ML14125A015
| ML14125A015 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 05/01/2014 |
| From: | George Gellrich Calvert Cliffs, Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML14125A015 (326) | |
Text
ExeLon Generation.
Ad~nowGeorge Gelirich Site Vice President Calvert Cliffs Nuclear Power Plant 1650 Calvert Cliffs Parkway Lusby. MD 20657 410 495 5200 Office 717 497 3463 Mobile www.exeloncorp.com george.gellrich@exeloncorp.com May 1, 2014 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Calvert Cliffs Nuclear Power Plant Unit Nos. 1 & 2; Docket Nos. 50-317 & 50-318 Facility Operating License Nos. DPR-53 and DPR-69
Subject:
License Amendment Request: Adoption of Technical Specification Task Force Traveler (TSTF) - 425, Revision 3, Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5b
Reference:
(a) Technical Specification Task Force Traveler (TSTF) - 425, Revision 3, Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5b The Calvert Cliffs Nuclear Power Plant, LLC hereby requests an Amendment to Renewed Operating License Nos. DPR-53 and DPR-69 for Calvert Cliffs Unit Nos. 1 and 2, respectively, with the submittal of the proposed changes to the Technical Specifications to incorporate the Nuclear Regulatory Commission-approved Reference (a).
The proposed amendment would modify the Technical Specifications by relocating specific surveillance frequencies to a licensee-controlled program with the implementation of Nuclear Energy Institute 04-10, "Risk Informed Method for Control of Surveillance Frequencies."
Attachment (1) provides a description and assessment of the proposed change, the requested confirmation of applicability and plant-specific verifications. Attachment (2) provides documentation of the probabilistic risk assessment technical adequacy. Attachment (3) provides the marked-up Technical Specification pages and Attachment (4) provides the marked-up Technical Specification Bases pages. A comparison of the Improved Standard Technical Specification Surveillance Frequencies and the Calvert Cliffs Technical Specification Surveillance Frequencies is presented in Attachment (5).
Calvert Cliffs requests approval of the proposed license amendment by April 30, 2015, with the amendment being implemented within 120 days. In accordance with 10 CFR 50.91(b)(1), a copy of this application is provided to the designated Maryland Official.
There are no regulatory commitments contained in this letter.
AwlD
Document Control Desk May 1, 2014 Page 2 Should you have questions regarding this matter, please contact Mr. Douglas E. Lauver at (410) 495-5219.
I declare under penalty of perjury that the foregoing is true and correct. Executed May 1, 2014.
Respectfully, George H. Gellrich Site Vice President GHG/PSF/bjd Attachments: (1) Description and Assessment of Proposed Changes (2) PRA Technical Adequacy (3) Marked-up Technical Specification Pages (4) Marked-up Technical Specification Bases Pages (5) Comparison Matrix cc: NRC Project Manager, Calvert Cliffs NRC Resident Inspector, Calvert Cliffs NRC Regional Administrator, Region I S. Gray, MD-DNR
ATTACHMENT (1)
DESCRIPTION AND ASSESSMENT OF PROPOSED CHANGES Calvert Cliffs Nuclear Power Plant, LLC May 1, 2014
ATTACHMENT (1)
DESCRIPTION AND ASSESSMENT OF PROPOSED CHANGES
1.0 DESCRIPTION
This letter is a request for an amendment to Renewed Operating Licenses DPR-53 and DPR-69 for Calvert Cliffs Nuclear Power Plant (CCNPP), Units 1 and 2. The proposed amendment would modify the Technical Specifications by relocating specific surveillance frequencies to a licensee-controlled program with the implementation of Nuclear Energy Institute (NEI) 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk Informed Method for Control of Surveillance Frequencies". Additionally, the change would add a new program, the Surveillance Frequency Control Program, to Technical Specification (TS) Section 5, Administrative Controls.
The changes are consistent with Nuclear Regulatory Commission (NRC) approved Technical Specification Task Force (TSTF)-425, Revision 3 (ADAMS Accession No. ML080280275). The Federal Register Notice published on July 6, 2009 (74 FR 31996) announced the availability of this TS improvement as part of the consolidated line item improvement process.
2.0 ASSESSMENT 2.1 Applicability of the Published Safety Evaluation Calvert Cliffs has reviewed the NRC Safety Evaluation dated July 6, 2009 as part of the consolidated line item improvement process. This review included a review of the NRC staffs evaluation, TSTF-425, Revision 3, and the requirements specified in NEI 04-10, Revision 1 (ADAMS Accession No. ML071360456). includes CCNPP documentation with regard to probabilistic risk assessment technical adequacy consistent with the requirements of Regulatory Guide 1.200, Revision 1 (ADAMS Accession No. ML070240001), Section 4.2, and describes any probabilistic risk assessment models without NRC endorsed standards, including documentation of the quality characteristics of those models in accordance with Regulatory Guide 1.200.
Calvert Cliffs has concluded that the justifications presented in the TSTF proposal and the NRC staff's Safety Evaluation are applicable to CCNPP and justify this license amendment request for the incorporation of the changes to the Calvert Cliffs TSs.
2.2 Optional Changes and Variations Calvert Cliffs is proposing variations or deviations as described below from the applicable Technical Specification changes described in TSTF-425, Revision 3 or the applicable portions of the NRC staffs model Safety Evaluation referenced in the Federal Register (74 FR 31996).
Note that Calvert Cliffs uses different numbering and titles than the Improved Standard Technical Specifications in several instances. Only TS changes consistent with Calvert Cliffs' design and TS are included. Attachment 5 provides specific information.
After NRC approval of TSTF-425, it was recognized that surveillance frequencies that have not been changed under the Surveillance Frequency Control Program (SFCP) may not be based on operating experience, equipment reliability or plant risk. Therefore, the TSTF and the NRC agreed that the TSTF-425 TS Bases insert, "The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program," should be revised to state, "The Surveillance Frequency is controlled under the Surveillance Frequency Control Program." The existing TS Bases information will be relocated to the licensee-controlled Surveillance Frequency Control Program.
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ATTACHMENT (1)
DESCRIPTION AND ASSESSMENT OF PROPOSED CHANGES The TSTF-425 TS Section 5.5.15 insert references NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies." The Calvert Cliffs TS 5.5.19 insert references NEI 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies." This is an administrative deviation from TSTF-425 with no impact on the NRC's model safety evaluation dated July 6, 2009 (74 FR 31996).
The Comparison Matrix (Attachment 5) is provided for information and is a comparison between the NUREG-1432, "Standard Technical Specifications Combustion Engineering Plants,"
Surveillance Requirements included in TSTF-425 and the Calvert Cliffs SRs included in this license amendment request. The comparison includes a summary description of the referenced SR, which is provided for information purposes only and is not intended to be a verbatim description of the SR. The comparison matrix contains the following information:
- Calvert Cliffs SRs that have identical numbers to the corresponding NUREG-1432 SRs are not deviations from TSTF-425, with the exception of administrative deviations (if any) such as formatting and plant-specific Frequencies. These deviations are administrative with no impact on the NRC's model safety evaluation dated July 6, 2009 (74 FR 31996).
" Calvert Cliffs SRs that have different numbering than the NUREG-1432 SRs are an administrative deviation from TSTF-425 with no impact on the NRC's model safety evaluation dated July 6, 2009 (74 FR 31996).
- For NUREG-1432 SRs that are not contained in the Calvert Cliffs TS, the corresponding TSTF-425 mark-ups for the SRs are not applicable to Calvert Cliffs. This is an administrative deviation from TSTF-425 with no impact on the NRC's model safety evaluation dated July 6, 2009 (74 FR 31996).
" The following TS and associated SRs are Calvert Cliffs plant specific:
o TS 3.4.17, Special Test Exception (STE) RCS Loops - Modes 4 and 5 Calvert Cliffs has determined that these Calvert Cliffs plant specific SRs involve fixed periodic Frequencies and do not meet any of the four exclusion criteria of TSTF-425. In accordance with TSTF-425, changes to the Frequencies for SRs with periodic Frequencies that do not meet the exclusion criteria would be controlled under the SFCP.
The SFCP provides the necessary administrative controls to require the SRs related to calibration, testing and inspection are conducted at a frequency to assure that the necessary quality of systems and components is maintained, the facility operation will be within safety limits, and that the limiting conditions for operation will be met. Change to frequencies in the SFCP would be evaluated using the methodology and probabilistic risk guidelines contained in NEI-04-10. The NEI 04-10 methodology includes qualitative considerations, risk analyses, sensitivity studies and bounding analyses, as necessary and recommended monitoring of the performance of systems, structures and components for which frequencies are changed to assure that the reduced testing does not adversely impact the SSCs. In addition, NEI 04-10 methodology satisfies the five key safety principles specified in Regulatory Guide 1.177 relative to changes in SR frequencies. Relocation of these frequencies is also consistent with TSTF-425 and with the NRC's model safety evaluation, including scope exclusions identified in Section 1.0 of the model safety evaluation.
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ATTACHMENT (1)
DESCRIPTION AND ASSESSMENT OF PROPOSED CHANGES
3.0 REGULATORY ANALYSIS
3.1 No Significant Hazards Consideration Determination Calvert Cliffs has reviewed the proposed no significant hazards determination published in the Federal Register on July 6, 2009 (74 FR 31996). Calvert Cliffs has concluded that the proposed no significant hazards consideration presented in the Federal Register notice is applicable to CCNPP, Unit Nos. 1 and 2 and is provided below, which satisfies the requirements of 10 CFR 50.90(a).
This change requests the adoption of an approved change to the standard technical specifications for Combustion Engineering plants (NUREG-1432) to allow relocation of specific Technical Specification (TS) surveillance frequencies to a licensee controlled program. The proposed change is described in Technical Specification Task Force (TSTF) traveler, TSTF-425, Revision 3 (ADAMS Accession No. ML080280275) related to the Relocation of Surveillance Frequencies to Licensee Control - RITSTF Initiative 5b and was described in the Notice of Availability published in the Federal Register on July 9, 2009 (74 FR 31996).
The proposed changes are consistent with Nuclear Regulatory Commission (NRC)-approved TSTF traveler, TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control -
RITSTF Initiative 5b." The proposed change relocates surveillance frequencies to a licensee controlled program, the Surveillance Frequency Control Program. This change is applicable to licensees using probabilistic risk guidelines contained in NRC-approved Nuclear Energy Institute (NEI) 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk Informed Method for Control of Surveillance Frequencies," (ADAMS Accession No. ML071360456).
Calvert Cliffs has evaluated the proposed changes to the Technical Specifications using the criteria in 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration, in that operation of the facility in accordance with the proposed amendment would not:
- i. Involve a significant increase in the probability or consequences of an accident previously evaluated; or Response: No.
The proposed change relocates the specified frequencies for periodic surveillance requirements to licensee control under a new Surveillance Frequency Control Program.
Surveillance frequencies are not an initiator to any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased.
The systems and components required by the Technical Specifications for which the surveillance frequencies are relocated are still required to be operable, meet the acceptance criteria for the surveillance requirements, and be capable of performing any mitigation function assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly increased.
Therefore, this proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
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ATTACHMENT (1)
DESCRIPTION AND ASSESSMENT OF PROPOSED CHANGES ii. Create the possibility of a new or different type of accident from any accident previously evaluated; or Response: No.
No new or different accidents result from utilizing the proposed change. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice.
Therefore, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.
iii. Involve a significant reduction in a margin of safety.
Response: No.
The design, operation, testing methods, and acceptance criteria for systems, structures and components specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the updated final safety analysis report and the bases to the TS), since these are not affected by changes to the surveillance frequencies. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. To evaluate a change in the relocated surveillance frequency, Calvert Cliffs will perform a probabilistic risk evaluation using the guidance contained in NRC approved NEI 04-10, Revision 1 in accordance with the TS Surveillance Frequency Control Program. Nuclear Energy Institute 04-10, Revision 1 methodology provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies consistent with Regulatory Guide 1.177.
Therefore, this proposed change does not involve a significant reduction in a margin of safety.
Based upon the reasoning presented above, Calvert Cliffs concludes that the requested change involves no significant hazards consideration, as set forth in 10 CFR 50.92(c).
4.0 ENVIRONMENTAL CONSIDERATION
The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR Part 20, and would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.
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ATTACHMENT (2)
PRA TECHNICAL ADEQUACY Calvert Cliffs Nuclear Power Plant, LLC May 1, 2014
ATTACHMENT (2)
PRA TECHNICAL ADEQUACY The Calvert Cliffs PRA Quality Statement in support of TSTF-425, Surveillance Frequency Control Program includes the following sections:
A. Internal Events PRA Quality B. Fire PRA quality C. Total CDF, LERF, and RG 1.174 A. Internal Events PRA Quality (27 Pages)
The PWROG performed a full scope internal events PRA peer review of CCNPP to determine compliance with ASME PRA Standard, RA-S-2008a and RG 1.200 (Reference 6.32) in June 2010. This review documented findings for all supporting requirements (SRs) which failed to meet at least Category II. The findings for that peer review are documented below in Table A-1.
This table also includes the disposition, status, and impact on the PRA.
The peer review found that 97% of the SR's evaluated Met Capability Category II or better.
There were 3 SR's that were noted as "not met" and 8 that were noted as Category 1. As noted in the peer review report the majority of the findings were documentation related. Of the 11 SR's which did not meet Category 2 or better, 7 were related to conservatisms or documentation in LERF and 2 were related to internal floods. There were 39 findings. Most of the findings have been addressed in the PRA model. No significant changes have been implemented in the internal events PRA. As there are no new methods applied, no follow on or focused peer reviews were required.
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ATTACHMENT (2)
PRA TECHNICAL ADEQUACY Table A-1 Internal Events PRA Peer Review - Facts and Observations Impact to F&O ID SR Topic FindinglObservation Status Disposition CDF/LERF 1-16 AS-B3 Systems Analysis Based on Sections 2.4 and 2.10 of the Complete The PRA Internal Events No significant impact.
SY-B6 System Analysis Introduction Accident Sequence Subsequent internal Notebook (CO-SY-00, Rev. 0) this SR Notebook, CO-AS-001, events accident appears to be met. However, there is Section 3.3, has been sequence analysis a potential issue related to this SR. updated with an engineering shows that Did not find reference to any analysis of this issue. The Containment Air engineering analysis needed to analysis identifies that during Cooler operation is support Containment Air Cooler the Loss of Offsite Power not challenged by operation when this system is sequences, the Containment containment heat up assumed to be available during LOSP Air Coolers are credited for during LOOP when the containment heats up prior to SBO conditions where the accident sequences electrical recovery. containment heats up, and that credit the CACs then, after power recovery, for recovery.
