satisfy the safety analyses. This change deletes a cross reference to the secondary containment Applicability and replaces with the specific Applicability for the SGT System. The proposed Applicability for the SGT System is the same as the Applicability for the Secondary Containment in ITS 3.6.4.1. Changes to the Applicability of the Secondary Containment is discussed in the Discussion of Changes for ITS 3.6.4.1. The changes identified in the Discussion of Changes for ITS 3.6.4.1 are applicable for the SGT System. This change is considered to be a format change consistent with the ISTS. This change is designated as administrative because it does not result in technical changes to the CTS. A.3 CTS 3.7.B.1 .c.2)(b)(1) and 3.7.B.1 .d state to immediately suspend movement of recently irradiated fuel assemblies in the secondary containment. ITS 3.6.4.3 ACTIONS C and E include the same requirement, however a Note has been added that states that LCO 3.0.3 is not applicable. This changes the CTS by adding this Note. The purpose of CTS 3.7.B.1 .c.2)(b)(1) and 3.7.B.1 .d, in part, is to provide the appropriate actions when one or two SGT subsystems, as applicable, are inoperable while the unit is moving recently irradiated fuel in the secondary containment. This change adds a Note that states LCO 3.0.3 is not applicable. This Note has been added because ITS LCO 3.0.3 has been added to ITS Section 3.0 in accordance with DOC M.1. This Note is necessary because if moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of recently irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown. Since ITS LCO 3.0.3 is not currently included in the CTS this change is Monticello Page 1 of 5 Attachment 1, Volume 11, Rev. 0, Page 358 of 431
Attachment 1, Volume 11, Rev. 0, Page 359 of 431 DISCUSSION OF CHANGES ITS 3.6.4.3, STANDBY GAS TREATMENT (SGT) SYSTEM considered administrative. This change is designated as administrative because it does not result in technical changes to the CTS. A.4 CTS 3/4.7.8.2 specifies the performance requirements for the SGT subsystems while CTS 3/4.7.B.3 specifies the post maintenance requirements for the SGT subsystems. ITS 3.6.4.3.2 requires the performance of the required SGT filter testing in accordance with the Ventilation Filter Testing Program (VFTP). CTS 3/4.7.B does not include a VFTP, but the requirements that make up the VFTP are being moved to ITS 5.5. This changes CTS by requiring testing in accordance with the VFTP, whose requirements are being moved to ITS 5.5. This change is acceptable because filter testing requirements are being moved to the VFTP as part of ITS 5.5, and ITS SR 3.6.4.3.2 references the VFTP for performing these tests. This change is designated as administrative because it does not result in technical changes to the CTS. A.5 CTS 4.7.B.2.d requires verification of automatic initiation of each SGT subsystem each "operating cycle." ITS SR 3.6.4.3.3 requires this same test however it is required to be performed every "24 months." This changes the CTS by changing the Frequency from "operating cycle" to "24 months." This change is acceptable because the current "operating cycle" is "24 months". In letter L-MT-04-036, from Thomas J. Palmisano (NMC) to the USNRC, dated June 30, 2004, NMC has proposed to extend the fuel cycle from 18 to 24 months and the same time has performed an evaluation in accordance with Generic Letter 91-04 to extend the unit Surveillance Requirements from 18 months to 24 months. CTS 4.7.B.2.d was included in this evaluation. This change is designated as administrative because it does not result in any technical changes to the CTS. A.6 CTS 3.7.8.1 allows one SGT subsystem to be inoperable with reactor water temperature > 2120 F for 7 days "provided that all active components in the other standby gas treatment system are operable." ITS 3.6.4.3 does not explicitly state this requirement in the ACTION for one inoperable SGT subsystem. This changes the CTS by deleting a provision to when the 7 day allowed outage time is applicable. The purpose of the CTS 3.7.8.1 statement is to ensure that the 7 days is only applicable if the other redundant SGT subsystem is OPERABLE. However, CTS 3.7.8.2 provides an action for when two SGT subsystems are inoperable. In addition, ITS 3.6.4.3 ACTION D provides the requirements when two SGT trains are inoperable. Thus, the deletion of this "provided" statement is acceptable since it is redundant and unnecessary. Both the CST and ITS ensure that proper actions are taken when two SGT subsystems are inoperable. This change is designated as administrative since it does not result in any technical changes to the CTS. A.7 These changes to CTS 3.7.8.1, CTS 3.7.B.1.a, and CTS 3.7.B.1.b, and the addition of CTS 3.7.B.1.c and d are provided in the Monticello ITS consistent with the Technical Specifications Change Request submitted to the USNRC for Monticello Page 2 of 5 Attachment 1, Volume 11, Rev. 0, Page 359 of 431
Attachment 1, Volume 11, Rev. 0, Page 360 of 431 DISCUSSION OF CHANGES ITS 3.6.4.3, STANDBY GAS TREATMENT (SGT) SYSTEM approval in NMC letter L-MT-05-013, from Thomas J. Palmisano (NMC) to USNRC, dated April 12, 2005. As such, these changes are administrative. A.8 This change to CTS 4.7.B.2.c is provided in the Monticello ITS consistent with the Technical Specifications Change Request submitted to the USNRC for approval in NMC letter L-MT-04-036, from Thomas J. Palmisano (NMC) to USNRC, dated June 30, 2004. As such, this change is administrative. MORE RESTRICTIVE CHANGES M.1 CTS 3.7.B.1.a states, in part, that if the inoperable SGT subsystem is not restored to OPERABLE status within 7 days, then 36 hours is allowed to be in a Cold Shutdown condition. ITS 3.6.4.3 ACTION B requires the unit to be in MODE 3 in 12 hours and MODE 4 in 36 hours under the same conditions. This changes the CTS by requiring the unit be in an intermediate condition (MODE 3) within 12 hours. The purpose of the 3.7.B.1.a is to place the unit in a MODE in which the LCO is no longer applicable in a reasonable amount of time. This change will now require the unit to be at an intermediate condition sooner than is currently required. The proposed Completion Time of 12 hours to be in MODE 3 is acceptable because it is consistent with the time proposed in other Specifications that requires the unit to be in MODE 3. This change is designated as more restrictive since in the ITS the unit will be required to be in an intermediate condition sooner than is currently required by the CTS when one SGT subsystem is inoperable. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA.1 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 3.7.B.1 states that two "separate and independent" Standby Gas Treatment (SGT) System trains shall be OPERABLE. ITS LCO 3.6.4.3 requires two SGT subsystems to be OPERABLE, but does not include the details of what constitutes OPERABILITY. This changes the CTS by moving the detail that the trains must be "separate and independent" to the ITS Bases. The removal of these details, which are related to system design, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement that two SGT subsystems shall be OPERABLE. Also, this change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of Monticello Page 3 of 5 Attachment 1, Volume 11, Rev. 0, Page 360 of 431
Attachment 1, Volume 11, Rev. 0, Page 361 of 431 DISCUSSION OF CHANGES ITS 3.6.4.3, STANDBY GAS TREATMENT (SGT) SYSTEM changes to the Bases to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications. LA.2 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 3.7.B.2.c requires the SGT subsystem be shown to be OPERABLE with automatic initiation upon receipt of the following inputs: Low Low Reactor Water Level; high drywell pressure; reactor building ventilation plenum high radiation; or refueling floor high radiation. ITS SR 3.6.4.3.3 requires verification that each SGT subsystem actuates on an initiation signal. This changes the CTS by moving the specific type of actuation signals to the ITS Bases. The removal of these details, which are related to system design, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement to verify each SGT subsystem actuates on an initiation signal. Also, this change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES L.1 (Category 3 - Relaxation of Completion Time) CTS 3.7.B.1.b requires the unit to be in Cold Shutdown (MODE 4) if two SGT subsystems are inoperable with reactor water temperature > 21 20F. ITS 3.6.4.3 ACTION D requires the unit to enter LCO 3.0.3 under the same conditions. This will require the unit to initiate action within 1 hour to place the unit in MODE 2 within 7 hours, MODE 3 within 13 hours, and MODE 4 within 37 hours. This changes the CTS by requiring the unit to enter LCO 3.0.3 instead of requiring a unit shutdown to MODE 4 within 36 hours, which effectively extends the time the unit is required to be in MODE 4 by 1 hour, however it also requires the unit to be at the specified intermediate conditions sooner. The purpose of CTS 3.7.B.1.b is to place the unit in a MODE in which the LCO is not applicable within a reasonable amount of time. This change is acceptable because the allowed Completion Times are reasonable, based on operating experience, to reach required unit conditions from full power conditions in an orderly manner and without challenging unit systems. This change is also acceptable because it requires the unit to be in intermediate conditions (MODE 2 and MODE 3) at 7 hours and 13 hours, respectively, which is not currently required. This portion of the change reduces the time the unit would be allowed to continue to operate in MODES 1 and 2 once the condition is identified. The consequences of a loss of coolant accident are reduced when the reactor is shutdown and a controlled cooldown is already in progress. This change is Monticello Page 4 of 5 Attachment 1, Volume 11, Rev. 0, Page 361 of 431