(This F&O originated from SR SY-B6) the air coolers are credited for containment pressure and temperature control. For these accident sequences, offsite power is restored in one hour, and the containment pressure and corresponding saturation temperature remain well below containment design parameters that would challenge the CACs.
Furthermore, failure of CACs is not risk significant, due to the potential availability of containment spray.
REFERENCE CO-AS-001I 2
ATTACHMENT (2)
PRA TECHNICAL ADEQUACY Table A-1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic FindinglObservation Status Disposition Impact to CDF/LERF 1-17 IFSO-A1 Internal Flooding Examined Internal Flooding Notebook Complete An engineering analysis has No significant impact QU-E3 (CO-IF-001, Rev. 1) Sections 3.0 and subsequently been performed on CDF/LERF.
3.1. Part of the Internal Flood analysis for AFW discharge piping may not be complete for assessing the flooding. The fraction of at-Aux Feedwater Discharge Piping as a power time during which the Flood Source. AFW system is in operation 0.6% and the AFW Discharge (This F&O originated from SR Piping flood may be screened IFSO-A1) due to their low impact on CDF (<1E-9).
REFERENCE CO-IF-001 1-18 IFSO-A4 Internal Flooding Examined Internal Flooding Notebook Open Human-induced impacts on No impact.
IFEV-A7 (CO-IF-001, Rev. 1) Section 3.3 and the flood initiating event Maintenance-5.3. Consideration of human-induced frequencies are not well induced floods are mechanisms as potential flood sources documented. The issue has included. Their not clear. Regarding human-induced been captured in the PRA contribution is based impacts on the flood frequency, configuration control database on industry data Section 5.3 of the IF report states that (CRMP), but not yet starting in 1985, and they were included, but their inclusion addressed. apportioned to should be better documented or various rooms based referenced from IF (e.g., a sample on the number of calculation showing human maintainable contribution would be helpful). components in the room. This is (This F&O originated from SR documented in RAN IFSO-A4)98-062, Rev 1, Section 5.0. The current flood analysis needs to be updated to reflect this.
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ATTACHMENT (2)
PRA TECHNICAL ADEQUACY Table A-1 Internal Events PRA Peer Review - Facts and Observations Impact to Topic FindinglObservation Status Disposition CDFILERF F&O ID SR 1-19 IFEV-B3 Internal Flooding While some items are included in Open In the Internal Flood No significant impact IFPP-B3 Section 7.0 of the IF report, many notebook, the discussions on as this is primarily a IFQU-B3 other instances of uncertainties and uncertainties and documentation issue.
IFSN-B3 assumptions are cited throughout the assumptions should be IFSO-B3 report, but not included in the expanded. This issue has discussion of Section 7.0 nor are the been captured in the PRA implications of these other configuration control database uncertainties and assumptions are (CRMP), but not yet discussed. addressed.
(This F&O originated from SR IFPP-B3) 1-25 DA-C7 Data For the most part actual plant-specific Complete The ESFAS logic train testing No significant impact data is used as a basis for the number has a very low risk on CDF/LERF.
of demands associated actual plant significance and generally experiences (See basis for DA-C6), does not take the logic OOS.
which includes both actual planned The train does go to 2-out-of-and unplanned activities. However, 3 logic. Occurrences where there are a few ESFAS testing and/or the train is in 2-out-of-3 logic other logic channel testing that are not is incorporated into the PRA tracked via the plant computer. Data Analysis Notebook, CO-DA-001, Section 2.6 and Created this F&O on non- 3.5. For the logic relays there documentation of ESFAS/logic train is a RAW of <1.04 and testing, which needs to include actual Birnbaum on the order of 4E-practice. 07. Any logic relay unavailability that does not (This F&O originated from SR DA-C7) cause the ESFAS channel to be OOS and bypassed, is therefore of low significance.
REFERENCE CO-DA-001 4
ATTACHMENT (2)
PRA TECHNICAL ADEQUACY Table A-I Internal Events PRA Peer Review - Facts and Observations Impact to F&O ID SR Topic FindinglObservation Status Disposition CDFILERF 2-7 IFPP-A5 Internal Flood Section 2.3 provides a discussion that Complete A walkdown was performed to No impact on CDF.
walkdowns used to confirm plant assess the susceptibility to jet This is an Internal arrangement. The following impingement or spray in Flood documentation note is contained in section 2.3: rooms 105A and 203. All issue.
equipment is considered Unfortunately, the walk-down failed by spray or documentation from the original impingement for flood sources flooding analysis no longer exists. A originating in the room.
plant walk-down was performed as a Notebook CO-IF-001 was part of this analysis to provide updated with this additional familiarity with the plant design as well documentation.
as confirm findings from the original walk-down. This walk-down is REFERENCE documented in a set of notes and CO-IF-001 photographs included in Appendix B.
Walkdown photos for room 105A and 203 show equipment and potential flood propagation paths.
However, there is not enough spatial information to develop specific targets for flood impingement or spray.
(This F&O originated from SR IFPP-A5) 5
ATTACHMENT (2)
PRA TECHNICAL ADEQUACY Table A-I Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to CDFILERF 2-9 DA-D4 Data Evidence of meeting this SR at Complete Table 2-6 of the Data No impact on CC-Il/Ill is found in the PRA Data Notebook CO-DA-001 listed CDF/LERF. This Notebook (CO-DA-001, Rev. 1) in incorrect data and Bayesian was a documentation Sections 2.1 and 2.7. Found update results for the SACMs. issue. The Internal inconsistencies in the value of total However, the correct values Events model has number components of different types were used in the models for been updated to (for both units) in Table 2-5 of the PRA peer review, include the correct Data Notebook with the actual total data.
number for Calvert Cliffs. Also, found For the SACM EDGs in Table an inconsistency between the prior 2-6, the correct plant-specific distribution and posterior distribution data are in Table 2-5. Table for SACM EDG fail to start in Table 2-6 2-6 lists incorrect data and of the Data Notebook. Bayesian update results for the SACMs. However, the (This F&O originated from SR DA-D4) correct values are used in the models.
The above errors have been corrected in CO-DA-001.
Other minor typographical errors were identified and corrected in the notebook.
REFERENCE CO-DA-001 6
ATTACHMENT (2)
PRA TECHNICAL ADEQUACY Table A-1 Internal Events PRA Peer Review - Facts and Observations SR Topic FindinglObservation Status Disposition Impact to F&O ID CDFILERF 3-3 SY-C2 Systems Analysis Section 2.3 of each system notebook Complete Marked-up system boundary No impact on states that marked up plant system drawings were generated for CDF/LERF. This is drawings are provided as supplements each system notebook, an Internal Events to the system notebook, which depicts Where Unit 1 and Unit 2 are documentation issue the boundary of the system in terms of similar, just the Unit 1 that has been PRA modeling. The drawings are not boundary is depicted. In addressed.
in the notebooks. addition, the system notebooks include drawing (This F&O originated from SR SY-C2) snippets, sketches, and descriptive text that also depict the system boundary.
REFERENCE CO-SY-[AII]
3-5 SY-A 11 Systems Analysis The fault tree does not include Complete A bounding sensitivity case This finding does not SY-A6 potential failures of the AFW was run to include failure of significantly impact accumulator system. the AFW accumulators failing CDF/LERF. The short-term AFW operation. random failure (This F&O originated from SR SY-A11) This issue has an insignificant probability of the contribution to CDF. Short- accumulators is two term failure of the AFW orders of magnitude operation is dominated by lower than active failure of electrical support hardware failures systems and failure of active that support the hardware (i.e., valves and same system instrumentation). The function.
applicable system notebooks were updated.
REFERENCES CO-SY-036 CO-SY-01 9 CO-SY-000 7
ATTACHMENT (2)
PRA TECHNICAL ADEQUACY Table A-1 Internal Events PRA Peer Review - Facts and Observations Impact to F&O ID SR Topic Finding/Observation Status Disposition CDFILERF 3-8 SY-Cl Systems Analysis Several system notebooks were Complete Some new flow diversions Flow diversion has SY-A13 reviewed (AFW, EDG, SI, 120 VAC were identified as part of the been documented electrical, etc.). In general, the Fire PRA Multiple Spurious and are in the documentation is complete and Operation review, and these internal events PRA thorough. In most cases it clearly were added to the system model. The safety follows the RG 1.200 SRs. models and system injection success notebooks. Furthermore, a criteria was verified In some places, assumptions were comprehensive review of PRA to be documented in imbedded in the documentation mechanical systems the Success Criteria without sufficient reference or notebooks and drawings was notebook. These justification. Examples include: performed to identify and were documentation document potential flow issues and do not Sy notebook page 11, last bullet'Only diversions. Flow diversion affect CDF/LERF.
one of the three HPSI pumps functions discussions were added to
- For a cold leg break, it is assumed Sections 3.4.d of the that only one-fourth pump discharge is applicable system notebooks.
spilled via the break. For a hot leg break, the entire pump discharge The success criteria for safety reaches the core.' injection pumps is developed, justified and documented in S1 notebook page 12, 2nd bullet'The the Success Criteria maximum time assumed for operation Notebook, CO-SC-001.
for the safety injection pumps is30 seconds following SIAS initiation.'
CO-SY-000 states that each system notebook addresses flow diversions (where applicable) in section 3.4.d.
Although flow diversions appear to be addressed (for example, the SW notebook talks about flow diversion),
there is no consistent discussion in each system notebook.
(This F&O originated from SR SY-Ci) 8
ATTACHMENT (2)
PRA TECHNICAL ADEQUACY Table A-I Internal Events PRA Peer Review - Facts and Observations Impact to F&O ID SR Topic FindinglObservation Status Disposition CDF/LERF 3-9 DA-B1 Data DA notebook table 2-5 contains the Complete The model has been updated The internal events grouping of components for plant to add additional component model includes the specific failure data. Many of the types and failure modes to updated data and groupings appear to take into account better reflect service failure modes.
differences in such things as size, conditions. Service Water type, mission type (e.g., FW TDP run and Salt Water pumps were vs. AFW TDP standby). However, in broken out. AFW pumps and some cases, it is not clear what the Safety Injection pumps were basis for the grouping is. For example, broken out. This resulted in SW MDP RUN and SRW MDP RUN changes to the associated are grouped together even though they failure rates. The change has are of different service conditions (salt been reflected in the Data water vs. clean water), voltages (480 Notebook, CO-DA-001.
VAC vs. 4160 VAC), size, etc.
Similarly, AFW MDP is included with REFERENCE HPSI MDP and LPSI MDP, even CO-DA-001 though the two SI pumps are pumping borated water, while the AFW pump is pumping condensate grade water. No documentation of the appropriateness of these groupings is provided.
(This F&O originated from SR DA-B1) 9
ATTACHMENT (2)
PRA TECHNICAL ADEQUACY Table A-1 Internal Events PRA Peer Review - Facts and Observations SR Topic Finding/Observation Status Disposition Impact to F&O ID CDFILERF 3-11 QU-B7 Quantification The mutually exclusive cutsets for Complete A comprehensive review of No impact on each system are described in the mutual exclusive modeling CDF/LERF. The system notebook section 3.4.e. was performed. Each system PRA model has been Several SY notebooks were reviewed notebook and each system updated where to determine appropriateness of the model was reviewed to needed.
mutually exclusive cutsets. All validate the appropriateness appeared reasonable. A review was of the modeling and reconcile performed of the MUTEX gate within any differences, and to verify the fault tree model and the that a documented basis appropriate combinations identified in exists for each mutually the SY notebooks appear have been exclusive event. The PRA included in the model. There are two model was updated to reflect gates under the MUTEX gate which new, deleted, or re-organized contain mutually exclusive cutsets mutually exclusive modeling which are not documented in the identified as part of this system notebooks. While the majority review.
of these are intuitively obvious (e.g.,
11 Steam Generator Tube Rupture REFERENCE occurs as an IE AND 12 Steam SY-CO-[ALL]
Generator Tube Rupture occurs as an IE), these should be included in an appropriate system notebook.
(This F&O originated from SR QU-B7) 10
ATTACHMENT (2)
PRA TECHNICAL ADEQUACY Table A-1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to CDFILERF 3-12 QU-D3 Quantification A review of the top cutsets from each Complete Documentation of the cutset No impact on event tree was performed. The utility reviews was presented to the CDF/LERF. The stated that during this review, cutsets peer review team; although, original internal were reviewed to determine if any the documentation was events cutset review mutually exclusive events were separate from the formal QU notes have now been contained within cutsets, ifany flag notebook package. A note archived.
settings were inappropriate or ifany was added to the QU recoveries were overlooked or added notebook directing the reader inappropriately. A review of a to the location of the cutset sampling of cutsets did not indicate review notes and any inappropriate results. However, spreadsheets. The PRA the QU notebook does not include a configuration control discussion of this review, procedure, CNG-CM-1.01-3003, requires a review of (This F&O originated from SR QU-D3) cutsets for PRA changes. In practice, the top CDF and LERF cutsets are examined for even the most innocuous model changes.
REFERENCES CNG-CM-1.01-3003 CO-QU-001 CO-FRQ-001 11
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PRA TECHNICAL ADEQUACY Table A-1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to CDF/LERF 4-5 IE-A10 Initiating Events The only mention in CO-SC-001 of Complete To address this finding, the No impact. The SY-Al0 shared systems between the units is Diesel Generator modeling finding has been IE-C3 the SBO EDG, noted in Section 4.1.2. was updated as described in addressed in the SC-A4 It states that the SBO diesel can power Appendix H of CO-SY-023- Internal Events any one bus on either unit. However, 024, PRA DG System model.
in the CAFTA model, there is an Notebook. EOP-7 directs to assumed bus preference of 11, then align the OC DG to the unit 24, then 12, then 23.* This is noted in with redundant safety the EDG system notebook but no equipment out-of-service, with basis is provided. The procedures do a goal to restore at least one not actually have a preference, which 4KV bus. Since 4KV Buses yields a potentially non-conservative 11 and 24 support AFW, analysis. For example, if there is a those busses would have a LOOP, the U2 diesels fail to start and preference over Busses 14 the Ul diesels fail to run after 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. and 21, all else being equal.
The SBO diesel would then be aligned No unit preference is to U2, and it is non-conservative to modeled. Ifthere is a conflict give the U1 bus 11 full credit. If such in the order-of-preference, for non-conservatism is negligible, some example, both 4KV Bus 11 analysis should be performed to and 4KV Bus 24 are not demonstrate this. powered, then a 50-50 probability is assumed as to (This F&O originated from SR IE-A10) the preferred bus.