Attachment 1, Volume 11, Rev. 0, Page 362 of 431 DISCUSSION OF CHANGES ITS 3.6.4.3, STANDBY GAS TREATMENT (SGT) SYSTEM designated as less restrictive because additional time is allowed to reach MODE 4 in the ITS than is allowed in the CTS. L.2 (Category 6 - Relaxation Of Surveillance Requirement Acceptance Criteria) CTS 3.7.B.2.c requires verification of the automatic actuation of the SGT subsystem upon a receipt of the specified inputs (i.e., test signal). ITS SR 3.6.4.3.3 specifies that the signal may be from either an "actual" or simulated (i.e., test) signal. This changes the CTS by explicitly allowing the use of an actual signal for the test. The purpose of CTS 3.7.B.2.c is to ensure that the SGT subsystem operates correctly upon receipt of an actuation signal. This change is acceptable because it has been determined that the relaxed Surveillance Requirement acceptance criteria are not necessary for verification that the equipment used to meet the LCO can perform its required functions. Equipment cannot discriminate between an "actual," "simulated," or "test" signal and, therefore, the results of the testing are unaffected by the type of signal used to initiate the test. This change allows taking credit for unplanned actuation if sufficient information is collected to satisfy the Surveillance test requirements. The change also allows a simulated signal to be used, if necessary. This change is designated as less restrictive because less stringent Surveillance Requirements are being applied in the ITS than were applied in the CTS. L.3 (Category I - Relaxation of LCO Requirements) CTS 3.7.B.1 requires the SGT System to be OPERABLE whenever the secondary containment integrity is required and CTS 3.7.C.2.c requires the secondary containment to be OPERABLE when the fuel cask is being moved within the reactor building. CTS 3.7.B.1.c.1)(b), CTS 3.7.B.1.c.2)(a), and CTS 3.7.B.1.d provides actions when this is not met. ITS 3.6.4.3 does not include this requirement. This changes the CTS by deleting the requirement to maintain the SGT System OPERABLE when the fuel cask is being moved within the reactor building. The purpose of the CTS 3.7.B.1, in part, is to ensure the SGT System is OPERABLE when the fuel cask is being moved in the reactor building. This change is acceptable because the LCO requirements continue to ensure that the structures, systems, and components are maintained consistent with the safety analyses. This change is also acceptable because the USAR contains restrictions on the movement of heavy loads based on the heavy loads analysis. The bounding design basis fuel handling accident assumes an irradiated fuel assembly is dropped onto an array of irradiated fuel assemblies seated within the reactor pressure vessel. The movement of other loads is administratively controlled based on available analysis for the individual load. USAR Section 12.2.5 includes a description of the manner in which Monticello controls heavy loads. This USAR Section provides a description of safe load paths and requires the establishment of load handling procedures in accordance with NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants." The removal of fuel cask controls from the Technical Specifications has been approved by the NRC in the FitzPatrick ITS. Therefore, these activities are not restricted in the Technical Specifications. This change is designated as less restrictive because less stringent LCO requirements are being applied in the ITS than were applied in the CTS. Monticello Page 5 of 5 Attachment 1, Volume 11, Rev. 0, Page 362 of 431
Attachment 1, Volume 11, Rev. 0, Page 363 of 431 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 363 of 431
Attachment 1, Volume 11, Rev. 0, Page 364 of 431 SGT System 3.6.4.3 CTS 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System 3.7.B.1 LCO 3.6.4.3 RTwoM SGT subsystems shall be OPERABLE. 0D 3.7.B.1 APPLICABILITY: MODES 1, 2, and 3, During movement of jrecentlyI irradiated fuel assemblies in the EsecondaryM containment, During operations with a potential for draining the reactor vessel
} 0D (OPDRVs).
ACTIONS CONDITION J REQUIRED ACTION j COMPLETION TIME 3.7.B.1.a, A. One SGT subsystem A.1 Restore SGT subsystem to 7 days 3.7.B.1.c.1) inoperable. OPERABLE status. 3.7.B.1.a B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not AND met in MODE 1, 2, or 3. B.2 Be in MODE 4. 36 hours C. Required Action and 0 NOTE--a 3.7.B.1.C.2) associated Completion LCO 3.0.3 is not applicable. Time of Condition A not met during movement of IrecentIA irradiated fuel assemblies in the isecondaryQ containment or during OPDRVs. C.1 OR Place OPERABLE SGT subsystem in operation. Immediately
} 0D C.2.1 Suspend movement of Immediately MrecentlyT irradiated fuel assemblies in Msecondaryl containment.
} 0D AND BWR/4 STS 3.6.4.3-1 Rev. 3.0, 03/31104 Attachment 1, Volume 11, Rev. 0, Page 364 of 431
Attachment 1, Volume 11, Rev. 0, Page 365 of 431 SGT System 3.6.4.3 CTS ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C.2.2 Initiate action to suspend Immediately OPDRVs. 3.7.B.1.b D. Two SGT subsystems D.1 Enter LCO 3.0.3. Immediately inoperable in MODE 1, 2,or3. 3.7.B.1.d E. Two SGT subsystems E.1 ------ NOTE-- --- inoperable during LCO 3.0.3 is not applicable. movement ofIrecently irradiated fuel assemblies in the Suspend movement of [recentlyT irradiated fuel Immediately 0 IsecondarA containment or during OPDRVs. assemblies in ISsecondaryA containment. AND E.2 Initiate action to suspend Immediately OPDRVs. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
+
4.7.B.1 SR 3.6.4.3.1 Operate each SGT subsystem for 2 01 OM continuous 31 days hoursfwith heaters operating. J-0 314.7B..2. SR 3.6.4.3.2 Perform required SGT filter testing in accordance In accordance 344.7.B.3 with the Ventilation Filter Testing Program (VFTP). with the VFTP 3.7.B.2.c. 4.73B2.d SR 3.6.4.3.3 Verify each SGT subsystem actuates on an actual [ emonths0( or simulated initiation signal. BWR/4 STS 3.6.4.3-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 365 of 431
Attachment 1, Volume 11, Rev. 0, Page 366 of 431 SGT System 3.6.4.3 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.6.4.3.4 ealech SGT filter cooler bypass dampec[18] months] be opened and the fan started. 0. BWR/4 STS 3.6.4.3-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 366 of 431
Attachment 1, Volume 11, Rev. 0, Page 367 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.4.3, STANDBY GAS TREATMENT (SGT) SYSTEM
- 1. The brackets have been removed and the proper plant specific information/value has been provided.
- 2. The bracketed requirement is deleted. The SGT subsystem arrangement to ensure the removal of decay heat from an idle train consists of a flow path containing a cooling air check valve in each subsystem, and a common, opened crosstie line with a restrictive orifice. This change is consistent with the current licensing basis.
Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 367 of 431
Attachment 1, Volume 11, Rev. 0, Page 368 of 431 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 368 of 431
Attachment 1, Volume 11, Rev. 0, Page 369 of 431 SGT System B 3.6.4.3 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.3 Standby Gas Treatment (SGT) System BASES S BACKGROUND The SGT System is required by 10 CFR 50, p ix A, GDC 41, 1"Containment os ere eanu (Ref. 1). The function of the SGT System is to ensure that radioactive materials that leak from the primary containment into the~secondarAj containment following a Design Basis Accident (DBA) are filtered and adsorbed prior to exhausting to the environment. The SGT System consists of two fully redundant subsystems, each with its own set of ductwork, dampers, charcoal filter train, and controls. Each charcoal filter train consists of (components listed in order of the direction of the air flow):
- a. A demister or moislTeparatoirm
- b. An electric heaters
- c. Wrefiiter, 0--q. A high efficiency particulate air (HEPA) filter 3 j-- . A charcoal adsorbe A second HEPA filterad E A centrifugal fan.
The sizing of the SGT System equipment and components is based on the results of an infiltration analysis, as well as an exfiltration analysis of the secondarA contain ment./The internal pressure gfhe SGT Syster, boundary region is m rna ned at a negative pressure of [0.25] inches/ water gauge when the system is in operation, whigh represents the internal pressure r quired to ensure zero exfiltrati6n of air from theI 3 _ ~~~building when ex osed to a [IO] mph wind lwn taageof[5°oi the building. I The demister is provided to remove entrained water in the air, while the electric heater reduces the relative humidity of the airstream to less than
- 17Cg% (Ref. 2). Thel prefiller removes lacart e mater we protects the charcoal from 0e HEPA filter removes fine particulate matter and fouling. The charcoal adsorber removes gaseous elemental iodine and organic iodides, and the final HEPA filter collects any carbon fines exhausted from the charcoal adsorber.