- Note: Peer review finding was not REFERENCE precise. It should have stated bus CO-SY-023-024 preference for Unit 1 is 11, then 24, and for Unit 2, is 24 then 21.
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PRA TECHNICAL ADEQUACY Table A-I Internal Events PRA Peer Review - Facts and Observations Impact to SR Topic Finding/Observation Status Disposition CDF/LERF F&O ID 4-12 HR-Cl Human Reliability One basic event calculated in the Complete The basic event has been No impact. The appendix (ESFOHFCISZDEFG) was added to the model. A missing basic event not included in the fault tree models. sensitivity run with the basic has been added to CCNPP staff noted that it had event included the current the internal events previously been modeled, but model showed no increase in model.
inadvertently deleted in an update. risk. The system notebook CO-SY-048 was updated.
(This F&O originated from SR HR-Cl)
REFERENCE CO-SY-048 4-15 IFEV-A6 Internal Flooding The internal flooding analysis did not Open CCNPP is typical of a PWR The review of have a formal process to gather plant that uses raw water for condition reports did specific design information, operating cooling. Raw water, in our not identify any practices, etc. that could potentially case salt water from the design issues or affect the generic flooding frequencies. Chesapeake Bay, is used to operating practices In response to an NRC RAI on the cool the Component Cooling that would affect the CCNPP ISl program plan, CCNPP Heat exchangers, Service generic flooding mentioned a review of Condition Water Heat exchangers, main frequencies.
Reports that did not find any items that condensers, and ECCS Pump Therefore, no would increase the flooding frequency. Room air coolers. Salt water significant impact is floods comprise the greatest expected.
The CR review meets part of the flood risk as there is an requirement, but the SR also calls for infinite volume. However, reviews of plant design, operating these systems are designed, practices, etc. that should be as typical, to be able to isolate considered. The evaluation should be a heat exchanger or documented in the PRA. condenser for draining and cleaning. CCNPP is unlikely (This F&O originated from SR to have any plant specific IFEV-A6) design or operating practices that would affect the generic flooding frequencies.
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PRA TECHNICAL ADEQUACY Table A-1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic FindinglObservation Status Disposition Impact to CDFILERF 4-19 LE-C13 Large Early The sources of uncertainty are well Complete Dominant LERF cutsets were No significant impact.
LE-F3 Release identified in Table 5-1 of the LE reviewed to identify The dominant LERF LE-G4 notebook and quantified in Table 5-2 uncertainties that could be contributors were of the QU notebook. However, no addressed. Two changes reviewed and model discussion of the uncertainties or have been implemented to changes insights from them is provided. For address significant implemented. The example, Sensitivity 1 shows a 74% uncertainties and reduced Calvert Cliffs LERF reduction in LERF, but this large LERF. First, a reverse-flow contribution is now reduction is not investigated, check valve in the CVCS similar to other Letdown line was credited as PWRs.
Also, conservatisms in the ISLOCA a potential ISLOCA recovery.
analyses were discussed in the AS Second, a new human action review. SGTR was treated in an was added with realistic overly conservative manner by timing for Steam Generator categorizing all SGTR as LERF. isolation and RCS depressurization on a SGTR.
(This F&O originated from SR LE-F3) These and less significant model updates resulted in a LERF-to-CDF ratio change from approximately 17% to approximately 10%. This newer ratio is in the typical range for other PWRs.
REFERENCE CO-LE-001 14
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PRA TECHNICAL ADEQUACY Table A-1 Internal Events PRA Peer Review - Facts and Observations StatsDiposiionImpact to F&O ID SR Topic FindinglObservation Status Disposition CDFILERF 4-20 LE-F1 Large Early The relative contribution to LERF is Complete The contributions to LERF are No impact. This is LE-G3 Release presented in the QU notebook by PDS documented in the an internal events and by initiating event, but not by Quantification Notebook and documentation issue.
accident progression sequence, are noted as such in the Level phenomena, containment challenges 2 Notebook. Accident or containment failure mode. progression sequences are located in Section 4.2.3 and (This F&O originated from SR LE-G3) Appendix C. The Level 2 notebook has been updated to point to additional phenomena and containment challenges and failure mode Tables/Figures in the QU Notebook.
REFERENCES CO-QU-001 CO-LE-001 4-21 LE-G5 Large Early The LE notebook states that limitations Complete Section 5.5.2.7 of CO-LE-001, No significant impact.
Release in the LE analysis that could impact Revision 2 - added discussion applications are documented in the QU of this issue and how it was notebook, but it is not. Given the addressed. Model changes conservative modeling of SGTR and included crediting the CVCS ISLOCA, the impact on applications reverse flow check valve and should include assessment of how this adding a new human action conservatism can skew the LERF with realistic timing for results. isolating the steam generators and de-pressurizing the RCS (This F&O originated from SR LE-G5) on a SGTR.
REFERENCE CO-LE-001 15
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PRA TECHNICAL ADEQUACY Table A-1 Internal Events PRA Peer Review - Facts and Observations Impact to SR Topic Finding/Observation Status Disposition CDF/LERF F&O ID 4-22 LE-Cl0 Large Early The LERF contributors have not been Complete The LERF results were No significant impact.
LE-C12 Release reviewed for reasonableness (per SR reviewed for conservatisms The dominant LERF LE-F2 LE-F2). The QU notebook discusses as described in the SRs. contributors were LE-C3 the top 20 LERi cutsets (which total After conservatisms were reviewed and model 73% of the total LERF). It notes addressed (see discussion for changes conservatism in the cutsets and says it F&0 4-19 above), no implemented. The will be evaluated in Section 5.2, but is significant issues were Calvert Cliffs LERF not. Section 4.3.6 of the QU notebook identified. contribution is now compares the total LERF of CCNPP to similar to other St. Lucie, but does not even break the REFERENCE PWRs.
results down by contributor (e.g., CO-LE-001 SGTR, ISLOCA, etc.).
Also, the ASME PRA Standard SRs C-3, C-la0 and C-1 3 require a review of the LERF results for conservatism in the following areas:
- 1. Engineering analyses to support continued equipment operation or operator actions during severe accident progression that could reduce the LERE.
- 2. Engineering analyses to support continued equipment operation or operations after containment failure.
- 3. Potential credit for repair of equipment.
No such review has been performed, despite the large conservatism noted in the containment bypasses.
(This F&O originated from SR LE-F2) 16
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PRA TECHNICAL ADEQUACY Table A-1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to CDF/1LERF 5-10 LE-D7 Large Early Following the failure of one or more Complete The merits have been Modeling of an Release containment penetrations to isolate on considered of adding an operator action to CIAS, a feasible operator action is to operator action in order close manually close failed manually close the failed valves from containment penetration from valves from the main the Main Control Room. the Main Control Room to control room would recover from a containment not significantly (This F&O originated from SR LE-D7) isolation failure. A review of reduce LERF, as cutsets shows that a recovery such an action is not is not feasible for top LERF feasible for the sequences, because the significant sequence includes either 1) a sequences where loss of CR indication, 2) containment isolation includes a station black-out has failed.
condition, or 3) includes non-recoverable pipe breaks.
REFERENCES CO-LE-001 Attachment S 5-17 IE-Cl Initiating Events Bayesian updates of non-time-based Complete CENG understands the No impact. The IE-C13 LOCA data were improper. The small general concern on Bayesian approach used for IE-C4 and medium LOCA frequencies were updating of rare events. LOCA frequencies obtained from draft NUREG 1829 then However, the method used has been validated Bayesian updated (in App E) with was based on a white paper by industry experts CCNPP experience from 2004 to developed by industry experts and is the same 2008. The Very Small LOCA prior regarding LOCA frequencies. approach as was having alpha = 0.4, Mean = 1.57E-03; These experts included INL, used for the NRC's was Bayesian updated to a Posterior NRC and Industry experts. In SPAR model.
having a mean value of 7.02E-04. addition, the approach used This represents an excessive drop for the Calvert PRA was the associated with CCNPP experience of same as used for the NRC 4 to 5 years. Similarly, the Small and SPAR model. This issue is Medium LOCAs were Bayesian captured in the PRA updated with the whole industry configuration control database 17
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PRA TECHNICAL ADEQUACY Table A-I Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to CDF/LERF experience rcy data. The draft (CRMP).
NUREG 1829 LOCA frequencies were obtained from expert elicitations (not REFERENCE time-based) that included crack CO-IE-001 propagation analysis. The Bayesian update for VSLOCA used the Alpha parameter and the mean value to justify that the prior mean was based on 255 rcy. This may not have been the basis for the expert elicitations in NUREG 1829.
Also, the Medium LOCA frequency may be classified as extremely rare event. It would require no Bayesian updating. The current CCNPP SLOCA and MLOCA frequencies are very close even though the source data in NUREG 1829 indicates a negative exponential drop in these frequencies.
(This F&O originated from SR IE-Cl)
(Note: rcy - reactor year) 5-18 IE-C2 Initiating Events Justify the exclusion of LOOP event at Complete The event is not counted No impact. The data IE-C7 CCNPP in 1987. No time trend following guidance provided in analysis is analysis was provided to justify the NUREG/CR-6928, based acceptable.
exclusion. upon trend analysis. A full discussion is included in the (This F&O originated from SR IE-C2) Initiating Event notebook, CO-IE-001.
REFERENCE CO-IE-001 18
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PRA TECHNICAL ADEQUACY Table A-I Internal Events PRA Peer Review - Facts and Observations Impact to SR Topic Finding/Observation Status Disposition CDF/LERF F&O ID 5-23 HR-A2 Human Reliability The Pre-Initiator HRAs did not include Complete It is agreed that the No impact. Given the miscalibration of SIT pressure. For miscalibration of SIT pressure the pressure of the example, in the event where SIT could have a negative impact CCNPP SITs they pressure is miscalibrated high, various on various accident scenarios are only required and accident scenarios requiring SI are involving LLOCA and provide significant negatively impacted. Add SIT VLLOCA initiators. However, benefit on Large pressure miscalibrated high or, justify this instrumentation is not LOCAs. The no impact on CDF / LERF. modeled explicitly and is frequency of a Large therefore deemed included LOCA times the pre-(This F&O originated from SR HR-A2) within the component initiator frequency is boundary for the SIT. As negligible.
such the miscalibration probability would be included in the SIT unavailability.
REFERENCE CO-HR-001 5-25 SC-Cl Success Criteria Simplify the traceability of Tsw. In the Complete Where applicable, the Tsw of No impact. This is HR-12 post initiator HRA details, the HRA each HFE that could be an internal events SC-C2 success criteria are often provided as traced to the Success Criteria documentation a positive re-statement of the HRA notebook (CS-SC-001) was finding.
title. And, the consequence of failure updated and referenced in the is often stated as core damage. HRA Calculator. CO-HR-001 Consider adding Tsw to the success was also updated.
criteria and linking that to the PCTran case where Tsw was developed. Also, REFERENCE in the SC report (Table B-3), consider CO-HR-001 adding the actual time to core uncovery (or core damage) instead of providing a "Yes" entry in the column of "core damage?"
(This F&O originated from SR HR-12) 19
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PRA TECHNICAL ADEQUACY Table A-1 Internal Events PRA Peer Review - Facts and Observations Impact to F&O ID SR Topic FindinglObservation Status Disposition CDF/LERF 5-30 LE-D1 Large Early Section 3.2.11 discussed the Complete CCNPP's Level 2 PRA follows No impact. The LE-B2 Release containment challenge from Hydrogen the analysis in WCAP-16341- methodology in Combustion. It concluded that the P, Simplified Level 2 Modeling WCAP-16341-P is challenge may be significant for some Guidelines. In the industry- appropriate for accident scenarios. The CCNPP entry supported analysis, the Calvert Cliffs level 2 in Table 6.11-2 of the Level 2 WCAP percentage of cladding analysis.
showed a potentially significant impact oxidation is the main factor from Hydrogen burn. Provide an used to develop a maximum estimate of the impact of Hydrogen H 2 concentration in the burn on containment pressure. Use an containment, and, in turn, a accident scenario that is likely to containment pressure is produce larger amounts of H2 with calculated if the H2 completely failed containment spray. The optimal burns. These are then time to estimate the impact of mapped to site-specific Hydrogen burn is approximately at 2 containment failure hours which is the time when the EOF probabilities.
and TSC personnel have convened and are ready to guide the Main A simplifying assumption is Control Room into periodic Hydrogen made that "no pre-burning of burns before the formation of explosive hydrogen generated in the mixtures. core melt progression is considered." Calvert Cliffs' (This F&O originated from SR LE-D1) severe action management procedures do include actions to reduce H 2 concentration in the containment, but these actions are not credited in the PRA model. Also, Containment Spray is not questioned for the LERF accident sequences.
Containment Spray is a factor in LATE containment failure accident sequences.
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PRA TECHNICAL ADEQUACY Table A-1 Internal Events PRA Peer Review - Facts and Observations Impact to F&O ID SR Topic Finding/Observation Status Disposition CDF/LERF REFERENCE CO-LE-001 5-31 DA-D4 Data The summary table for Bayesian Complete The aforementioned footnote No impact. This is a updated parameters (on Page 53 of was incorporated into Table minor internal events the PRA Data Notebook, CO-DA-001, 2-6 of CO-DA-001. documentation issue Rev. 1) shows the CS-MDP was and no changes Bayesian updated with plant REFERENCE were required for the experience containing 1 failure and CO-DA-001 CS-MDP failure rate.
Zero run-hours. The CCNPP PRA staff responded to this issue as an isolated case. There is an actual FTR
> 1 hr (This F&O originated from SR DA-D4) 21
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PRA TECHNICAL ADEQUACY Table A-1 Internal Events PRA Peer Review - Facts and Observations Impact to F&O ID SR Topic FindinglObservation Status Disposition CDF/LERF 6-3 SC-B2 Success Criteria Expert judgment was not used as the Complete The approach for SLOCA No impact.
sole basis for any success criteria, break size analysis is However, upon inspection of the discussed in the Success PCTran run tables in the SC report Criteria notebook.
appendices, many instances of Furthermore, a review was surrogate or inferred results were conducted of this issue; in found. Instead of running specific addition, TH analyses were PCTran calculations to cover the completed to verify the break-whole SLOCA break size spectrum, size ranges. Itwas found that intermediate break sizes have been the computer simulations calculated supplemented with expert adequately represented the judgment to derive limiting time delay various break-size ranges.
for operators to actuate SI (30 min) or Notwithstanding the argument limiting time delay for OTCC above, two additional SLOCA (SGL<350'+1 0min). computer simulation runs were made and documented (This F&O originated from SR SC-B2) in Success Criteria Notebook CO-SC-002 - PCTran simulations. Success Criteria Notebook CO-SC-001 was updated to clarify and incorporate the additional reviews.
REFERENCE CO-SC-001 22
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PRA TECHNICAL ADEQUACY Table A-I Internal Events PRA Peer Review - Facts and Observations Impact to F&O ID SR Topic FindinglObservation Status Disposition CDFILERF 6-5 SY-A20 Systems Analysis When appropriate, the simultaneous Complete AFW basic event No impact. The unavailability within a system is AFWOTMMAINT6-F7 was offending basic event documented in the system notebooks determined to not be needed was removed from and included in the PRA model. in the plant model. The basic the model. A review However, a further review of these event was removed. All did not discover items is required for completeness. remaining AFW equipment other missing or unavailability events in the incorrect (This F&O originated from SR SY-A20) model and notebooks were simultaneous reviewed for consistency. unavailability events.
AFWOTMMAINT-TF was determined to be modeled correctly, its description was found to be in error in the system notebook. Notebook CO-SY-036 was updated. A review for concurrent maintenance was previously performed and documented in the Data Notebook.
REFERENCE CO-SY-036 23
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PRA TECHNICAL ADEQUACY Table A-I Internal Events PRA Peer Review - Facts and Observations Impact to CDFILERF F&O ID SR Topic Finding/Observation Status Disposition 6-8 HR-H2 Human Reliability Some recovery actions included in the Complete For each screening HRA, the No impact. The model (thus credited) are set to internal events analysis was documentation for screening values. In the HEP updated to include a specific internal events HRAs evaluation (appendices of the HR reference to the earlier HRA was updated to report) there are no indications that analysis. Included are the address this finding.
procedures, training, or other shaping applicable success criteria for factors are available on a plant-specific each recovery. Refer to basis. CO-HR-001, Internal Events Human Reliability Analysis, (This F&O originated from SR HR-H2) and the associated HRA Calculator file.
REFERENCES CO-HR-001 CO-HRA-001 6-9 HR-I1 Human Reliability The HR report is well documented in Complete Updated the notebooks in the No impact. This is a general and will facilitate upgrades, reference section so HRA documentation however, some basic event names are designator names and finding. HRA names not consistent between the HR report descriptions are the same in in the model and and the system notebooks. the HR Calculator, HR notebook are now notebook, CAFTA Model 6.0. consistent.
(This F&O originated from SR HR-I1) Changes included adding the
"-B" extension and removing the "(-2)" event where applicable.
REFERENCES CO-HR-001 CO-SY-[Many]
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PRA TECHNICAL ADEQUACY Table A-1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic FindinglObservation Status Disposition Impact to CDF/LERF 6-10 IFPP-A2 Internal Flooding Plant design features such as open Complete The Internal Flood notebook No impact.
IFSN-A2 rooms or as built divisions are used to has been updated to define the flood areas and was well incorporate an analysis documented. More detail is needed as describing the screening of to why the containment buildings were the containment building from screened from the analysis. flooding analysis. Essentially, the containment is designed (This F&O originated from SR for LOCA condition, which IFPP-A2) screens reactor coolant system and related piping system. Other piping systems have limited inventory, are normally isolated, or have a low flow rate. Reference CO-IF-001.
REFERENCE CO-IF-001 6-14 IFSO-B1 Internal Flooding While the flooding calculations have Open This is a documentation No impact as this is a IFSN-A9 been performed and are thought to be finding for the internal floods documentation issue.
correct and well done, additional notebook. The issue has documentation of data would enhance been captured in the PRA the IF report. It appears that the input configuration control database reports and references are based on (CRMP), but not yet closed-poorly documented or non-officially out.
revisioned reports and information sources.
(This F&O originated from SR IFSN-A9) 25
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PRA TECHNICAL ADEQUACY Table A-I Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to CDFILERF 6-16 IFQU-A1 1 Internal Flooding Walkdowns have been conducted and Open This is an internal floods No impact. This is a IFPP-B2 are documented in Appendix B of the documentation finding. The documentation issue.
IF report. It is stated in the IF report finding has been captured in that prior information is no longer the PRA configuration control available; this fact should be corrected database (CRMP), but not yet as required for analysis updates and addressed.
information verifications.
(This F&O originated from SR IFQU-A1 1) 26
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PRA TECHNICAL ADEQUACY Table A-1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to CDFILERF 6-17 IFQU-A10 Internal Flooding By including the flooding events under Open This internal flood issue is This finding may the transient fault tree, the LERF captured in the PRA drive some changes impacts are automatically accounted configuration control database to the model for for in the same manner as the general (CRMP), but not yet internal floods but it transient events in the LERF analysis. addressed. There is some is not expected to be Very little documentation is found potential that the analysis, in significant. The level related to the IF analysis in the LE addition to providing the of modeling detail in report, although the IF report states required documentation, may the CCPRA is that the LERF impacts due to flooding drive changes to the model. sufficiently robust are documented and analyzed in the such that the model LE report. logic for flood impacts propagate (This F&O originated from SR appropriately through IFQU-A10) the system fault trees so that the equivalent general transient initiator (e.g. loss of CCW) is appropriately defined in the transient fault tree. In addition, cutset reviews have not revealed the current modeling to be deficient in this regard.
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PRA TECHNICAL ADEQUACY Table A-1 Internal Events PRA Peer Review - Facts and Observations Topic FindinglObservation Status Disposition Impact to F&O ID SR CDFILERF 6-18 HR-H2 Human Reliability The system time window Tsw for post Complete It was determined that the text No impact. As initiator HRAs was frequently in Section 3.1.5.7 was described in this associated with 'core damage'. Post incorrect and does not F&O for internal initiator HRAs that appear in the top capture how stress is actually events, the stress cutsets may require success criteria applied in the EPRI HRA levels in the model linked to beginning of core uncovery Calculator. CO-HR-001, are appropriate, but (about 20 minutes before 'core Internal Events PRA Human updates to the damage'). Or, the operator actions Reliability Analysis, has been documentation are that may fall into that final 20-minute updated to show the stress required. The time period should be overridden to level applied to each HFE and internal events assume a high stress level. While the justification for stress documentation was section 3.1.5.7 described this selection. Also included is a updated.
approach, there is no evidence of its correlation between stress proper application in the HRA level and failure of execution quantifications. probability. New text has been provided for inclusion in (This F&O originated from SR HR-H2) a future update of the HRA notebook.
REFERENCE CO-HR-001 CO-HRA-001 28
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PRA TECHNICAL ADEQUACY Table A-1 Internal Events PRA Peer Review - Facts and Observations SR Topic Finding/Observation Status Disposition Impact to F&O ID CDFILERF 6-22 HR-El Human Reliability Upon RAS, LPSI stops and EOP-5, Complete As documented in CR-2009- No impact. The Step S.1(d) requires the Operators to 005881, shutting the RWT system is operable
'Shut RWT OUT Valves SI-4142, outlet valves upon a RAS without the manual 4143'. This manual action was not does not impact station action to shut the modeled in the PRA. The CCNPP operability. The Safety RWT outlet valves.
PRA staff provided reasonable Injection Pumps and There is no impact response to this issue. Based on Containment Spray Pumps on internal events CR-2009-005581, there is no impact will not fail if the RWT CDF/LERF. The on pump operability. Also, the staff will isolation valves do not closed issue was added to continue to track the CR. If there are with a RAS signal. A design the plant's margin any changes to the disposition of margin issue has been management pump operability, then a new HRA identified. This issue has program.
may be added to the PRA model (if been added to the plant's warranted). margin management program. No model changes (This F&O originated from SR HR-E1) have been made, but the PRA configuration management program, CNG-CM-1.01-3003, would capture any design changes concerning this issue.
REFERENCES CO-SY-052 CR-2009-005881 CNG-CM-1.01-3003 29
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PRA TECHNICAL ADEQUACY Table A-1 Internal Events PRA Peer Review - Facts and Observations Impact to F&O ID SR Topic FindinglObservation Status Disposition CDF/LERF 6-23 HR-G7 Human Reliability When the Calculator reads in the Open New HRA events, This specific issue combinations, it assumes that actions CVCOHFOTA8HRS and with time delay and occur in the order of the time delay AFWOHFCCSGDEC8HR CST depletion has (Td). However, the time delay is not were added to model Td been addressed and the same for all sequences, and care variances where CST incorporated into the must be taken to make the depletion occurs early and PRA model.
combinations appropriate for the when it occurs later. This sequences in which they occur. Page account for appropriate 88 of the HRA notebook indicates this sequencing of events.
was considered, since the Td was modified for events occurring prior to An updated dependency reactor trip, and also for OTCC after analysis has been performed, SG overfill. However, not all which includes these new occurrences have been addressed. HRA events. The The combination examined by the dependency analysis shows review team is Combination 770 that these new HRA actions (OTCC after CST depletion). In this are not significant for CDF or event the CST depletion should come LERF. A PRA configuration first. control database (CRMP) item has been initiated to (This F&O originated from SR HR-G7) formally incorporate the updated dependency analysis into the model.
REFERENCES CO-HR-001 CO-HRA-001 30
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PRA TECHNICAL ADEQUACY Table A-1 Internal Events PRA Peer Review - Facts and Observations Impact to CDFILERF F&O ID SR Topic Finding/Observation Status Disposition 7-13 QU-A2 Quantification Discrepancy between documentation Complete The top flood cutset was No impact. This was and result files. SB0037 and SB0038 incorrectly flagged as being an Internal events sequences appear to be inverted in SBO sequence 37 (offsite documentation issue.
Tables D-1, 4.2.2, 4.2.4, 4.2.5, B-3). power recovered < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) instead of sequence 38 (This F&O originated from SR QU-A2)- (offsite power not recovered).
Updated tables B-2, C-1, and D 1 in CO-QU-001. Spot-check was performed to identify other errors. In C0-QU-002, fixed sequence 12 table 4.2-5, which incorrectly showed sequence 37 instead of 38.
REFERENCES CO-QU-001 CO-QU-002 31
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PRA TECHNICAL ADEQUACY B. Fire PRA Quality (27 Pages)
In accordance with RG 1.205 position 4.3:
"The licensee should submit the documentation described in Section 4.2 of Regulatory Guide 1.200 to address the baseline PRA and application-specific analyses. For PRA Standard"supportingrequirements"important to the NFPA 805 risk assessments, the NRC position is that Capability Category II is generally acceptable. Licensees should justify use of Capability Category I for specific supporting requirements in their NFPA 805 risk assessments, if they contend that it is adequate for the application. Licensees should also evaluate whether portions of the PRA need to meet Capability Category Ill, as described in the PRA Standard."
The CCNPP FPRA peer review was performed January 16-20, 2012 using the NEI 07-12 Fire PRA peer review process, the ASME PRA Standard (ASME/ANS RA-Sa-2009) and Regulatory Guide 1.200, Rev. 2. The purpose of this review was to establish the technical adequacy of the FPRA for the spectrum of potential risk-informed plant licensing applications for which the FPRA may be used. The 2012 Calvert FPRA peer review was a full-scope review of all of the technical elements of the CCNPP at-power FPRA (2012 model of record) against all technical elements in Section 4 of the ASME/ANS Combined PRA Standard, including the referenced internal events SRs. The peer review noted a number of facts and observations (F&Os). The findings and their dispositions are provided in Table B-1. All findings are being provided and have been dispositioned. All F&Os that were defined as suggestions have been dispositioned and will be available for NRC review. The FPRA is adequate to support the NFPA 805 licensing basis.
The FPRA now meets at least Capability Category II in all cases. Eleven ASME/ANS SRs were identified by the peer review team as meeting Capability Category I only requirements or a level of "not met" for the requirement. The capability categories are defined in ASME/ANS RA-Sa-2009. An evaluation of the impact of those areas where only the Capability Category I requirement was met or the requirement was "not met" is provided in Table B-2 along with the basis for now meeting Capability Category I1.
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PRA TECHNICAL ADEQUACY Table B-1 Fire PRA Peer Review - Facts and Observations - Findings F&O ID SR Topic Status Finding Disposition PP-B3-01 PP-B3 Plant Complete The containment is partitioned into 2 CO-PP-001, Calvert Cliffs Fire PRA Plant PP-B6 Partitioning PAUs. There are intervening Partitioning Notebook, was updated to include an PP-C3 combustibles and this was accounted analysis that justifies the partitioning of the for in the PRA by treating the 20 feet as containment into two plant partitioning units with a an overlap region and failing 20-foot spatial separation (known as the buffer components affected in both PAUs. zone). The only potential intervening combustibles There is no justification given for the 20 in this buffer zone were identified as qualified foot assumption. The turbine deck is cables that were verified to be encased within continuous from unit 1 to unit 2. This marinate covered raceways. The covers prevent area is divided into 2 PAUs, TURB1 and the cables from becoming potential combustibles TURB2, but there is no discussion for and therefore are not considered intervening the basis of the partitioning. Finding combustibles.
level of significance is based on crediting spatial separation with no The unit 1 and unit 2 Turbine Deck was walked requisite justification. down to assess for the acceptability of the Appendix R partitioning into distinct PAUs. The Maintain the containment as 1 PAU and boundary was assessed to have at least a 20-foot discern the separation of east from west separation between potential ignition sources and in the fire modeling. Document the potential targets, assessed for intervening spatial separation and no intervening combustibles, and the Turbine deck volume combustibles for the turbine deck. assessed for damaging hot gas layer development.
The partitioning was found acceptable and consistent with NUREG/CR-6850, Section 1.5.2, where main turbine decks are typical applications where spatial separation has been credited.
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PRA TECHNICAL ADEQUACY Table B-1 Fire PRA Peer Review - Facts and Observations - Findings F&O ID SR Topic Status Finding Disposition PP-B5-01 PP-B5 Plant Complete The water curtain in the CCW room was The Component Cooling Water room water curtain PP-C3 Partitioning credited as an active fire barrier. The is an approved Appendix R exemption, as justification was that the water curtain identified in the exemption issued by the NRC in was part of the original regulatory fire response to Calvert Cliffs exemption request protection program. This meets CAT 1, ER820816. The validity of crediting CCW Room but needs enhancement for CAT Il/111. Water Curtains is discussed in Southwest Finding level was used because the Research Institute Report No. 01-0763-201. A requirements for CAT Il/111 were not met. reference to the Southwest Research Institute report was added to CO-PP-001, Plant Partitioning Calvert Cliffs should provide a direct Notebook.
reference to their Appendix R program as the basis for the acceptability for this or provide a design basis justification for the water curtain and document that in the PP notebook ifthe Appendix R program reference cannot be found.