BWR/4 STS B 3.6.4.3-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 369 of 431
Attachment 1, Volume 11, Rev. 0, Page 370 of 431 B 3.6.4.3 0 INSERT I Exfiltration from the secondary containment does not exceed 4,000 cfm with wind speeds on the order of 40 mph, from a start condition of negative internal pressure of 0.25 inches water gauge under calm wind conditions. Insert Page B 3.6.4.3-1 Attachment 1, Volume 11, Rev. 0, Page 370 of 431
Attachment 1, Volume 11, Rev. 0, Page 371 of 431 SGT System B 3.6.4.3 BASES BACKGROUND (continued) The SGT System automatically starts and operates in response to [ INSERT22 actuation signals indicative of conditions or an accident that could require A operation of the s stern.Aollowinginitiation both c arcoa ter traina t l hAgo v rficatinthat l redundant subsysn is normally shut down subsystems areoperatig J_ both __ 3and4 APPLICABLE The design basis for the SGT System is to mitigate the consequences of SAFETY a loss of coolant accident and fuel handling accidentsainvolving handling / (9 ANALYSES dfuel (i.e., fuel that has occupiedart of a critical reacto core within the previou a (Reh. . For all events analyzed, the ID SGT System is shown to be automatically initiated to reduce, via Titratlonj and adsorption, the radioactive material released to the environment. The SGT System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO Following a DBA, a minimum of one SGT subsystem is required to maintain theasecondarn containment at a negative pressure with respect to the environment and to process gaseous releases. Meeting the LCO 0 requirements for two OPERABLE subsystems ensures operation of at least one SGT subsystem in the event of a single active failure. APPLICABILITY In MODES 1, 2, and 3, a DBA could lead to a fission product release to primary containment that leaks to secondary containment. Therefore, SGT System OPERABILITY is required during these MODES. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the SGT System in OPERABLE status is not required in MODE 4 or 5, except for other situations under which significant releases of radioactive material can be postulated, such as during operations with a potential for draining the reactor vessel (OPDRVs) or during movement ofgrecently irradiated fuel assemblies in the secondaryM containment. RDue to radioactive decay, the SGT System is only required to be OPERABLE during fuel handling involving handling 0 irradiated fuel (i.e., fuel that has occupied part of a critical reactor hus core within the previousl lXav ACTIONS A.1 With one SGT subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status in 7 days. In this Condition, the remaining OPERABLE SGT subsystem is adequate to perform the required 0 radioactivity release control function. However, the overall system reliability is reduced because a single failure in the OPERABLE BWR/4 STS B 3.6.4.3-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 371 of 431
Attachment 1, Volume 11, Rev. 0, Page 372 of 431 B 3.6.4.3 ( ) INSERT 2 The SGT System is initiated by Reactor Vessel Water Level - Low Low, Drywell Pressure - High, Reactor Building Ventilation Exhaust Radiation - High, and Refueling Floor Radiation - High signals. 0, INSERT 3 the SGT subsystem A starts and both the inlet and outlet dampers of the reactor building
- ventilation ducts are isolated. A failure of the SGT subsystem A to start within the required time delay will initiate the automatic start and alignment of SGT subsystem B.
Automatic valves provide for isolation of each SGT subsystem. Each subsystem can draw air to remove radioactive decay heat from the charcoal adsorber. Insert Page B 3.6.4.3-2 Attachment 1, Volume 11, Rev. 0, Page 372 of 431
Attachment 1, Volume 11, Rev. 0, Page 373 of 431 SGT System B 3.6.4.3 BASES ACTIONS (continued) subsystem could result in the radioactivity release control function not being adequately performed. The 7 day Completion Time is based on consideration of such factors as the availability of the OPERABLE redundant SGT System and the low probability of a DBA occurring during this period. B.1 and B.2 If the SGT subsystem cannot be restored to OPERABLE status within the required Completion Time in MODE 1, 2, or 3, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. C.1, C.2.1. and C.2.2 During movement ofjrecentlyM irradiated fuel assemblies, in the Msecondaryl containment or during OPDRVs, when Required Action A.1 v cannot be completed within the required Completion Time, the OPERABLE SGT subsystem should immediately be placed in operation. This action ensures that the remaining subsystem is OPERABLE, that no failures that could prevent automatic actuation hVE occurEd, and that (5 ) any other failure would be readily detected. An alternative to Required Action C.1 is to immediately suspend activities that represent a potential for releasing a significant amount of radioactive material to theasecondaryl containment, thus placing the plant in a condition that minimizes risk. If applicable, movement of RErecentlyj 0 irradiated fuel assemblies must immediately be suspended. Suspension of these activities must not preclude completion of movement of a component to a safe position. Also, if applicable, actions must immediately be initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended. BWR/4 STS B 3.6.4.3-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 373 of 431
Attachment 1, Volume 11, Rev. 0, Page 374 of 431 SGT System B 3.6.4.3 BASES ACTIONS (continued) The Required Actions of Condition C have been modified by a Note stating that LCO 3.0.3 is not applicable. If moving recentlyl irradiated 0 fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If movingarecentlA irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of jrecentlyl irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown. D.1 If both SGTS subsystems are inoperable in MODE 1, 2, or 3, the SGT system may not be capable of supporting the required radioactivity release control function. Therefore, actions are required to enter LCO 3.0.3 immediately. E.1 and E.2 When two SGT subsystems are inoperable, if applicable, movement of grecently] irradiated fuel assemblies in [secondary4 containment must immediately be suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must immediately be initiated to suspend 0 OPDRVs in order to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended. Required Action E.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If movingirecentll irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving mrecentlh irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of Irecently irradiated fuel ( assemblies would not be a sufficient reason to require a reactor shutdown. BWR/4 STS B 3.6.4.3-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 374 of 431
Attachment 1, Volume 11, Rev. 0, Page 375 of 431 SGT System B 3.6.4.3 BASES SURVEILLANCE SR 3.6.4.3.1 REQUIREMENTS Operating each SGT subsystem for 2 10j continuous hours ensures that abothM subsystems are OPERABLE and that all associated controls are 0D functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. Operationgwith the heaters on (automatic heater cycling to maintain temperatures for 0 2 1O0continuous hours every 31 days eliminates moisture on the adsorbers and HEPA filters. The 31 day Frequency was developed in 0 consideration of the known reliability of fan motors and controls and the I redundancy available in the system. SR 3.6.4.3.2 This SR verifies that the required SGT filter testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). Specific test frequencies and additional information are discussed in detail in the VFTP. SR 3.6.4.3.3 This SR verifies that each SGT subsystem starts on receipt of an actual m4 or simulated initiation signal. While this Surveillance can be performed with the reactor at power, operating experience has shown that these components usually pass the Surveillance when performed at the Lco3.3.6.2.Secondary month Frequency. The LOGIC SYSTEM FUNCTIONAL TEST in( Containment Isolation
.i
-[E 6.2 overlaps this SR to provide complete testing of the safety function. Therefore, the Frequency was found to be acceptable from a 0
reliability standpoint. [ SR 3.6.4.3.4/ This SR verifies that t filter cooler bypass damper can b pened and the fan started. Th' ensures that the ventilation mode SGT System operation is ava' le. While this Surveillance can be erformed with the reactor at po r, operating experience has shown at these components 0D usually pa the Surveillance when performed a e [18] month Frequen , which is based on the refueling c e. Therefore, the Freq ncy was found to be acceptable fro a reliability standpoint. ] BWR/4 STS B 3.6.4.3-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 375 of 431
Attachment 1, Volume 11, Rev. 0, Page 376 of 431 SGT System B 3.6.4.3 K) 0 0 03 BWRI4 STS B 3.6.4.3-6 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 376 of 431
Attachment 1, Volume 11, Rev. 0, Page 377 of 431 JUSTIFICATION FOR DEVIATIONS ITS 3.6.4.3 BASES, STANDBY GAS TREATMENT (SGT) SYSTEM
- 1. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the plant specific nomenclature, number, reference, system description, analysis description, or licensing basis description.
- 2. The brackets have been removed and the proper plant specific information/value has been provided.
- 3. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, NEI 01-03, Section 5.1.3.
- 4. These changes have been made for consistency with similar phrases in other parts of the Bases and/or to be consistent with the Specification.
- 5. Changes made to be consistent with changes made to the Specification.
Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 377 of 431
Attachment 1, Volume 11, Rev. 0, Page 378 of 431 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 11, Rev. 0, Page 378 of 431
Attachment 1, Volume 11, Rev. 0, Page 379 of 431 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.6.4.3, STANDBY GAS TREATMENT (SGT) SYSTEM There are no specific NSHC discussions for this Specification. Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 379 of 431
Attachment 1, Volume 11, Rev. 0, Page 380 of 431 ATTACHMENT 16 Improved Standard Technical Specifications (ISTS) not adopted in the Monticello ITS, Attachment 1, Volume 11, Rev. 0, Page 380 of 431
, Volume 11, Rev. 0, Page 381 of 431 ISTS 3.6.1.4, Drywell Pressure , Volume 11, Rev. 0, Page 381 of 431
Attachment 1, Volume 11, Rev. 0, Page 382 of 431 ISTS 3.6.1.4 Markup and Justification for beviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 382 of 431
Attachment 1, Volume 11, Rev. 0, Page 383 of 431 Drywell Pressure 3.6.1.4 3.6 CONTAIN ENT SYSTEMS 3.6.1.4 D ell Pressure LCO 3.6.1.4 Drywell pressure sh I be [< 0.75 psig]. APPLICABILI : MODES 1, 2, and 3. ACTIONS C DITION lEQUIRED ACTION COMPLETION TIME A. Drywel pressure not A.1 estore drywell pressure to 1 hour within Iit. [within limit. B. Requi d Action and associ ted Completion B.1 Be in MODE 3. 12 hours 0 Time ot met. AND B.2 Be in MODE 4. 36 hours SURVEIL NCE REQUIREMENTS _ _l SURVEIL NCE FREQUENCY SR 3.6. .4.1 Verify drywell pre sure is within limit. 12 hours BWRI4 TS 3.6.1.4-1 Rev. 3.0, 03/31104 Attachment 1, Volume 11, Rev. 0, Page 383 of 431
Attachment 1, Volume 11, Rev. 0, Page 384 of 431 JUSTIFICATION FOR DEVIATIONS ISTS 3.6.1.4, DRYWELL PRESSURE
- 1. ISTS 3.6.1.4 has not been adopted since it is not applicable to Monticello. The Monticello containment analyses assume an initial drywell pressure of 2.0 psig.