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PRA TECHNICAL ADEQUACY Table B-1 Fire PRA Peer Review - Facts and Observations - Findings F&O ID SR Topic Status Finding Disposition PP-B7-01 PP-B7 Plant Complete 1. The walk down nomenclature does A table was created to correlate the building or PP-C3 Partitioning not match the PP notebook. Example area nomenclature that was used for the plant PP-C4 Qualitative page 561 of the walkdown walkdown documentation, to the plant analysis unit QLS-AI Screening documentation uses nomenclature in identifiers used in the Fire PRA analysis. This the containment that does not match the table was added to CO-PP-001, Calvert Cliffs Fire PP notebook. PRA Plant Partitioning Notebook as Table 17.
- 2. There are many areas inaccessible The facilities and rooms that were not originally such as: #23 Charging Pump Room, U1 walked-down were reviewed. Supplemental Service Water Pump Room, U1 East walkdowns were performed and supplemental Battery Room, E/W Corridor. These walkdown datasheets were generated. For areas areas appear to be accessible with a that were not accessible at the time of the little effort. In some of the areas supplemental walkdowns (for radiological safety screened out in QLS, the areas were reasons, personnel safety concerns, or access inaccessible and did not have a otherwise denied), The reason for inaccessibility confirmatory walkdown. Finding level was added to Table 17.
assessed due to the incompleteness of the walkdown documentation.
- 1. Prepare a table that correlates the PAUs from the PP notebook with the area nomenclature used in the walkdown documentation.
- 2. Complete the walkdowns, particularly for areas screened in the QLS task.
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PRA TECHNICAL ADEQUACY Table B-I Fire PRA Peer Review - Facts and Observations - Findings F&O ID SR Topic Status Finding Disposition CS-Bl-01 CS-B1 Fire PRA Cable Complete Current Breaker coordination study still The breaker coordination study has been CS-C4 Selection and in progress. This study needs to be completed. As described in ECP-1 3-000321, Form Location completed in order to receive a category 12, Engineering Evaluation, all PRA common II met for CS-B1. power supplies are assumed to meet - or will meet
- the coordination requirements of NFPA 805, Complete the breaker coordination except as noted in CO-CS-001, Fire PRA Cable study. Selection Notebook. As described in the cable selection notebook, two 120VAC lighting panels are not validated as coordinated, and these panels are assumed to fail for all Fire PRA scenarios.
Also, as described in the PRA notebook a breaker for 480V motor control center MCC101 BT has not been validated as coordinated. This breaker, 52-10150, is modeled so that a fire-induced electrical fault on the breaker's power cabling will fail MCC101BT. Finally, the notebook identifies that selected 120V power panels have coordination issues, but that these will be addressed by design changes and referenced in Attachment S -
Modifications and Implementation Items.
PRM-B3-01 PRM-B3 Fire PRA/Plant Complete The FPRA model did not address Loss of Control Room HVAC can affect the PRM-B4 Response events involving loss of both HVAC operability and availability of equipment in the PRM-B5 Model trains to the MCR, long term heatup of control room and cable spreading room. As MCR and need for operator actions described in Calvert PRA System Analysis outside the MCR to compensate for the Notebooks CO-SY-002, CO-SY-017, and loss of electronic controls in the MCR, CO-SY-030, loss of HVAC is modeled to have the which was assumed as a CCDP of 1.0 effect of increasing the failure rate of 120VAC and for the plant. The basis for excluding 125VDC instruments and controls in the cable this potential Core Damage sequence spreading room. For the control room, degradation was addressed in questions to the of the 125VDC system is used as a conservative Calvert Cliffs PRA team. This sequence surrogate for control room I&C degradation.
is a new sequence outside the current model FPRA model logic trees. Loss of Control Room HVAC and subsequent temperature increases may adversely affect 36
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PRA TECHNICAL ADEQUACY Table B-1 Fire PRA Peer Review - Facts and Observations - Findings F&O ID SR Topic Status Finding Disposition Consider using a combination of MCR operator responses. The model reflects heatup calculations to define the time degradation of human actions by the degradation when operators would leave MCR and of the 125VDC system used for instruments and consider a recovery action for restoring controls. Loss of Control Room HVAC is not cooling the MCR. expected to cause abandonment by operations staff of the control room due to high temperatures.
On complete loss of HVAC with no mitigation, such as no use of emergency fans, calculation CA02725 shows a CR temperature of 123 deg F at 24-hours.
While this is a challenging environment, this temperature is assessed as insufficient to solely drive a complete CR abandonment scenario.
NUREG/CR-6738 describes operational experience where operators will continue to occupy the control room even under severe environments.
Operations staff says that in consideration of high temperatures in the control room, that Operations would do what was needed to keep the cores safe and covered. The site safety director says that for a temperature of 123 deg F, the site would implement a mitigation strategy which would include stay-times, assessment of individuals for heat-related conditions, use of ice vests, and call-in of additional qualified operations staff to rotate into the control room.
The above discussion was included in CO-SY-030, Control Room HVAC PRA System Notebook.
FSS-A5-01 FSS-A5 Fire Scenario Complete A range of ignition source / target set FDS modeling was used for fire scenario Selection and combinations has been represented for evaluations in the Cable Spreading Rooms and Analysis unscreened PAUs. These combinations Switchgear Rooms. In both cases, thermocouple are identified in relevant calculation location was adjusted as identified in F&O sheets for unscreened PAUs. In some FSS-D3-02. For the CSR, consequences were 37
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PRA TECHNICAL ADEQUACY Table B-I Fire PRA Peer Review - Facts and Observations - Findings F&O ID SR Topic Status Finding Disposition PAUs, sub-PAUs are defined and divided into scenarios based on mitigation damage from a potential fire within the potential. First, ifthe scenario was suppressed by sub-PAU isaddressed. However, it is the Halon system then the limit of damage was not clear how or why damage would be based on what was predicted by FDS in terms of limited to the specified sub-PAU temperature and energy. If it was unsuppressed it because there are no physical barriers went to total room burn, which assumes failure of between specified sub-PAUs. The all targets in the room, regardless of the initial documentation is such that it cannot be scenario boundary. For the Switchgear Room FDS determined if the selected fire scenarios analysis, the analysis was updated to add clarity to provide reasonable assurance that the the analysis. A discussion of the application of risk contribution of each unscreened sub-PAUs has been added to Addendum 1 to CO-PAU can be characterized. Another FSS-004, Fire PRA Detailed Fire Modeling issue that influences the potential for Notebook. Damage was not limited to specified fire propagation across sub-PAU sub-PAUs. Specific examples of the treatment of boundaries isthat the temperature fire growth and the application of sub-PAUs have measurement locations specified in the been provided.
detailed FDS fire modeling evaluations do not generally coincide with locations As described in CO-FSS-004, the sub-PAU where maximum temperature are analysis included spatial information from expected (e.g., within the fire plume). walkdown, along with engineering judgment, to determine iffire sources could fail additional As a consequence, for some fire components, cables, or other combustibles, scenarios damage to targets is not potentially leading to more damage to surrounding predicted when it should be based on equipment or cables. For scenarios that leveraged the specified damage criteria. Some FDT modeling, the issue related to whether the scenarios are screened on the basis of analysis had correctly addressed the impact of temperature measurements that do not transients along the edge of a boundary interface represent conditions at targets within for a sub-PAU. A comparable consideration was the fire plume. (See F&O FSS-D3-02) also related to secondary combustion and oil fires.
This could have a significant impact on Resolution involved selection of several the potential for fire propagation across representative PAUs for a sensitivity study that sub-PAU boundaries and needs to be expanded the existing sub-PAUs and examined discussed more thoroughly. secondary ignition potential.
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PRA TECHNICAL ADEQUACY Table B-I Fire PRA Peer Review - Facts and Observations - Findings F&O ID SR Topic Status Finding Disposition FSS-A5-01 FSS-A5 Fire Scenario Complete There were indications that Calvert The PAUs were considered representative of the Selection and Cliffs had the tools and information in work performed based on several criteria. The Analysis place to properly evaluate the analysis indicated that the methods mentioned propagation of fires across the sub-PAU were indeed appropriate. Sub-PAU impacts did boundaries given no physical barriers not change from the expanded assessment and but there were no examples showing that secondary ignition was bounded by the that this evaluation was performed or existing analysis and was appropriately addressed.
any explicit descriptions of how they The analysis was incorporated into the were performed in general. The documentation for CO-FSS-004.
concern here isthat without an explicit description of the process for evaluating the spread of fires across sub-PAU boundaries with no physical barriers and detailed examples, there is the potential that in the future, new people updating the PRA may not know that they have to evaluate this.
Calvert Cliffs needs to describe their process for evaluating fire growth and propagation between sub-PAUs and as applicable, between PAUs. Specific examples of the sub-PAU fire growth need to be provided. If fire propagation from sub-PAU to sub-PAU was not treated, Calvert Cliffs needs to evaluate all sub-PAUs to determine if there is any potential for fire spread and then model the potential for spreading fires and for damage occurring across sub-PAU boundaries.
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PRA TECHNICAL ADEQUACY Table B-1 Fire PRA Peer Review - Facts and Observations - Findings F&O ID SR Topic Status Finding Disposition FSS-D2-01 FSS-D2 Fire Scenario Complete Where used, the FDS model was FDS modeling was used for fire scenario Selection and generally used with a level of grid evaluations in the Cable Spreading Rooms and Analysis resolution that was below the level of Switchgear Rooms.
grid resolution documented in the NUREG-1824 Verification and For the Cable Spreading Room FDS fire scenarios, Validation study for the FDS model. A a grid study was performed on the updated FDS validation study was not conducted to model. The study recommended a grid size that support the use of this lower level of was within the range in NUREG/CR-1 824. That grid resolution. Grid resolution has a grid size was used for CSR FDS scenario bearing on the results of FDS evaluations. The study and results were calculations. Grid resolutions outside incorporated into CO-FSS-004, Fire PRA Detailed the validation range in NUREG-1824 Fire Modeling Notebook.
should be justified and validated.
The Unit 1 27' and 45' Switchgear Rooms were Increase the level of grid resolution in updated to increase the level of grid resolution to a the FDS PAU Fire Evaluations value that is within the validation range (CO-FSS-004 R1) so that the grid documented in NUREG/CR-1824. Results resolution is within the validation range calculated in the Unit 1 FDS models were applied documented in NUREG-1824. to Unit 2. Results of the updated model are incorporated into CO-FSS-004 as Addendum 1.
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PRA TECHNICAL ADEQUACY Table B-1 Fire PRA Peer Review - Facts and Observations - Findings F&O ID SR Topic Status Finding Disposition FSS-D3-01 FSS-D3 Fire Scenario Complete This SR is not met because detailed FDS modeling was used for fire scenario FSS-B2 Selection and FDS fire modeling evaluations of PAUs evaluations in the Cable Spreading Rooms and FSS-D4 Analysis 302, 306, 311, 317, 407 and 430 Switchgear Rooms.
assume that material surfaces are "inert." As noted on p. 44 of For the Cable Spreading Room FDS fire scenarios, CO-FSS-004 R1, this assumption was the Unit I CSR was modified to include actual made "... so that no objects in the PAU material properties and sensitivity analysis. Actual or the PAU structure (walls, floor, or material properties were used in the updated ceiling) itself would absorb any heat U1CSR FDS model rather than the prior use of from the various fire scenarios, "inert" material conditions. Adiabatic conditions producing a more conservative or worst were used for any items with material properties case result for all fire scenarios' impacts that are unknown or of a high uncertainty to bound to the components and cables within the the analysis and prevent heat transfer into those PAU model. As such, no detailed objects. The CSR FDS model was executed and material properties were required to be the results compared to the baseline results. This defined in FDS for the scenarios to study was then documented in FSS-004. The function correctly." However, results were applied to Unit 2 CSR. This study specification of material surfaces as was then documented in FSS-004, Fire PRA "inert" in FDS does not prevent heat Detailed Fire Modeling Notebook.
absorption into material surfaces. On the contrary, this specification maintains The Unit 1 27' and 45' Switchgear Rooms were material surfaces at ambient updated to specify representative material temperature in FDS, which tends to properties as referenced by NUREG 1805. This maximize heat absorption into these adjustment enabled the analysis to obtain more surfaces. realistic estimates of environmental conditions for these fire scenarios. Results calculated in the Unit 1 FDS models were applied to Unit 2. Results of the updated model are incorporated into CO-FSS-004 as Addendum 1.
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PRA TECHNICAL ADEQUACY Table B-I Fire PRA Peer Review - Facts and Observations - Findings F&O ID SR Topic Status Finding Disposition FSS-D3-01 FSS-D3 Fire Scenario Complete To prevent heat absorption into material FSS-B2 Selection and surfaces, they should have been FSS-D4 Analysis specified as "adiabatic" rather than as "inert." The "inert" parameter in FDS maximizes heat transfer to surfaces rather than minimize it. This can result in lower calculated gas temperatures.
Specify materials surfaces as
'adiabatic" rather than as "inert" in FDS to prevent them from absorbing heat in order to achieve the stated goal of producing a more conservative or worst case result. This may prove to be overly conservative, in which case specification of realistic material properties could be used to achieve more realistic estimates of environmental conditions for these fire scenarios.
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PRA TECHNICAL ADEQUACY Table B-1 Fire PRA Peer Review - Facts and Observations - Findings F&O ID SR Topic Status Finding Disposition FSS-D3-02 FSS-D3 Fire Scenario Complete Temperature measurement locations FDS modeling was used for fire scenario FSS-A5 Selection and specified in the detailed FDS fire evaluations in the Cable Spreading Rooms and Analysis modeling evaluations do not generally Switchgear Rooms.
coincide with locations where maximum temperature are expected (e.g., within For the Cable Spreading Room FDS fire scenarios, the fire plume). As a consequence, for new measurement devices were included in the some fire scenarios damage to targets updated UICSR FDS model. The thermocouples is not predicted when it should be based were placed directly above the fire source in the on the specified damage criteria. Some updated FDS model and the scenarios re-scenarios are screened on the basis of evaluated. The results were applied to Unit 2 temperature measurements that do not CSR. This study and the results were then represent conditions at targets within documented in FSS-004, Fire PRA Detailed Fire the fire plume. Modeling Notebook.
Re-run FDS simulations with The Unit 1 27' and 45' SWGR rooms were updated temperature measurement probes to alter the location of the thermocouples such that located within the fire plume or use the centerline plume temperature was recorded other fire modeling tools such as FDTs and used to determine target impacts. Results to calculate fire plume temperatures for calculated in the Unit 1 FDS models were applied these scenarios, to Unit 2. Results of the updated model are incorporated into CO-FSS-004 as Addendum 1.