However, if the pressure in the drywell were to reach this pressure, the unit will automatically shutdown because the Reactor Protection System Drywell Pressure - High Allowable Value is < 2.0 psig. Therefore, ISTS 3.6.1.4 is not needed to ensure drywell pressure is within limit because ITS 3.3.1.1, Reactor Protection System
.(RPS) Instrumentation, will ensure the drywell pressure remains below 2.0 psig. This is also consistent with current licensing basis since Monticello does not include a specific requirement to maintain drywell pressure within a specified limit.
Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 384 of 431
Attachment 1, Volume 11, Rev. 0, Page 385 of 431 ISTS 3.6.1.4 Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 385 of 431
Attachment 1, Volume 11, Rev. 0, Page 386 of 431 Drywell Pressure B 3.6.1.4 B 3.6 CONTAI MENT SYSTEMS B 3.6.1.4 D ell Pressure BASES BACKGROUND The drywell pressur is limited during normal oper tions to preserve the initial conditions ass med in the accident analysis for a Design Basis Accident (DBA) or I ss of coolant accident (LOCA . APPLICABL Primary containme performance is evaluated fo the entire spectrum of SAFETY break sizes for post lated LOCAs (Ref. 1). Among the inputs to the DBA ANALYSES is the initial primary containment internal pressur (Ref. 1). Analyses assume an initial dr ell pressure of [0.75 psig]. This limitation ensures that the safety anal sis remains valid by maintai ing the expected initial conditions and ens res that the peak LOCA d ell internal pressure does not exceed th maximum allowable of [62] sig. The maximum cal lated drywell pressure occu during the reactor blowdown phase the DBA, which assumes a instantaneous recirculation line b eak. The calculated peak dIelI pressure for this limiting event is [5 .5] psig (Ref. 1). 0D Drywell pressure atisfies Criterion 2 of 10 CF 50.36(c)(2)(ii). In the event of a BA, with an initial drywell pr sure s [0.75 psig], the resultant peak d rwell accident pressure will be maintained below the drywell design pr ssure. { In MODES 1,2, 2nd 3, a DBA could cause a release of radioactive material to primafy containment. In MODES 4jand 5, the probability and consequences of these events are reduced duie to the pressure and temperature limitations of these MODES. Therefore, maintaining drywell pressure within I mits is not required in MODE!4 or 5. A.1 With drywell pr' sure not within the limit of thLCO, drywell pressure must be restore within 1 hour. The Require Action is necessary to return operatio to within the bounds of the p imary containment analysis. The 1 hour Co pletion Time is consistent wih the ACTIONS of LCO 3.6.1.1, Frimary Containment," which quires that primary containment b restored to OPERABLE stat s within 1 hour.
- ,
BWR/41STS I B 3.6.1.4-1 I Rev. 3.0, 03/31/04 AWRI4InTS B3u6 I14 R1 Attachment 1, Volume 11, Rev. 0.Page 386 of 431
Attachment 1, Volume 11, Rev. 0, Page 387 of 431
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I Drywell Pressure BASES I ACTIONS (co inued) B 3.6.1.4 B.1 and B.2 If drywell pressure ca not be restored to within limi within the required Completion Time, th plant must be brought to a DE in which the LCO does not apply. To a hieve this status, the plant ust be brought to at least MODE 3 within 12 hours and to MODE 4 wit in 36 hours. The allowed Completion imes are reasonable, based n operating experience, to reach the required plant conditions rom full power conditions in an ordrly manner and without chal ing plant systems. SURVEILLA CE SR 3.6.1.4.1 REQUIREM TS Verifying that dryweI pressure is within limit ensu es that unit operation remains within the i it assumed in the primary c ntainment analysis. The 12 hour Frequ ncy of this SR was develope based on operating experience related o trending of drywell pressur variations during the applicable MODES Furthermore, the 12 hour F quency is considered adequate in view o other indications available in the control room, including alarms, t alert the operator to an abn rmal drywell pressure condition. '0 REFEREN4ES 1. FSAR, Sectio [6.2]. BWR/4 TS B 3.6.1.4-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 387 of 431
Attachment 1, Volume 11, Rev. 0, Page 388 of 431 JUSTIFICATION FOR DEVIATIONS ISTS 3.6.1.4 BASES, DRYWELL PRESSURE
- 1. Changes are made to be consistent with changes made to the Specification.
Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 388 of 431
Attachment 1, Volume 11, Rev. 0, Page 389 of 431 ISTS 3.6.1.9, Main Steam Isolation Valve (MSIV) Leakage Control System (LCS) Attachment 1, Volume 11, Rev. 0, Page 389 of 431
Attachment 1, Volume 11, Rev. 0, Page 390 of 431 ISTS 3.6.1.9 Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 390 of 431
Attachment 1, Volume 11, Rev. 0, Page 391 of 431
, i ,. r . -
MSIV LCS 3.6.1.9 3.6 CONTAIN ENT SYSTEMS 3.6.1.9 Ma n Steam Isolation Valve (M V) Leakage Control System LCS) LCO 3.6.1.9 Two MSIV LCS subs stems shall be OPERABLE. APPLICABILI MODES 1, 2, and 3. ACTIONS CO DITION EQUIRED ACTION COMPLETION TIME A. One M IV LCS A.1 estore MSIV LCS 30 days subsy em inoperable. ubsystem to OPERABLE B. Two iIV LCS B.1 Restore one MSIV LCS 7 days 0 subsy tems inoperable. subsystem to OPERABLE l [ status. C. Requ red Action and C.1 Be in MODE 3. 12 hours asso iated Completion Time not met. AND C.2 *Be in MODE 4. 36 hours SURVEI NCE REQUIREMENTS l SURVEI LANCE FREQUENCY SR 3. .1.9.1 Operate each M IV LCS blower 2 [15] minute . 31 days SR 3. .1.9.2 Verify electrical ontinuity of each inboard MS V 31 days LCS subsystem eater element circuitry. BWR/4/STS I 3.6.1.9-1 I Attachment 1, Volume 11, Rev. 0, Page 391 of 431 Rev. 3.0, 03/31/04
, Volume 11, Rev. 0, Page 392 of 431 MSIV LCS 3.6.1.9 REQUIREMENTS (contin ed) -. l SURVEILLAN l FREQUENCY Perform a system func onal test of each MSIV LCS [18] months subsystem.
A' 0 Attachment 1, Volume 11, Rev. 0, Page 392 of 431
Attachment 1, Volume 11, Rev. 0, Page 393 of 431 JUSTIFICATION FOR DEVIATIONS ISTS 3.6.1.9, MAIN STEAM ISOLATION VALVE (MSIV) LEAKAGE CONTROL SYSTEM (LCS)
- 1. ISTS 3.6.1.9 has not been adopted since the Monticello design does not include a Main Steam Isolation Valve (MSIV) Leakage Control System (LCS). Therefore, this Specification has not been adopted.
Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 393 of 431
Attachment 1, Volume 11, Rev. 0, Page 394 of 431 ISTS 3.6.1.9 Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 394 of 431
Attachment 1, Volume 11, Rev. 0, Page 395 of 431
.i -!I -
MSIV LCS B 3.6.1.9 B 3.6 CONTAI MENT SYSTEMS B 3.6.1.9 Mai Steam Isolation Valve (MSI ) Leakage Control System ( CS) BASES BACKGROU 0 The MSIV LCS supp ements the isolation function of the MSIVs by processing the fissio products that could leakthr ugh the closed MSIVs after a Design Basis Accident (DBA) loss of coola t accident (LOCA). The MSIV LCS con ists of two independent subs stems: an inboard subsystem, conne d between the inboard and utboard MSIVs, and an outboard subsyste, connected immediately do stream of the outboard MSIVs. Each subs stem is capable of processi leakage from MSIVs following a DBA L A. Each subsystem consis of blowers (one blower for the inboard sub ystem and two blowers for th outboard subsystem), valves, piping, and eaters (for the inboard subs stem only). Four electric heaters in e inboard subsystem are pr vided to boil off any condensate prior t the gas mixture passing thr gh the flow limiter. Each subsystem o erates in two process mode : depressurization and bleedoff. The dep essurization process reduce the steam line pressure 0 to within the oper ing capability of equipment ed for the bleedoff mode. During ble doff (long term leakage cont 1), the blowers maintain a negative pressu e in the main steam lines (R f. 1). This ensures the leakage through t e closed MSIVs is collected nd processed by the MSIV LCS. In bosh process modes, the efflue is discharged to the secondary containment and ultimately filtered y the Standby Gas Treatment (SGT) System. The MSIV LCS i manually initiated approxim ely 20 minutes following a DBA LOCA (Ref 2). APPLICA LE The MSIV LCS itigates the consequences o a DBA LOCA by ensuring SAFETY that fission prod cts that may leak from the cl sed MSIVs are diverted to ANALYSS the secondary c ntainment and ultimately filt red by the SGT System. The operation o the MSIV LCS prevents a re ease of untreated leakage for this type of ent. The MSIV LCS atisfies Criterion 3 of 10 CF 50.36(c)(2)(ii). LCO One MSIV LC subsystem can provide the r quired processing of the MSIV leakage. To ensure that this capabilit is available, assuming worst case single fail re, two MSIV LCS subsystems must be OPERABLE. I B 3.6.1.9-1 Rev. 3.0, 03/31/04 BWR/4ISTS I Attachment 1, Volume 11, Rev. 0, Page 395 of 431
Attachment 1, Volume 11, Rev. 0, Page 396 of 431
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I I I MSIV LCS B 3.6.1.9 BASES APPLICABILI In MODES 1, 2, and :, a DBA could lead to a fissioji product release to primary containment.l Therefore, MSIV LCS OPE R)ABILITY is required during these MODEt. In MODES 4 and 5,the pro ability and consequences of the e events are reduced due to he pressure and temperature limitatiors in these MODES. Therefo e, maintaining the MSIV LCS OPERABE is not required in MODE 4 or 5 to ensure MSIV leakage is processe. ACTIONS A.1 With one MSIV LCS subsystem inoperable, the in perable MSIV LCS subsystem must be estored to OPERABLE statu within 30 days. In this Condition, the rema ning OPERABLE MSIV LCS ubsystem is adequate to perform the requ ed leakage control function. However, the overall reliability is reduce because a single failure in t e remaining subsystem could result in a tot I loss of MSIV leakage contr I function. The 30 day Completion Time i based on the redundant cap bility afforded by the remaining OPERA LE MSIV LCS subsystem an the low probability of a DBA LOCA occurri g during this period. 8.1 0D With two MSIV L subsystems inoperable, at east one subsystem must be restored to OP RABLE status within 7 days The 7 day Completion Time is based on he low probability of the occ rrence of a DBA LOCA. C.1 and C.2 If the MSIV LCS ubsystem cannot be restore to OPERABLE status within the requir d Completion Time, the plant must be brought to a MODE in which te LCO does not apply. To chieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Tim s are reasonable, based on operating ex erience, to reach the require plant conditions from full power condition in an orderly manner and wi hout challenging plant systems. SURVEI LANCE SR 3.6.1.9.1 REQUIR MENTS Each MSIV LOC blower is operated for 2 [15 minutes to verify OPERABILITY The 31 day Frequency was eveloped considering the known reliabilit of the LCS blower and cont Is, the two subsystem redundancy, ad the low probability of a sig ificant degradation of the MSIV LCS su ystems occurring between surveillances and has been shown to be a ceptable through operating e perience. BWR/41STS B 3.6.1.9-2 I Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 396 of 431
Attachment 1, Volume 11, Rev. 0, Page 397 of 431 MSIV LCS B 3.6.1.9 BASES SURVEILLAN E REQUIREMENTS (conti ued) SR 3.6.1.9.2 The electrical contin ity of each inboard MSIV LC subsystem heater is verified by a resistan e check, by verifying that th rate of temperature increase meets spe fications, or by verifying that he current or wattage draw meets specific tions. The 31 day Frequenc is based on operating experience that has hown that these component usually pass this Surveillance when p rformed at this Frequency. SR 3.6.1.9.3 A system functiona test is performed to ensure t at the MSIV LCS will operate through its perating sequence. This inn udes verifying that the automatic positioni g of the valves and the oper tion of each interlock and timer are corr t, that the blowers start and evelop the required flow rate and the neces ary vacuum, and that the up ream heaters meet current or wattage raw requirements (if not use to verify electrical continuity in SR 3. .1.9.2). The [18] month Fre uency is based on the need to perform t-s Surveillance under the con itions that apply during a plant outage and the potential for an unplanned ransient if the Surveillance werelperformed with the reactor at power. Operating experience has shown that these components sually pass the Surveillance wheq performed at the [18] month Frequency. Therefore, the Frequency w s concluded to be acceptabl from a reliability standpoint. l REFERE CES 1. FSAR, Secti n [6.5].
- 2. Regulatory uide 1.96, Revision [1].
BWR/4 STS B 3.6.1.9-3 -Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 397 of 431
Attachment 1, Volume 11, Rev. 0, Page 398 of 431 JUSTIFICATION FOR DEVIATIONS ISTS 3.6.1.9 BASES, MAIN STEAM ISOLATION VALVE (MSIV) LEAKAGE CONTROL SYSTEM (LCS)
- 1. Changes are made to be consistent with changes made to the Specification.
Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 398 of 431
Attachment 1, Volume 11, Rev. 0, Page 399 of 431 ISTS 3.6.2.5, Drywell-to-Suppression Chamber Differential Pressure Attachment 1, Volume 11, Rev. 0, Page 399 of 431
Attachment 1, Volume II, Rev. 0, Page 400 of 431 ISTS 3.6.2.5 Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 400 of 431
Attachment 1, Volume 11, Rev. 0, Page 401 of 431
^. .. .
D ell-to-Suppression Chamb r Differential Pressure 3.6.2.5 3.6 CONTAIN ENT SYSTEMS 3.6.2.5 D ell-to-Suppression Chamb r Differential Pressure LCO 3.6.2.5 The drywell pressur shall be maintained 2 [1.5] p id above the pressure of the suppression c amber. APPLICABILIl: MODE 1 during the ime period:
- a. From [24] hour after THERMAL POWER is [15]% RTP following startup, to
- b. [24] hours prio to reducing THERMAL PO ER to < [15]% RTP prior to the next sc duled reactor shutdown.
ACTIONS C NDITION REQUIRED ACTION COMPLETION TIME 0 A. Dryw 1i-to-suppression A.1 Restore differential 8 hours cham er differential pressure to within limit. press re not within limit. B. Requ red Action and B.1 Reduce THERMAL 12 hours asso iated Completion POWER to * [15]% RTP. Time not met. SURVEI ANCE REQUIREMENQTS lSURVEI NCE FREQUENCY SR 3. .2.5.1 Verify drywell-to- uppression chamber differe ial 12 hours l ~pressure is withif limit.l BWRI4 STS 3.6.2.5-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 401 of 431
Attachment 1, Volume 11, Rev. 0, Page 402 of 431 JUSTIFICATION FOR DEVIATIONS ISTS 3.6.2.5, DRYWELL-TO-SUPPRESSION CHAMBER DIFFERENTIAL PRESSURE
- 1. ISTS 3.6.2.5 has not been adopted since it is not applicable to Monticello. The Monticello containment analyses for a DBA LOCA do not assume a drywell-to-suppression chamber differential pressure to reduce the hydrodynamic loads on the torus during a LOCA blowdown. Therefore, ISTS 3.6.2.5 is not needed to ensure a drywell-to-suppression chamber differential pressure limit. This is consistent with the current Monticello licensing basis.
Monticello Page 1 of I Attachment 1, Volume 11, Rev. 0, Page 402 of 431
Attachment 1, Volume 11, Rev. 0, Page 403 of 431 ISTS 3.6.2.5 Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 403 of 431
Attachment 1, Volume 11, Rev. 0, Page 404 of 431 D eli-to-Suppression Chamb r Differential Pressure B 3.6.2.5 B 3.6 CONTAI MENT SYSTEMS B 3.6.2.5 D ell-to-Suppression Chambe Differential Pressure BASES BACKGROU D The toroidal shaped uppression chamber, which ontains the suppression pool, is onnected to the drywell (pa of the primary containment) by [eigt] main vent pipes. The mai vent pipes exhaust into a continuous vet header, from which [96] do ncomer pipes extend into the suppressio pool. The pipe exit is [4] ft b low the minimum suppression pool w ter level required by LCO 3. .2.2, "Suppression Pool Water Level." Duri g a loss of coolant accident ( OCA), the increasing drywell pressure wi force the waterleg in the do ncomer pipes into the suppression pool a substantial velocities as the 'blowdown" phase of the event begins.. The ength of the waterleg has a s gnificant effect on the resultant primary c ntainment pressures and loa s. APPLICAB The purpose of m intaining the drywell at a slig ly higher pressure with SAFETY respect to the sup ression chamber is to minimi e the drywell pressure increase necessa to clear the downcomer pip s to commence ANALYSES condensation of s am in the suppression pool nd to minimize the mass 0 of the accelerated water leg. This reduces the ydrodynamic loads on the torus during t e LOCA blowdown. The req red differential pressure results in a down omer waterleg of [3.06 to 3.5 ft. Initial drywell-to-s ppression chamber different I pressure affects both the dynamic pool loads on the suppression ch mber and the peak drywell pressure during owncomer pipe clearing duri g a Design Basis Accident LOCA. Drywell-t -suppression chamber differ ntial pressure must be maintained withi the specified limits so that t safety analysis remains valid. Drywell-to-suppr ssion chamber differential p ssure satisfies Criterion 2 of 10 CFR 50.3 (c)(2)(ii). LCO A drywell-to-su ression chamber differential pressure limit of [1.5] psid is required to ens re that the containment cond ions assumed in the safety analyses are met. A drywell-to-suppression hamber differential pressure of < [1.5] psid c rresponds to a downcomer ater leg of > [3.58] ft. Failure to main in the required differential pessure could result in excessive forc s on the suppression chamber due to higher water clearing loads om downcomer vents and hi her pressure buildup in the drywell. BWR/4ISTS l B 3.6.2.5-1 I Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 404 of 431
Attachment 1, Volume 11, Rev. 0, Page 405 of 431 I l well-to-Suppression Chambe Differential Pressure l/l C / B 3.6.2.5 BASES APPLICABILI Drywell-to-suppressi chamber differential pressu e must be controlled when the primary co ainment is inert. The prima containment must be inert in MODE 1, sinc this is the condition with the highest probability for an event that could p oduce hydrogen. It is also th condition with the highest probability of n event that could impose I rge loads on the primary containment Inerting primary con inment is an operational pro lem because it prevents primary co tainment access without an ppropriate breathing apparatus. Therefoo, the primary containment is inerted as late as possible in the unit artup and is de-inerted as s n as possible in the unit shutdown. As I ng as reactor power is < [15] /o RTP, the probability of an event that ge rates hydrogen or excessiv loads on primary containment occurri g within the first [24] hours f Ilowing a startup or within the last [24] ours prior to a shutdown is I enough that these "windows," with the primary containment not ie ed, are also justified. The [24] hour time eriod is a reasonable amoun time to allow plant personnel to perfor inerting or de-inerting. ACTIONS A.1 0D If drywell-to-suppr ssion chamber differential pr ssure is not within the limit, the condition assumed in the safety anal es are not met and the differential pressue must be restored to within e limit within 8 hours. The 8 hour Compl tion Time provides sufficient time to restore differential pressure to within limit and takes into account t e low probability of an event that would reate excessive suppression hamber loads occurring during this time p riod. B.1 If the differential ressure cannot be restoredl within limits within the associated Corn letion Time, the plant must b placed in a MODE in which the LCO es not apply. This Is done reducing power to s [15]% RTP wit in 12 hours. The 12 hour C mpletion Time is reasonable, basbd on operating experience, t reduce reactor power from full power condi ions in an orderly manner an without challenging plant systems. BWR4I4STS l B 3.6.2.5-2 Rev. 3.0, 03/31/04 I Attachment 1, Volume 11, Rev. 0, Page 405 of 431
Attachment 1, Volume 11, Rev. 0, Page 406 of 431 I D ell-to-Suppression Chambe Differential Pressure lbe lB 3.6.2.5 BASES SURVEILLAN E SR 3.6.2.5.1 REQUIREMEJ S The drywell-to-suppr ssion chamber differential pr ssure is regularly monitored to ensure that the required limits are sat fied. The 12 hour
- Frequency of this SR was developed based on op rating experience relative to differential pressure variations and pres ure instrument drift during applicable M DES and by assessing the pAximity to the specified LCO differential pres ure limit. Furthermore, the 1 hour Frequency is
- considered adequat in view of other indications a ailable in the control 4 room, including alar s, to alert the operator to an bnormal pressure condition.