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PRA TECHNICAL ADEQUACY Table B-1 Fire PRA Peer Review - Facts and Observations - Findings F&O ID SR Topic Status Finding Disposition FSS-D8-01 FSS-D8 Fire Scenario Complete Fire detection timing is evaluated for FDS modeling was used for fire scenario Selection and detailed fire modeling cases that use evaluations in the Cable Spreading Rooms and Analysis FOS. This fire detection timing isthen Switchgear Rooms.
used to estimate automatic fire For the updated Cable Spreading Room FDS fire suppression timing and fire brigade scenarios, cable tray obstructions were placed in response timing for these scenarios. the ceiling area of the updated U1CSR FDS model.
However, the fire detection timing is Additional thermocouple and heat flux data based on modeling that does not recording devices were added to the U1CSR include obstructions located beneath model under the new cable tray obstructions in the the ceiling that could have an impact on vicinity of the fire source. The scenarios were re-fire detector response. The fire evaluated. The results were applied to Unit 2. A detection timing is also based on an sensitivity study was also performed. The study unjustified assumption regarding the and new scenario results were incorporated into type of smoke detectors installed in the CO-FSS-004, Fire PRA Detailed Fire Modeling affected PAUs. Obstructions to the flow Notebook.
of fire gases can have an impact on smoke concentrations and velocities, The Unit 1 27' and 45' SWGR rooms were also which in turn influence smoke detector updated to include significant obstructions such as response. Without including such cable trays and beam pockets within the obstructions in fire modeling switchgear rooms. Results calculated in the Unit 1 simulations, their impact on fire FDS models were applied to Unit 2. Results and detection times is not evaluated. details of this analysis are documented in CO-FSS-004 as Addendum 1.
Include obstructions located beneath the ceiling for the affected fire scenarios in order to evaluate their impact on fire detection timing. Provide justification for the selection of the type of smoke detector specified in the FDS simulations for these fire scenarios.
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PRA TECHNICAL ADEQUACY Table B-I Fire PRA Peer Review - Facts and Observations - Findings F&O ID SR Topic Status Finding Disposition FSS-F3-01 FSS-F3 Fire Scenario Complete To achieve CC Il/111 for this SR, a The Turbine Building was reviewed for potential Selection and quantitative assessment of the risk of fire scenarios where structural steel can be Analysis the selected fire scenarios involving a) adversely affected. From the scenarios examined, exposed structural steel and b) the those that can damage structural steel were presence of a high-hazard fire sources selected for further analysis. The frequency, must be completed consistent with the severity factor and non-suppression probability of FQ requirements including the collapse each scenario were developed and included in the of the exposed structural steel and any Structural Failure Analysis Notebook. These attendant damage. Such an impacts were then added to FRANX database and assessment has not been done or was quantified as part of the final Fire PRA risk not documented in a readily discernible quantification in Fire Quantification Notebooks manner. This has a potential impact on CO-FRQ-001 and CO-FRQ-002.
fire risk quantification.
Complete a quantitative assessment of the risk of the selected exposed structural steel fire scenarios consistent with the FQ requirements.
FSS-G4-01 FSS-G4 Fire Scenario Complete An assessment of the effectiveness, Generic probabilities were used for credited Selection and reliability and availability of credited passive fire barrier features in the multi-Analysis passive fire barrier features has not compartment analysis. At Calvert Cliffs, the fire been documented in the multi- barriers are verified to be effective through test compartment analysis. To achieve a procedures. An unreliability value was applied to CC II capability assessment, the all normally closed doors that represents the effectiveness, reliability and availability probability of the door being propped open given a of credited passive fire barrier features fire in the exposing compartment. The probability must be assessed. of finding a failed sealed wall penetration is assumed to be very small to warrant propagation Assess the effectiveness, reliability and scenarios. A discussion of the effectiveness, availability of credited passive fire reliability, and availability of fire barriers was added barrier features and document this to CO-FSS-008, Calvert Fire PRA Multi-assessment. Compartment Analysis.
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PRA TECHNICAL ADEQUACY Table B-I Fire PRA Peer Review - Facts and Observations - Findings F&O ID SR Topic Status Finding Disposition FSS-G5-01 FSS-G5 Fire Scenario Complete The effectiveness, reliability and Active fire barriers were evaluated as effective in Selection and availability of credited active fire barrier studies used to support Appendix R analysis. An Analysis features have not been quantified in the unreliability value has been applied to all normally multi-compartment analysis. To open, self closing dampers and doors; A achieve a CC II capability assessment, discussion of the effectiveness of credited active the effectiveness, reliability and fire barriers was added to CO-FSS-008, Calvert availability of credited active fire barrier Fire PRA Multi-Compartment Analysis.
features must be quantified.
Quantify the effectiveness, reliability and availability of credited active fire barrier features and document this assessment.
HRA-B2-01 HRA-B2 Human Complete Improve documentation of the adverse CO-HRA-001, Fire Human Reliability notebook, Reliability operator actions needed to address the was updated to detail the adverse operator actions Analysis impact of grounded or shorted electrical added to the model following the fire AOP review buses that might have an impact on process. Table 3 was added to Section 2.2 other plant buses ifnot isolated and re detailing each basic event, set to true (1.0) used in energized in the areas identified. Very the model to annotate the adverse operator actions difficult to find the information within the in the model. These include actions to de-energize HRA notebook alone, because the electrical busses to isolate them from potential actions are modeled as inputs to shorts and grounds. Table 2 shows the HFEs FRANX. added to the model as part of the AOP review, including actions to restore AC power to busses Provide new tables listing the actions lost due to fire failure sequences.
considered or references to specific locations.
HRA-E1-01 HRA-E1 Human Complete Documentation for what was done was CO-HRA-001, Fire Human Reliability Notebook, Reliability very good, however, the details for not was updated detailing the Alarm Response Analysis selecting any spurious alarms is not Procedure review process. Table 12 was clear. The documentation of the expanded to show the ARP review of alarm impact adverse actions put into the model as and operator interview notes for CR annunciators "true" are not in the HRA report, actions that could result in a manual reactor trip. No identified in the cutset reviews are not annunciators were identified that would cause the 46
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PRA TECHNICAL ADEQUACY Table B-1 Fire PRA Peer Review - Facts and Observations - Findings F&O ID SR Topic Status Finding Disposition clearly identified, rational for not using operator to terminate a systems or components specific HFEs in the RCP trip actions, operation based solely on the alarm itself, but for identifying actions from procedures several were identified that could potentially result and the process for assigning in the operator tripping the Unit unnecessarily.
uncertainty range for the combos.
Doesn't permit verification of the rational CO-HRA-001 was also updated to detail the for judgments made in deciding what is adverse operator actions added to the model in and out of the Fire HRA. Also, from following the fire AOP review process. Table 3 the calculation viewpoint the need to was added to Section 2.2 detailing each basic know the use of all manpower event, set to true (1.0) used in the model to requirements during early time after fire annotate the adverse operator actions in the initiator for dependency analysis. model. These include actions to de-energize electrical busses to isolate them from potential Enhance documentation of the specific shorts and grounds. Table 2 shows the HFEs issues needed to reproduce the added to the model as part of the AOP review, assumptions and calculations used in including actions to restore AC power to busses the HRA. lost due to fire failure sequences.
New HFEs added as part of the cutset review process are identified in Table I of CO-HRA-001, Fire Human Action Reliability notebook. These are annotated with "identified during the development of the PRM Notebook." The cutset reviews are described in CO-QNS-001, Fire PRA Quantitative Screening Notebook. A new dependency analysis was performed after the new HFEs were added to the model, ensuring new dependency combinations are considered.
Additional information was added to Table 1 of the Human Reliability Analysis Notebook, CO-HRA-001, detailing why each HFE was either retained or removed. For example, event FGAFWOSGTRISOL, Operator Feeds Affected SG with SGTR to Assure Heat Removal, was "Not 47
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PRA TECHNICAL ADEQUACY Table B-I Fire PRA Peer Review - Facts and Observations - Findings F&O ID SR Topic Status Finding Disposition retained for fire scenarios, because these actions are SGTR specific. Modeling was not necessary to ensure these actions did not appear in the cut sets, because the SGTR initiator is not being used for fire scenarios."
Combination event multipliers are used in cutsets of multiple HEP actions to account for dependencies between HEP actions. To account for the uncertainty in HEP actions, an uncertainty parameter is added to the HEP action. When performing uncertainty analysis, the uncertainty parameters for combination events is increased proportionally when they are multiplied by the combination event multipliers.
Based on interviews, there are sufficient non-control room personnel for fire recovery actions.
Appendix D of CO-HRA-001 notes that there are no control room operators assigned to the fire brigade.
There were no identified staffing issues or interferences between operators performing fire recovery actions and members of the fire brigade.
FQ-Al-01 FQ-Al Fire Risk Complete Treatment of 0 CCDPs scenarios is not The fire risk quantification process has been Quantification clear and appears to result in an updated in notebooks CO-FRQ-001 and CO-FRQ-underestimate of total risk (the 002 to address the issue with FRANX fire underestimate appears to be small scenarios having a zero conditional probability for based on the sensitivity evaluations CDF and LERF.
performed):
1 - with respect to opposite unit 1. When documented analysis shows that selected quantification, use CCDP for reactor trip fire scenarios for one unit are screened from initiator unless confirmation of no trip is impact for the opposite unit (typically, no trip would documented; be initiated), then that scenario may be excluded 2 - address use of 0 CCDP for control from the opposite unit's fire risk quantification.
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PRA TECHNICAL ADEQUACY Table B-1 Fire PRA Peer Review - Facts and Observations - Findings F&O ID SR Topic Status Finding Disposition room HVAC loss scenarios, apply Otherwise, a nominal conditional probability, as CCDP consistent with control room described in item 3 below, would apply.
abandonment
- 2. F&O PRM-B3-01 identifies the concern with loss 3 - for scenarios with limited impact with of Control Room HVAC with control room a 0 CCDP, due to cutsets below abandonment. As discussed in more detail with truncation limit, apply a baseline CCDP the resolution to PRM-B3-01, subsequent based on reactor trip initiator investigation revealed that loss of CR HVAC is not expected to cause abandonment by the operations More than 50% of the scenarios have a staff of the control room due to high temperatures.
0 CCDP but no clear discussion of the Loss of CR HVAC and subsequent temperature basis for the 0 CCDP is provided, increases may adversely affect operator responses, and the model reflects degradation of Treatment of 0 CCDPs scenarios: human actions with loss of CR HVAC. CO-SY-030, Control Room HVAC PRA System Notebook, was 1 - with respect to opposite unit updated to include this discussion.
quantification, use CCDP for reactor trip initiator unless confirmation of no trip is 3. The new quantification process described in the documented; FRQ notebooks is to assure a nominal conditional 2 - address use of 0 CCDP for control value is calculated for these low significant room HVAC loss scenarios, apply scenarios by 1) recalculating the zero-conditional CCDP consistent with control room scenarios at a lower truncation value to assure abandonment resolution in the scenario cutset file and conditional 3 - for scenarios with limited impact with probabilities , and/or to 2) use a baseline a 0 CCDP, due to cutsets below conditional probability for CDF and LERF for the truncation limit, apply a baseline CCDP internal events reactor trip initiating vent - IEOPT based on reactor trip initiator for Unit 1 or IEOPT-2 for Unit 2 FQ-Bl-01 FQ-B1 Fire Risk Complete We observed zero CCDPs for some The fire risk quantification process has been Quantification PAU CDF and LERF values in the updated in notebooks CO-FRQ-001 and CO-FRQ-FRANX tables (e.g., PAU 512) which 002 to address the issue with FRANX fire eliminated loss of HVAC to the MCR as scenarios having a zero conditional probability for a potential MCR abandonment CDF and LERF.
sequence. Treatment of 0 CCDPs scenarios: 1. When documented analysis shows that selected 49
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PRA TECHNICAL ADEQUACY Table B-I Fire PRA Peer Review - Facts and Observations - Findings F&O ID SR Topic Status Finding Disposition fire scenarios for one unit are screened from 1 - with respect to opposite unit impact for the opposite unit (typically, no trip would quantification, use CCDP for reactor trip be initiated), then that scenario may be excluded initiator unless confirmation of no trip is from the opposite unit's fire risk quantification.
documented; Otherwise, a nominal conditional probability, as 2 - address use of 0 OCOP for control described in item 3 below, would apply.
room HVAC loss scenarios, apply OCOP consistent with control room 2. F&O PRM-B3-01 identifies the concern with loss abandonment (F&O FQ-Al-01 (fl)) of Control Room HVAC with control room 3 - for scenarios with limited impact with abandonment. As discussed in more detail with a 0 CCDP, due to cutsets below the resolution to PRM-B3-01, subsequent truncation limit, apply a baseline CCDP investigation revealed that loss of CR HVAC is not based on reactor trip initiator expected to cause abandonment by the operations Allowing zero CCDPs allows scenarios staff of the control room due to high temperatures.
in the fire model to quantify with no Loss of CR HVAC and subsequent temperature contribution to the COF or LERF value increases may adversely affect operator and this under represents those responses, and the model reflects degradation of frequencies especially when human actions with loss of CR HVAC. CO-SY-030, considering delta risk evaluations. Control Room HVAC PRA System Notebook, was updated to include this discussion.
Replace the zero entries with the lowest CCPD for a plant trip with only random 3. The new quantification process described in the failures of the safety equipment as in FRQ notebooks is to assure a nominal conditional the internal events model. We value is calculated for these low significant discussed this with the Calvert Cliffs scenarios by 1) recalculating the zero-conditional PRA team and some of the zeros are scenarios at a lower truncation value to assure due to fire areas in one unit potentially resolution in the scenario cutset file and conditional contributing to the CCDP of the probabilities, and/or to 2) use a baseline opposite unit. With the exception of conditional probability for CDF and LERF for the these cases a method for handling the internal events reactor trip initiating vent - IEOPT zeros needed to be developed and for Unit 1 or IEOPT-2 for Unit 2 applied in the frequency quantifications.
50
ATTACHMENT (2)
PRA TECHNICAL ADEQUACY Table B-2 Fire PRA - Category I Summary SR Topic Status PP-B3 2012 Peer Review: SR Not Met Now: Met Cat Il/111 The containment is partitioned into 2 PAUs. There CO-PP-001, Calvert Cliffs Fire PRA Plant Partitioning Notebook, was updated to include are intervening combustibles and this was accounted an analysis that justifies the partitioning of the containment into two plant partitioning units for in the PRA by treating the 20 feet as an overlap with a 20-foot spatial separation (known as the buffer zone). The only potential region and failing components affected into both intervening combustibles in this buffer zone were identified as qualified cables that were PAUs. There is no justification given for the 20 verified to be encased within marinate covered raceways. The covers prevent the cables assumption. The turbine deck is continuous from unit from becoming potential combustibles and therefore are not considered intervening 1 to unit 2. This area is divided into 2 PAUs, TURB1 combustibles.
and TURB2, but there is no discussion for the basis of the partitioning. The unit 1 and unit 2 Turbine Deck was walked down to assess for the acceptability of the Appendix R partitioning into distinct PAUs. The boundary was assessed to have at least a Associated F&O: PP-B3-01 20-foot separation between potential ignition sources and potential targets, assessed for intervening combustibles, and the Turbine deck volume assessed for damaging hot gas layer development. The partitioning was found acceptable and consistent with NUREG/CR-6850, Section 1.5.2, where main turbine decks are typical applications where spatial separation have been credited.
PP-B5 2012 Peer Review: SR Met: (CC-I) Now: Met Cat Il/111 The water curtain in the CCW room was credited as The Component Cooling Water room water curtain is an approved Appendix R exemption, an active fire barrier. The justification was that the as identified in the exemption issued by the NRC in response to Calvert Cliffs exemption water curtain was part of the original regulatory fire request ER820816. The validity of crediting CCW Room Water Curtains is discussed in protection program. This meets CAT 1, but needs Southwest Research Institute Report No. 01-0763-201. A reference to the Southwest enhancement for CAT Il/111 Research Institute report was added to CO-PP-001, Plant Partitioning Notebook.
Associated F&O: PP-B5-01 51
ATTACHMENT (2)
PRA TECHNICAL ADEQUACY Table B-2 Fire PRA - Category I Summary SR Topic Status PP-B6 2012 Peer Review: SR Not Met Now: Met Cat 1/11/111 The containment has a 20 foot area that overlaps between the E and W section. The overlap is CO-PP-001, Calvert Cliffs Fire PRA Plant Partitioning Notebook, was updated to include specifically addressed in the PP notebook. The an analysis that justifies the partitioning of the containment into two plant partitioning units standard does not allow for an overlap, with a 20-foot spatial separation (known as the buffer zone). The only potential intervening combustibles in this buffer zone were identified as qualified cables that were Associated F&O: PP-B3-01 verified to be encased within marinate covered raceways. The covers prevent the cables from becoming potential combustibles and therefore are not considered intervening combustibles.
CS-B1 2012 Peer Review: SR Met: (CC I) Now: Met Cat Il/111 Supporting Requirement CS-B1 met with a category I. The breaker coordination study has been completed. As described in ECP-1 3-000321, A breaker coordination study is currently being Form 12, Engineering Evaluation, all PRA common power supplies are assumed to meet -
performed and is planned to be incorporated in the or will meet - the coordination requirements of NFPA 805, except as noted in CO-CS-001, future. See Fact and Observation CS-B1-01. Fire PRA Cable Selection Notebook. As described in the cable selection notebook, two 120VAC lighting panels are not validated as coordinated, and these panels are assumed Associated F&O: CS-B1-01 to fail for all Fire PRA scenarios. Also, as described in the PRA notebook a breaker for 480V motor control center MCC101 BT has not been validated as coordinated. This breaker, 52-10150, is modeled so that a fire-induced electrical fault on the breaker's power cabling will fail MCC101 BT. Finally, the notebook identifies that selected 120V power panels have coordination issues, but that these will be addressed by design changes and referenced in Attachment S - Modifications and Implementation Items.
52
ATTACHMENT (2)
PRA TECHNICAL ADEQUACY Table B-2 Fire PRA - Category I Summary SR Topic Status PRM-B3 2012 Peer Review: SR Not Met Now: Met Cat 1/11/111 No new initiating events were identified in the course Loss of Control Room HVAC can affect the operability and availability of equipment in the of the fire PRA model generation. control room and cable spreading room. As described in Calvert PRA System Analysis Notebooks CO-SY-002, CO-SY-017, and CO-SY-030, loss of HVAC is modeled to have the The failure of the control room HVAC does not lead to effect of increasing the failure rate of 120VAC and 125VDC instruments and controls in a control room abandonment CCDP (1.0 or other the cable spreading room. For the control room, degradation of the 125VDC system is value justified by analysis as corresponding to used as a conservative surrogate for control room I&C degradation.
shutdown from outside the control room)
Loss of Control Room HVAC and subsequent temperature increases may adversely affect Associated F&O: PRM-B3-01 operator responses. The model reflects degradation of human actions by the degradation of the 125VDC system used for instruments and controls. Loss of Control Room HVAC is not expected to cause abandonment by operations staff of the control room due to high temperatures. On complete loss of HVAC with no mitigation, such as no use of emergency fans, calculation CA02725 shows a CR temperature of 123 deg F at 24-hours.
While this is a challenging environment, this temperature is assessed as insufficient to solely drive a complete CR abandonment scenario. NUREG/CR-6738 describes operational experience where operators will continue to occupy the control room even under severe environments.
Operations staff says that in consideration of high temperatures in the control room, that Operations would do what was needed to keep the cores safe and covered. The site safety director says that for a temperature of 123 deg F, the site would implement a mitigation strategy which would include stay-times, assessment of individuals for heat-related conditions, use of ice vests, and call-in of additional qualified operations staff to rotate into the control room.
The above discussion was included in CO-SY-030, Control Room HVAC PRA System Notebook.
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ATTACHMENT (2)
PRA TECHNICAL ADEQUACY Table B-2 Fire PRA - Category I Summary SR Topic Status PRM-B4 2012 Peer Review: SR Not Met Now: Met Cat 1/11/111 See PRM-B3-01 F&O, not met due to no new The potential new initiator has been assessed (failure of CR HVAC leading to CR initiators identified and the identification of a potential abandonment as discussed in PRM-B3). Loss of Control Room HVAC and subsequent new initiator that was not quantified in the fire PRA temperature increases may adversely affect operator responses. The model reflects model. degradation of human actions by the degradation of the 125VDC system used for instruments and controls. Loss of Control Room HVAC is not expected to cause Associated F&O: PRM-B3-01 abandonment by operations staff of the control room due to high temperatures. On complete loss of HVAC with no mitigation, such as no use of emergency fans, calculation CA02725 shows a CR temperature of 123 deg F at 24-hours. While this is a challenging environment, this temperature is assessed as insufficient to solely drive a complete CR abandonment scenario. NUREG/CR-6738 describes operational experience where operators will continue to occupy the control room even under severe environments.
Operations staff says that in consideration of high temperatures in the control room, that Operations would do what was needed to keep the cores safe and covered. The site safety director says that for a temperature of 123 deg F, the site would implement a mitigation strategy which would include stay-times, assessment of individuals for heat-related conditions, use of ice vests, and call-in of additional qualified operations staff to rotate into the control room.
The above discussion was included in CO-SY-030, Control Room HVAC PRA System Notebook.
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ATTACHMENT (2)
PRA TECHNICAL ADEQUACY Table B-2 Fire PRA - Category I Summary SR Topic Status FSS-A5 2012 Peer Review: SR Not Met Now: Met Cat 1/11 A range of ignition source / target set combinations FDS modeling was used for fire scenario evaluations in the Cable Spreading Rooms and has been represented for unscreened PAUs. These Switchgear Rooms. In both cases, thermocouple location was adjusted as identified in combinations are identified in relevant calculation F&O FSS-D3-02. For the CSR, consequences were divided into scenarios based on sheets for unscreened PAUs (filenames mitigation potential. First, if the scenario was suppressed by the Halon system then the RSC-CALKNX-201 1-xxx.pdf). However, it is not clear limit of damage was based on what was predicted by FDS in terms of temperature and how the potential for spreading fires and for fire and energy. If it was unsuppressed it went to total room burn, which assumes failure of all smoke spread between sub-PAUs is addressed and targets in the room, regardless of the initial scenario boundary. For the Switchgear Room consequently it cannot be determined if the selected FDS analysis, the analysis was updated to add clarity to the analysis. A discussion of the fire scenarios provide reasonable assurance that the application of sub-PAUs has been added to Addendum 1 to CO-FSS-004, Fire PRA risk contribution of each unscreened PAU can be Detailed Fire Modeling Notebook. Damage was not limited to specified sub-PAUs.
characterized. Specific examples of the treatment of fire growth and the application of sub-PAUs has been provided.
Associated F&O: FSS-A5-01 As described in CO-FSS-004, the sub-PAU analysis included spatial information from walkdown, along with engineering judgment, to determine if fire sources could fail additional components, cables, or other combustibles, potentially leading to more damage to surrounding equipment or cables. For scenarios that leveraged FDT modeling, the issue related to whether the analysis had correctly addressed the impact of transients along the edge of a boundary interface for a sub-PAU. A comparable consideration was also related to secondary combustion and oil fires. Resolution involved selection of several representative PAUs for a sensitivity study that expanded the existing sub-PAUs and examined secondary ignition potential.
The PAUs were considered representative of the work performed based on several criteria. The analysis indicated that the methods mentioned were indeed appropriate.
Sub-PAU impacts did not change from the expanded assessment and that secondary ignition was bounded by the existing analysis and was appropriately addressed. The analysis was incorporated into the documentation for CO-FSS-004.
55
ATTACHMENT (2)
PRA TECHNICAL ADEQUACY Table B-2 Fire PRA - Category I Summary SR Topic Status FSS-D3 2012 Peer Review: SR Not Met Now: Met Cat. II.
This SR is not met for multiple reasons. First, FDS modeling was used for fire scenario evaluations in the Cable Spreading Rooms and detailed FDS fire modeling evaluations of PAUs 302, Switchgear Rooms.
306, 311, 317, 407 and 430 assume that material surfaces are "inert." As noted on p. 44 of CO-FSS-004 Material Properties R1, this assumption was made " so that no objects in the PAU or the PAU structure (walls, floor, or ceiling) For the Cable Spreading Room FDS fire scenarios, the Unit 1 CSR was modified to itself would absorb any heat from the various fire include actual material properties and sensitivity analysis. Actual material properties were scenarios, producing a more conservative or worst used in the updated U1CSR FDS model rather than the prior use of "inert" material case result for all fire scenarios' impacts to the conditions. Adiabatic conditions were used for any items with material properties that are components and cables within the PAU model. As unknown or of a high uncertainty to bound the analysis and prevent heat transfer into such, no detailed material properties were required to those objects. The CSR FDS model was executed and the results compared to the be defined in FDS for the scenarios to function baseline results. This study was then documented in FSS-004. The results were applied correctly." However, specification of material to Unit 2 CSR. This study was then documented in FSS-004, Fire PRA Detailed Fire surfaces as "inert" in FDS does not prevent heat Modeling Notebook.
absorption into material surfaces. On the contrary, this specification maintains material surfaces at The Unit 1 27' and 45' Switchgear Rooms were updated to specify representative material ambient temperature in FDS, which tends to properties as referenced by NUREG 1805. This adjustment enabled the analysis to maximize heat absorption into these surfaces. To obtain more realistic estimates of environmental conditions for these fire scenarios.
meet the specified goal of preventing heat absorption Results calculated in the Unit 1 FDS models were applied to Unit 2. Results of the into material surfaces, they should have been updated model are incorporated into CO-FSS-004 as Addendum 1.
specified as "adiabatic" rather than as "inert." (See Finding FSS-D3-01) Temperature Measurement Locations Second, temperature measurement locations For the Cable Spreading Room FDS fire scenarios, new measurement devices were specified in the detailed FOS fire modeling included in the updated U1CSR FDS model. The thermocouples were placed directly evaluations do not generally coincide with locations above the fire source in the updated FDS model and the scenarios re-evaluated. The where maximum temperatures are expected (e.g., results were applied to Unit 2 CSR. This study and the results were then documented in within the fire plume). As a consequence, for some FSS-004, Fire PRA Detailed Fire Modeling Notebook.
fire scenarios damage to targets is not predicted when it should be based on the specified damage The Unit 1 27' and 45' SWGR rooms were updated to alter the location of the criteria. (See Finding FSS-D3-02) thermocouples such that the centerline plume temperature was recorded and used to determine target impacts. Results calculated in the Unit 1 FDS models were applied to Associated F&Os: FSS-D3-01 and FSS-D3-02 Unit 2. Results of the updated model are incorporated into CO-FSS-004 as Addendum 1.
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ATTACHMENT (2)
PRA TECHNICAL ADEQUACY Table B-2 Fire PRA - Category I Summary SR Topic Status FSS-F3 2012 Peer Review: SR Met: (CC I) Now: Met Cat. Il/111 A number of potential scenarios are selected and a This subject of this SR are fire-induced damage to structural steel. As described in qualitative assessment of the associated risk is CO-FSS-005, Calvert Cliffs Fire PRA Structural Failure Analysis Notebook, the un-performed for the selected fire scenarios, screened structural steel scenarios are in the Turbine Building.
Associated F&O: FSS-F3-01 The Turbine Building analysis was reviewed for potential fire scenarios where structural steel can be adversely affected. From the scenarios examined, those that can damage structural steel were selected for further analysis. The frequency, severity factor and non-suppression probability of each scenario were developed and included in the Structural Failure Analysis Notebook. These impacts were then added to FRANX database and quantified as part of the final Fire PRA risk quantification in Fire Quantification Notebooks CO-FRQ-001 and CO-FRQ-002.
FSS-G4 2012 Peer Review: SR Met: (CC I) Now: Met Cat. II Passive fire barriers are credited in the multi- Generic probabilities were used for credited passive fire barrier features in the multi-compartment analysis consistent with fire resistance compartment analysis. At Calvert Cliffs, the fire barriers are verified to be effective ratings, but the effectiveness, reliability and through test procedures. An unreliability value was applied to all normally closed doors availability of credited passive fire barriers have not that represents the probability of the door being propped open given a fire in the exposing been assessed. compartment. The probability of finding a failed sealed wall penetration is assumed to be very small to warrant propagation scenarios. A discussion of the effectiveness, reliability, Associated F&O: FSS-G4-01 and availability of fire barriers was added to CO-FSS-008, Calvert Fire PRA Multi-Compartment Analysis.
FSS-G5 2012 Peer Review: SR Met: (CC I) Now: Met Cat. Il/111 The effectiveness, reliability and availability of active Active fire barriers were evaluated as effective in studies used to support Appendix R fire barrier elements has been assessed qualitatively, analysis. An unreliability value has been applied to all normally open, self closing but has not been quantified. dampers and doors; A discussion of the effectiveness of credited active fire barriers was added to CO-FSS-008, Calvert Fire PRA Multi-Compartment Analysis.