REFERENCES None. 0 BWRI4 TS B 3.6.2.5-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 406 of 431
Attachment 1, Volume 11, Rev. 0, Page 407 of 431 JUSTIFICATION FOR DEVIATIONS ISTS 3.6.2.5 BASES, DRYWELL-TO-SUPPRESSION CHAMBER DIFFERENTIAL PRESSURE
- 1. Changes are made to be consistent with changes made to the Specification.
- I Monticello Page1 of 1 Attachment 1, Volume 11, Rev. 0, Page 407 of 431
, Volume 11, Rev. 0, Page 408 of 431 ISTS 3.6.3.1, Drywell Cooling System Fans Attachment 1, Volume 11, Rev. 0, Page 408 of 431
Attachment 1, Volume 11, Rev. 0, Page 409 of 431 ISTS 3.6.3.1 Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 409 of 431
Attachment 1, Volume 11, Rev. 0, Page 410 of 431 [Drywell ooling System Fans] 3.6.3.1 3.6 CONTAIN ENT SYSTEMS 3.6.3.1 [D0 el Cooling System Fans] LCO 3.6.3.1 Two [drywell cooling ystem fans] shall be OPE A LE. APPLICABILI MODES 1 and 2. ACTIONS _ _ _ _ _ CO DITION EQUIRED ACTION COMPLETION TIME A. One [r quired] [drywell A.1 estore [required] [drywell 30 days coolin system fan] cooling system fan] to inoper ble. OPERABLE status. B. Two [r quired] [drywell B.1 Verify by administrative 1 hour 0 coolin system fans] means that the hydrogen inope ble. control function is AND maintained. Once per 12 hours thereafter AND B.2 Restore one [required] 7 days [drywell cooling system fan to OPERABLE status. C. Req ired Action and C.1 Be in MODE 3. 12 hours ass ciated Completion Tim not met. BWR/4 STS 3.6.3.1-1 Rev. 3.0, 03131/04 Attachment 1, Volume 11, Rev. 0, Page 410 of 431
Attachment 1, Volume 11, Rev. 0, Page 411 of 431 [Drywell ooling System Fans] ...... -..- --- SURVEILLAN E REQUIREMENTS SURVEILLAN E I I ._I [M FREQUENCY 3.6.3.1 SR 3.6.3.1.1 Operate each [require ] [drywell cooling system fan 92 days for > [15] minutes. SR 3.6.3. 1. [ Verify each [require' [drywell cooling system fan] [18] months ] flow rate is 2 [500] sc'. 0D BWRI4 STS 3.6.3.1-2 Rev. 3.0, 03/31/04 Aftachment 1; Volume 11, Rev. 0, Page 411 of 431
Attachment 1, Volume 11, Rev. 0, Page 412 of 431 JUSTIFICATION FOR DEVIATIONS ISTS 3.6.3.1, DRYWELL COOLING SYSTEM FANS
- 1. ISTS 3.6.3.1 has not been adopted since it is not applicable to Monticello. The Monticello containment analyses for a DBA LOCA do not assume drywell cooling system fans are used, post-LOCA, to perform mixing t6 minimize the potential for local hydrogen burns. Therefore, ISTS 3.6.3.1 is not needed to ensure drywell atmosphere mixing. This is consistent with the current Monticello licensing basis.
Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 412 of 431
Attachment 1, Volume 11, Rev. 0, Page 413 of 431 ISTS 3.6.3.1 Bases Markup and Justification for Deviations (JFDs)- Attachment 1, Volume 11, Rev. 0, Page 413 of 431
Attachment 1, Volume 11, Rev. 0, Page 414 of 431 [Drywell ooling System Fans] B 3.6.3.1 B 3.6 CONTAI MENT SYSTEMS B 3.6.3.1 [D ell Cooling System Fans] BASES BACKGROU D The [Drywell Coolin System fans] ensure a unifo ly mixed post accident primary co tainment atmosphere, thereb minimizing the potential for local hy rogen burns due to a pocket of hydrogen above the flammable concentr tion. The [Drywell Coolin System fans] are an Engine red Safety Feature and are designed to wit stand a loss of coolant accid nt (LOCA) in post accident environme ts without loss of function. he system has two independent subsy tems consisting of fans, fan oil units, motors, controls, and ducti . Each subsystem is sized circulate [500] scfm. The [Drywell Cooli g System fans] employ both orced circulation and natural circulation ensure the proper mixing o hydrogen in primary containment. The ecirculation fans provide the rced circulation to mix hydrogen while th fan coils provide the natural rculation by increasing the density throug the cooling of the hot gases t the top of the drywell causing the coole gases to gravitate to the bo om of the drywell. The 0 two subsystems a e initiated manually since fla mability limits would not be reached until s veral days after a LOCA. E ch subsystem is powered from a separate e ergency power supply. Sin e each subsystem can provide 100% oft e mixing requirements, the stem will provide its design function w th a worst case single activeeilure. The [Drywell Coo ing System fans] use the D ell Cooling System recirculating fans to mix the drywell atmospher. The fan coil units and recirculation fan are automatically disengage during a LOCA but may be restored to s ice manually by the operat r. In the event of a loss of offsite power, all an coil units, recirculating fa s, and primary containment wat r chillers are transferred to t eemergency diesels. The fan coil units an recirculating fans are starte automatically from diesel power upon los of offsite power. APPLIC LE The [Drywell C ling System fans] provide th capability for reducing the SAFETY local hydrogen oncentration to approximatel the bulk average ANALYS S concentration f Ilowing a Design Basis Accid nt (DBA). The limiting DBA relative to hydr gen generation is a LOCA. Hydrogen may ccumulate in primary contai ment following a LOCA as a result of:
- a. A metal st am reaction between the zir onium fuel rod cladding and the reacto, coolant or BWR/41STS I B 3.6.3.1-1 I Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 414 of 431
Attachment 1, Volume 11, Rev. 0, Page 415 of 431 [Drywell Cooling System Fans] B 3.6.3.1 BASES APPLICABLE AFETY ANALYSIS (contin ed)
- b. Radiolytic deco position of water in the Reac or Coolant System.
To evaluate the pote tial for hydrogen accumulati n in primary containment followina LOCA, the hydrogen gen ration as a function of time following the i jation of the accident is calc ated. Conservative assumptions recoinended by Reference I are u ed to maximize the amount of hydroge calculated. The Reference 2 culations show that hydroge assumed to be released to the d elI within 2 minutes following a DBA LOCA raises drywell hydrogen c ncentration to over 2.5 volu e percent (vlo). Natural circulation phenom na result in a gradient conce tration difference of less then 0.5 v/o in the rywell and less than 0.1 v/o i the suppression chamber. Even th ugh this gradient is acceptably small and no credit for mechanical mixing was assumed in the analysis two [Drywell Cooling System fans] are [equired] to be OPERABLE (t pically four to six fans are required to ke p the drywell cool during ope ation in MODE 1 or 2) by this LCO. 0 The [Drywell Cool ng System fans] satisfy Crite ion 3 of 10 CFR 50.36(c)( )(ii). l LCO Two [Drywell Co ing System fans] must be O ERABLE to ensure operation of at lest one fan in the event of a rst case single active failure. Each oft ese fans must be powered f m an independent safety related bus. Operation with a least one fan provides the c pability of controlling the bulk hydrogen c ncentration in primary conta ment without exceeding the flammability mit. APPLICA ILITY In MODES I an 2, the two [Drywell Cooling ystem fans] ensure the capability to pre ent localized hydrogen conc ntrations above the flammability limi of 4.0 v/o in drywell, assumi g a worst case single active failure. In MODE 3, bo the hydrogen production ra and the total hydrogen produced after LOCA would be less than t at calculated for the DBA LOCA. Also, b cause of the limited time in t is MODE, the probability of an accident re uiring the [Drywell Cooling S stem fans] is low. Therefore, the Drywell Cooling System fans are not required in MODE 3. BWR/41STS B 3.6.3.1-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 415 of 431
Attachment 1, Volume 11, Rev. 0, Page 416 of 431 I I [Drywell ooling System Fans] yf B 3.6.3.1 BASES APPLICABILI (continued) In MODES 4 and 5, t e probability and consequenies of a LOCA are reduced due to the p essure and temperature limitations in these MODES. Therefore, he [Drywell Cooling System ans] are not required in these MODES. ACTIONS A.1 With one [required] rywell Cooling System fan] i operable, the inoperable fan must be restored to OPERABLE s tus within 30 days. In this Condition, the r maining OPERABLE fan is a equate to perform the hydrogen mixing fu ction. However, the overall r liability is reduced because a single fa lure in the OPERABLE fan could result in reduced hydrogen mixing ca ability. The 30 day Completion Time is based on the availability of the s cond fan, the low probability f the occurrence of a LOCA that would g nerate hydrogen in amounts capable of exceeding the flammability liit, the amount of time availabl after the event for operator action to Orevent exceeding this limit, a d the availability of the Containment Atmosphere Dilution System. 0 B.1 and B.2
-- -REVIEWER'S NOTE This Condition is nly allowed for units with an Iternate hydrogen control system acceptabl to the technical staff.
With two [Drywell Cooling System fans] inoper ble, the ability to perform the hydrogen co trol function via alternate cap bilities must be verified by administrative m ans within 1 hour. The alter ate hydrogen control capabilities are ovided by the [Primary Cont inment Inerting System or one subsystem f the Containment Atmosphe e Dilution System]. The 1 hour Completi n Time allows a reasonable eriod of time to verify that a loss of hydroge control function does not exi t.
---- REVIEWER'S NO E - ------ _
The following is to be used if a non-Technical Specification alternate hydrogen contr Ifunction is used to justify thi Condition: In addition, the alternate hydro en control system capability ust be verified once per 12 hours there fter to ensure its continued a ailability. BWR/4ISTS I I I B 3.6.3.1-3 Attachment 1, Volume 11, Rev. 0, Page 416 of 431 I Rev. 3.0, 03/31/04
Attachment 1,Volume 11, Rev. 0, Page 417 of 431 I [Drywell Cooling System Fans] B 3.6.3.1 BASES ACTIONS (co tinued) [Both] the [initial] veri cation [and all subsequent v rifications] may be performed as an adm nistrative check by examininlogs or other information to deter ne the availability of the alter ate hydrogen control system. It does not ean to perform the Surveilla ces needed to demonstrate OPER ILITY of the alternate hydroen control system. If the ability to perform the hydrogen control function is maintained, continued operation s permitted with two [Drywell ooling System fans] inoperable for up to days. Seven days is a reasnable time to allow two [Drywell Cooling Sy tem fans] to be inoperable b cause the hydrogen control function is m intained and because of the w probability of the occurrence of a LO A that would generate hydro en in amounts capable of exceeding the fla nmability limit. CA1 If any Required Ac on and associated Completi nTime cannot be met, the plant must be brought to a MODE in which t e LCO does not apply. To achieve this status, the plant must be brough to at least MODE 3 f 12 hours is reasonable, 0 within 12 hours. The allowed Completion Time based on operatin experience, to reach MODE 3 from full power conditions in an o erly manner and without ch lenging plant systems. SURVEIL NCE SR 3.6.3. 1.1 REQUIRE ENTS Operating each [r quired] [Drywell Cooling Sys em fan] for 2 15 minutes ensures that eac subsystem is OPERABLE a d that all associated controls are funct oning properly. It also ensur s that blockage, fan or motor failure, or xcessive vibration can be de cted for corrective action. The 92 day Freq ency is consistent with the In ervice Testing Program Frequencies, op rating experience, the known reliability of the fan motors and controls, an the two redundant fans avail ble. BWR/4 gTS l B 3.6.3.1-4 I Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 417 of 431
Attachment 1, Volume 11, Rev. 0, Page 418 of 431
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[Drywell ooling System Fans] [/ B 3.6.3.1 BASES SURVEILLAN E REQUIREMENTS (conti ued) [SR 3.6.3.1.2l Verifying that each [r quired] [Drywell Cooling Syslem fan] flow rate is d2 [500] scfm ensure that each fan is capable of m intaining localized hydrogen concentratons below the flammability Ii it. The [18] month Frequency is based n the need to perform this S rveillance under the conditions that appl during a plant outage and th potential for an unplanned transient f the Surveillance were perfo med with the reactor at power. Operating e perience has shown these armponents usually pass the Surveillance whn performed at the [18] mont Frequency. Therefore, the Freq ency was concluded to be a ceptable from a reliability standpoinc ] REFERENC S 1. Regulatory Gu e 1.7, Revision [1].
- 2. FSAR, Sectio [6.2.5].
0 BWR/4 TS B 3.6.3.1-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 418 of 431
Attachment 1, Volume 11, Rev. 0, Page 419 of 431 JUSTIFICATION FOR DEVIATIONS ISTS 3.6.3.1 BASES, DRYWELL COOLING SYSTEM FANS
- 1. Changes are made to be consistent with changes made to the Specification.
Monticello . Page 1 of I Attachment 1, Volume 11, Rev. 0, Page 419 of 431
Attachment 1, Volume 11, Rev. 0, Page 420 of 431 ISTS 3.6.3.3, Containment Atmosphere Dilution (CAD) System Attachment 1, Volume 11, Rev. 0, Page 420 of 431
Attachment 1, Volume 11, Rev. 0, Page 421 of 431 ISTS 3.6.3.3 Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 421 of 431
Attachment 1, Volume 11, Rev. 0, Page 422 of 431 I CAD System 3.6.3.3 3.6 CONTAIN ENT SYSTEMS 3.6.3.3 Co tainment Atmosphere Diluti n (CAD) System LCO 3.6.3.3 Two CAD subsyste shall be OPERABLE. APPLICABILIY: MODES 1 and 2. ACTIONS CO DITION EQUIRED ACTION COMPLETION TIME A. One C D subsystem A.1 estore CAD subsystem to 30 days inoper ble. OPERABLE status. 1 hour B. [Two AD subsystems inoper ble. B.1 Verify by administrative means that the hydrogen 0D control function is AND maintained. Once per 12 hours thereafter AND B.2 Restore one CAD 7 days] subsystem to OPERABLE status. C. Req ired Action and C.1 Be in MODE 3. 12 hours ass ciated Completion Ti not met. BWRI4 TS 3.6.3.3-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 422 of 431
Attachment 1, Volume 11, Rev. 0, Page 423 of 431
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CAD System S IOMIMII I AkMtLLr t~~l10 Of Jrv~lL.LJHI'4 RC M I.Ju~L;IcvL;I' I .i) SURVEILLAN E
.I I FREQUENCY 3.6.3.3 SR 3.6.3.3.1 Verify 2[4350] gal of li nitrogen are contained in uid 31 days the CAD System.
SR 3.6.3.3. Verify each CAD subs stem manual, power 31 days operated, and automatic valve in the flow path that is not locked, sealed, r otherwise secured in position is in the corr ct position or can be aligned to the correct position 0 BWRI4 TS 3.6.3.3-2 Rev. 3.0, 03/31/04
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Attachment 1, Volume 11, Rev. 0, Page 423 of 431
Attachment 1, Volume 11, Rev. 0, Page 424 of 431 JUSTIFICATION FOR DEVIATIONS
- ISTS 3.6.3.3, CONTAINMENT ATMOSPHERE DILUTION (CAD) SYSTEM
- 1. ISTS 3.6.3.3 has not been adopted since it is not applicable to Monticello. The Monticello containment analyses for a DBA LOCA do not assume the use of a CAD system, post-LOCA, to maintain combustible gas concentrations within the primary containment at or below flammability limits. Therefore, ISTS 3.6.3.3 is not needed to maintain combustible gas concentrations. This is consistent with the current Monticello licensing basis.
Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 424 of 431
Attachment 1, Volume 11, Rev. 0, Page 425 of 431 ISTS 3.6.3.3 Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 11, Rev. 0, Page 425 of 431
Attachment 1, Volume 11, Rev. 0, Page 426 of 431 CAD System B 3.6.3.3 B 3.6 CONTAI MENT SYSTEMS B 3.6.3.3 Con ainment Atmosphere Dilutio (CAD) System BASES BACKGROU D The CAD System fu ctions to maintain combustib gas concentrations within the primary ccntainment at or below the fla inability limits following a postulated loss of Coolant accident (LOCA) by diluting hydrogen and oxygen with nitroge . To ensure that a combustible gas mixture does not occur, oxygen conc ntration is kept < [5.0] volum percent (v/o), or hydrogen concentraion is kept < 4.0 vlo. The CAD System i manually initiated and consits of two independent, 100% capacity sub ystems. Each subsystem in ludes a liquid nitrogen supply tank, ambie t vaporizer, electric heater, a d connected piping to supply the drywell nd suppression chamber vol mes. The nitrogen storage tanks eac contain 2 [4350] gal, which i adequate for [7] days of CAD subsystem o eration. The CAD System perates in conjunction with e ergency operating procedures that a used to reduce primary con ainment pressure 0 periodically during CAD System operation. Thi combination results in a feed and bleed ap roach to maintaining hydrog n and oxygen concentrations be ow combustible levels. APPLICAB E To evaluate the p tential for hydrogen and oxy en accumulation in SAFETY primary containm nt following a LOCA, hydrog n and oxygen generation ANALYSE is calculated (as function of time following th initiation of the accident). The assumption stated in Reference 1 are us d to maximize the amount of hydrogen and xygen generated. The calclation confirms that when the mitigating sytems are actuated in accord nce with emergency operating proce res, the peak oxygen conce tration in primary containment is [5.0] v/o (Ref. 2). Hydrogen and o ygen may accumulate within primary containment following a LOC as a result of:
- a. A metal wa er reaction between the zirc ium fuel rod cladding and the reactor oolant or
- b. Radiolytic ecomposition of water in the Reactor Coolant System.
The CAD Syst m satisfies Criterion 3 of 10 FR 50.36(c)(2)(ii). BWR/4ISTS l B 3.6.3.3-1 I Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 426 of 431
Attachment 1, Volume 11, Rev. 0, Page 427 of 431
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CAD System B 3.6.3.3 BASES I LCO Two CAD subsystemj must be OPERABLE. This 4 nsures operation of at least one CAD subsy tem in the event of a worst c se single active failure. Operation of At least one CAD subsystem i designed to maintain primary containment ost-LOCA oxygen concentra ion < 5.0 vlo for 7 days. l APPLICABILI In MODES 1 and 2, t e CAD System is required t maintain the oxygen concentration within rimary containment below th flammability limit of 5.0 vto following a L CA. This ensures that the rlative leak tightness of primary containmen is adequate and prevents da age to safety related equipment and instr ments located within primary containment. In MODE 3, both th hydrogen and oxygen produ tion rates and the total amounts produced fter a LOCA would be less th n those calculated for the Design Basis A cident LOCA. Thus, if the an lysis were to be performed starting ith a LOCA in MODE 3, the t me to reach a flammable concention would be extended bey nd the time conservatively calc lated for MODES 1 and 2. T e extended time would allow hydrogen rer oval from the primary contai ment atmosphere by other means and also allow repair of an inopera le CAD subsystem, if CAD were not avai able. Therefore, the CAD S tem is not required to be OPERABLE in M DE 3. 0D In MODES 4 and ,the probability and consequ nces of a LOCA are reduced due to thpressure and temperature Ii itations of these MODES. There e, the CAD System is not re uired to be OPERABLE in MODES 4 and 5. . ACTIONS A._ If one CAD subs tem is inoperable, it must b restored to OPERABLE status within 30 ays. In this Condition, the r aining OPERABLE CAD subsystem is ad quate to perform the oxygen ontrol function. However, the overall reliab lity is reduced because a sin le failure in the OPERABLE sub ystem could result in reduce oxygen control capability. The 30 day Co pletion Time is based on the w probability of the occurrence of a OCA that would generate h rogen and oxygen in amounts capabl of exceeding the flammabili limit, the amount of time available after t e event for operator action tol prevent exceeding this limit, and the av liability of the OPERABLE C D subsystem and other hydrogen mitig ting systems. BWR/4 WTS I B 3.6.3.3-2 Rev. 3.0, 03/31/04 I Attachment 1, Volume 11, Rev. 0, Page 427 of 431
Attachment 1, Volume II, Rev. 0, Page 428 of 431 I I I CAD System B 3.6.3.3 BASES ACTIONS (co tinued I) 1.1 and B.2
/__ REVIEWER'S NOTE----
. ----- /
'his Condition is onl allowed for plants with an altrnate hydrogen ontrol system accegiable to the technical staff.
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VVith two CAD subsy tems inoperable, the ability t perform the hydrogen Control function via Iternate capabilities must be erified by administrative mean within 1 hour. The alternat hydrogen control Capabilities are prov ed by the [Primary Contain ent Inerting System or Cine hydrogen reco biner and one Drywell Cooli System fan]. The 1 hour Completion ime allows a reasonable perid of time to verify that a Ikoss of hydrogen co trol function does not exist.
------REVIEWER'S NOTE--
1'he following is to e used if a noni-Technical Sp cification alternate hydrogen control f nction is used to justify this C ndition: In addition, the aIternate hydrogen I 2 hours thereafte control system capability mu t be verified once per to ensure its continued avail bility. 0 [Both] the [initial] rification [and all subsequen verifications] may be performed as an ministrative check by exami ing logs or other information to det rmine the availability of the a ternate hydrogen control system. It does n mean to perform the Survei ances needed to demonstrate OP RABILITY of the alternate hy rogen control system. If the ability to pe rm the hydrogen control fun ion is maintained, continued operati n is permitted with two CAD ubsystems inoperable for up to 7 days. Se en days is a reasonable tim to allow two CAD subsystems to b inoperable because the hyd gen control function is maintained and 'ecause of the low probability f the occurrence of a LOCA that woul generate hydrogen in amouts capable of exceeding the flammability mit. With two CAD s bsystems inoperable, one C D subsystem must be restored to OPE BLE status within 7 days. he 7 day Completion Time is based on the ow probability of the occurre ce of a LOCA that would generate hydro en in the amounts capable o exceeding the flammability . limit, the amour t of time available after the e nt for operator action to prevent exceeding this limit, and the availabil ty of other hydrogen mitigating syst s. BWR/4 hTS l B 3.6.3.3-3 I Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 428 of 431
Attachment 1, Volume 11, Rev. 0, Page 429 of 431
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CAD System B 3.6.3.3 BASES ACTIONS (co tinuedI) C.1 If any Required Actio cannot be met within the as ociated Completion Time, the plant must be brought to a MODE in whih the LCO does not apply. To achieve t is status, the plant must be b ought to at least MODE 3 within 12 h urs. The allowed Completio Time of 12 hours is reasonable, based operating experience, to re ch MODE 3 from full power conditions in n orderly manner and witho challenging plant systems. SURVEILLA CE SR 3.6.3.3.1 REQUIREM NTS Verifying that there is 2 [4350] gal of liquid nitrogf n supply in the CAD System will ensure t least [7] days of post-LOC CAD operation. This minimum volume o liquid nitrogen allows sufficiqnt time after an accident to replenish the nit ogen supply for long term ine ing. This is verified every 31 days to e sure that the system is capa le of performing its intended function hen required. The 31 day F quency is based on operating experie ce, which has shown 31 day to be an acceptable period to verify th liquid nitrogen supply and o the availability of other 0 hydrogen mitigati g systems. SR 3.6.3.3.2 Verifying the corrct alignment for manual, po r operated, and automatic valves n each of the CAD subsyste flow paths provides assurance that the proper flow paths exist for ystem operation. This SR does not apply t valves that are locked, sealep, or otherwise secured in position, since thse valves were verified to b in the correct position prior to locking, sealin , or securing. A valve is also a lowed to be in the nonaccide t position provided it can be aligned to th accident position within the me assumed in the accident analysi. This is acceptable becaus the CAD System is manually initiat .This SR does not apply to valves that cannot be inadvertently imaligned, such as check valv s. This SR does not require any testing or v lve manipulation; rather, it in olves verification that those valves capable f being mispositioned are in he correct position. BWR/41STS l B 3.6.3.3-4 I Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 429 of 431
Attachment 1, Volume 11, Rev. 0, Page 430 of 431 I I I CAD System B 3.6.3.3 BASES SURVEILLAN E REQUIREMENTS (conti ued) The 31 day Frequen y is appropriate because the alves are operated under procedural co trol, improper valve position ould only affect a single subsystem, th probability of an event requi ng initiation of the system is low, and esystem is a manually initi d system. REFERENC S 1. Regulatory Guile 1.7, Revision E2]. l 2. FSAR, Section ]-
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0 BWR/S TS B 3 .6.3.3-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 11, Rev. 0, Page 430 of 431
Attachment 1, Volume 11, Rev. 0, Page 431 of 431 JUSTIFICATION FOR DEVIATIONS ISTS 3.6.3.3 BASES, CONTAINMENT ATMOSPHERE DILUTION (CAD) SYSTEM
- 1. Changes are made to be consistent with changes made to the Specification.
Monticello Page 1 of 1 Attachment 1, Volume 11, Rev. 0, Page 431 of 431}}