Associated F&O: FSS-G5-01 57
ATTACHMENT (2)
PRA TECHNICAL ADEQUACY C. Total CDF, LERF and RG 1.174 The PRA scope that is currently developed in accordance with the ASME/ANS RA-Sa-2009 standard and Reg. Guide 1.200 are fire and internal events. Table C-1 includes the values for the following:
" Fire: The CDF and LERF are obtained from the Fire PRA. These risk metrics take into account the proposed plant modifications credited for the NFPA 805 transition. Section B discusses PRA quality for the fire model.
- Internal events (including internal floods): CDF and LERF used to evaluate the total plant risk are based on preliminary estimates, taking into account the proposed plant modifications credited for the NFPA 805 transition. Section A discusses PRA quality for the internal events model.
The following external hazards were screened in the IPEEE as not significantly influencing total risk. There is no PRA model for these hazards:
- External Flooding,
- Transportation accidents, and
- Industrial Accidents.
Calvert Cliffs does not have a low power or shutdown PRA model.
Of the hazards evaluated in the Individual Plant Examination of External Events (IPEEE),
tornadoes/high winds and seismic events were of note. Table C-1 uses the IPEEE CDF values for tornadoes/high winds and seismic events, and for these, Table C-1 uses an estimate of the LERF values based on a reasonable 13% fraction of CDF for the following:
- Seismic events: PRA risk for seismic events is based upon IPEEE values. Since the Calvert Seismic PRA has not been updated since the IPEEE, the results of the seismic PRA must be qualitatively evaluated for impact on the STI extension.
The NRC evaluated seismic risk generically in ML100270756. As shown for the 2008 Seismic Hazard Curves in Table D-1 of ML100270756, the "weakest link" estimated CDF risk for Calvert Cliffs Unit 1 is 1.OE-05 and for Unit 2 is 1.2E-05. These updated CDF values from ML100270756 are bounded by the IPEEE results presented in Table C-1.
The Calvert Cliffs SPRA was performed using the methodology outlined in NUREG-1407.
The Seismic hazard curves used are those developed for CCNPP by Lawrence Livermore National Laboratory (LLNL) and provided in NUREG-1488. In late 2013 EPRI provided Calvert Cliffs a new Ground Motion Response Spectrum (GRMS). A preliminary comparison of the new Calvert Cliffs GMRS to the Calvert Cliffs IPEEE HCLPF Spectra (IHS) shows that the GMRS is lower than the IHS at all frequencies. This is true for GMRS comparisons utilizing both the mean and median hazards as inputs in the IHS development. Based on these comparisons, the new GMRS is not expected to have a significant impact on seismic CDF at Calvert Cliffs.
" Wind events: PRA risk for tornadoes and high winds are based upon IPEEE values.
Calvert Cliffs has maintained and updated a high wind PRA model in order to perform risk assessment of tornado missile impacts and hurricane force winds. Although this model has not been peer reviewed in compliance with the ASME/ANS standard, the model is 58
ATTACHMENT (2)
PRA TECHNICAL ADEQUACY based upon accepted methodology and utilizes the ASME/ANS compliant internal events model. A recent quantification of the wind initiating events using the updated internal events model estimates CDF risk for Calvert Cliffs Unit 1 at 9.4E-07 and for Unit 2 at 9.4E-
In addition, the tornado/high wind and the seismic results in Table C-1 do not credit risk reduction for the NFPA 805 modifications. This further conservatively bounds the summary of total plant risk values in the table.
Table C Summary of Total Plant Risk for Calvert Cliffs Unit I (Irx-yr) Unit 2 (Irx-yr)
Event Type CDF LERF CDF LERF Comments Fire 3.2E-05 3.2E-06 3.6E-05 4.4E-06 Fire PRA quantification with NFPA 805 modifications credited.
Internal Events 1.3E-05 1.3E-06 1.OE-05 1.3E-06 Internal Events PRA quantification (including internal with NFPA 805 modifications floods) credited.
Seismic Events 1.3E-05 1.7E-06 15E-05 2.OE-06 IPEEE for CDF. Estimate for LERF based on 13% of CDF.
Tornadoes/High 4.4E-06 5.7E-07 4.4E-06 5.7E-07 IPEEE for CDF. Estimate for LERF Winds based on 13% of CDF.
Plant-Level 6.2E-05 6.8E-06 6.5E-05 8.3E-06 Total The data from Table C-1 is derived either from an ASME/ANS compliant PRA model or from reasonable or bounding PRA analysis for models that have not had an ASME/ANS peer review.
There is high confidence that the overall CDF is less than the Regulatory Guide 1.174 limits for total plant risk of 1 E-04 per year and LERF is less than 1E-05 per year 59
ATTACHMENT (3)
MARKED-UP TECHNICAL SPECIFICATION PAGES Calvert Cliffs Nuclear Power Plant, LLC May 1, 2014
ATTACHMENT (3)
MARKED-UP TECHNICAL SPECIFICATION PAGES Insert 1 In accordance with the Surveillance Frequency Control Program Insert 2 5.5.19 Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.
- a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.
- b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk Informed Method for Control of Surveillance Frequencies," Revision 1.
- c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.
1
Definitions
- 1.1 1.1 Definitions verification have been previously reviewed and approved by the NRC.
SHUTDOWN MARGIN (SDM) SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full length control element assemblies (CEAs) (shutdown and regulating) are fully inserted except for the single CEA of highest reactivity worth, which is assumed to be fully withdrawn. However, with all CEAs verified fully inserted by two independent means, it is not necessary to account for a stuck CEA in the SDM calculation. With any CEAs not capable of being fully inserted, the reactivity worth of these CEAs must be accounted for in the determination of SDM.
THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
CALVERT CLIFFS - UNIT 1 1.1-6 Amendment No. 286 CALVERT CLIFFS - UNIT 2 Amendment No. 263
SDM 3.1.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.1.1 Verify SDM is within limits specified in the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> COLR.
SR 3.1.1.2 ------------------ NOTE ---------------
Only required in MODE 5 with pressurizer level < 90 inches.
Verify Reactor Coolant System level is above Once within the bottom of the hot leg nozzles. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after achieving MODE 5 with pressurizer level
< 90 inches AND 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter CALVERT CLIFFS - UNIT 1 3.1.1-3 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
CEA Alignment 3.1.4 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME F. Required Action and F.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition C, D, or E not met.
OR One or more CEAs untrippable.
OR Two or more CEAs misaligned by
> 15 inches.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.1 Verify the indicated position of each CEA to Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> be within 7.5 inches of all other CEAs in following any its group. CEA movement of
> 7.5 inches AND T -e-I SR 3.1.4.2 Verify the CEA motion inhibit is OPERABLE. q-i' CALVERT CLIFFS - UNIT 1 3.1.4-3 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
CEA Alignment 3.1.4 SURVEILLANCEREQUIREMENTS__(conti~nued) _________
SURVEILLANCE FREQUENCY 1*
SR 3.1.4.3 Verify the CEA deviation circuit is OPERABLE.
SR 3.1.4.4 Verify CEA freedom of movement (trippability) by moving each individual CEA that is not fully inserted into the reactor core 7.5 inches in either direction.
SR 3.1.4.5 Perform a CHANNEL FUNCTIONAL TEST of the V-arl ul +*rk Ahei÷in +":nnCmi+M" hknnnal SR 3.1.4.6 Verify each CEA drop time is
- 3.1 seconds. Prior to reactor criticality, after each removal of the reactor head CALVERT CLIFFS - UNIT 1 3.1.4-4 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
Shutdown CEA Insertion Limits 3.1.5 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME B. One shutdown CEA B.1 Restore shutdown 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> withdrawn CEA(s) to within
_>121.5 inches and limit.
< 129 inches for
> 7 days per occurrence or
> 14 days per 365 days.
< 121.5 inches.
OR Two or more shutdown CEAs not within limit.
C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met.
SURVEILLANCE REQUIREMENTS SURVE I LLANCE FREQUENCY SR 3.1.5.1 Verify each shutdown CEA is withdrawn
_>129 inches.
CALVERT CLIFFS - UNIT 1 3.1.5-2 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
Regulating CEA Insertion Limits 3.1.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY I
SR 3.1.6.1 Verify each regulating CEA group position is within its insertion limits.
SR 3.1.6.2 Verify the accumulated times during which the regulating CEA groups are inserted beyond the steady state insertion limits, but within the transient insertion limits.
SR 3.1.6.3 Verify power dependent insertion limit alarm circuit is OPERABLE.
I CALVERT CLIFFS - UNIT 1 3.1.6-3 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
STE-SDM 3.1.7 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME A. (Continued)
All CEAs inserted and the reactor subcritical by less than the above shutdown reactivity equivalent.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.1 Verify that the position of each CEA not 1-!2 1 fully inserted is within the acceptance criteria for available negative reactivity addition.
SR 3.1.7.2 - -------------------NOTE -------------------
Not required to be performed during initial power escalation following a refueling outage if SR 3.1.4.6 has been met.
Verify that each CEA not fully inserted is Once within capable of full insertion when tripped from 7 days prior to at least the 50% withdrawn position. reducing SDOI to less than the limits of LCO 3.1.1 CALVERT CLIFFS - UNIT I 3.1.7-2 Amendment No. 239 CALVERT CLIFFS - UNIT 2 Amendment No. 213
STE-MODES 1 and 2 3.1.8 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Suspend PHYSICS I hour associated Completion TESTS.
Time not met.
AND B.2 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.8.1 Verify THERMAL POWER is equal to or less than the test power plateau.
CALVERT CLIFFS - UNIT 1 3.1.8-2 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
LHR 3.2.1 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 2. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met.
SURVEILLANCE REQUIREMENTS
NOTE-- ----------------------------
Either the Excore Detector Monitoring System or the Incore Detector Monitoring System shall be used to determine LHR.
SURVEILLANCE FREQUENCY SR 3.2.1.1 Deleted SR 3.2.1.2 -------------------- NOTE----------------
Only applicable when the Excore Detector Monitoring System is being used to determine LHR.
Verify ASI alarm setpoints are within the limits specified in the COLR.
I CALVERT CLIFFS - UNIT 1 3.2.1-2 Amendment No. 297 CALVERT CLIFFS - UNIT 2 Amendment No. 273
LHR 3.2.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.2.1.3 NOTES---------------
- 1. Only applicable when the Incore Detector Monitoring System is being used to determine LHR.
- 2. Not required to be performed below 20% RTP.
Verify incore detector local power density alarms satisfy the requirements of the core power distribution map, which shall be updated at least once per 31 days of accumulated operation in MODE 1.
SR 3.2.1.4 NOTES---------------
- 1. Only applicable when the Incore Detector Monitoring System is being used to determine LHR.
- 2. Not required to be performed below 20% RTP.
Verify incore detector local power density alarm setpoints are less than or equal to the limits specified in the COLR.
CALVERT CLIFFS - UNIT 1 3.2.1-3 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
F3 3.2.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 --- NOTE---------------
F' shall be determined by using the incore detectors to obtain a power distribution map with all full length control element assemblies at or above the long-term steady state insertion limit as specified in the COLR.
Prior to Verify the value of F'r. operation
> 70% RTP after each fuel 1loadi ng *' u-AND "- -
CALVERT CLIFFS - UNIT 1 3.2.3-2 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
Tq 3.2.4 CALVERT CLIFFS - UNIT 1 3.2.4-2 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
ASI 3.2.5 3.2 POWER DISTRIBUTION LIMITS 3.2.5 AXIAL SHAPE INDEX (ASI)
LCO 3.2.5 The ASI shall be maintained within the limits specified in the COLR.
APPLICABILITY: MODE 1 with THERMAL POWER > 20% RTP.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. ASI not within A.1 Restore ASI to within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> limits, limits.
B. Required Action and B.1 Reduce THERMAL POWER 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated Completion to
- 20% RTP.
Time not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.5.1 Verify ASI is within limits specified in the 2hrs COLR.
CALVERT CLIFFS - UNIT 1 3.2.5-1 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
RPS Instrumentation-Operating 3.3.1 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME F. Required Action and F.1 Reduce THERMAL POWER 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion to < 15% RTP.
Time not met for Axial Power Distribution-High and Loss of Load Trip Functions.
G. Required Action and G.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met except for Axial Power Distribution-High and Loss of Load Trip Functions.
SURVEILLANCE REQUIREMENTS
NOTE- ------------------------------
Refer to Table 3.3.1-1 to determine which Surveillance Requirement shall be performed for each RPS Function.
SURVEILLANCE FREQUENCY SR 3.3.1.1 Perform a CHANNEL CHECK of each RPS instrument channel except Loss of Load.
CALVERT CLIFFS - UNIT 1 3.3.1-5 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
RPS Instrumentation-Operating 3.3.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.3.1.2 NOTES---------------
- 1. Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after THERMAL POWER is
Ž 15% RTP.
- 2. The daily calibration.may be suspended during PHYSICS TESTS, provided the calibration is performed upon reaching each major test power plateau, and prior to proceeding to the next major test power plateau.
Perform a calibration (heat balance only) and adjust the excore power range and AT power channels to agree with calorimetric calculation if the absolute difference is
> 1.5%.
SR 3.3.1.3 NOTE---------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after THERMAL POWER is : 20% RTP and required to be performed prior to operation above 90% RTP.
Calibrate the power range excore channels using the incore detectors.
l I
CALVERT CLIFFS - UNIT I 3.3.1-6 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
RPS Instrumentation-Operating 3.3.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.3.1.4 Perform a CHANNEL FUNCTIONAL TEST of each --days RPS instrument channel except Loss of Load and Rate of Change of Power-High.
SR 3.3.1.5 ----------------- NOTE- -----------------
Neutron detectors are excluded from CHANNEL CALIBRATION.
Perform a CHANNEL CALIBRATION on excore power range channels.
SR 3.3.1.6 Perform a CHANNEL FUNCTIONAL TEST of each Once within Rate of Change of Power-High and Loss of 7 days prior to Load instrument channel, each reactor startup SR 3.3.1.7 Perform a CHANNEL FUNCTIONAL TEST on each automatic bypass removal feature.
SR 3.3.1.8 ----------------- NOTE---------------
Neutron detectors are excluded from CHANNEL CALIBRATION.
Perform a CHANNEL CALIBRATION of each 24mo this instrument channel, including applicable automatic bypass removal functions.
CALVERT CLIFFS - UNIT 1 3.3.1-7 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
RPS Instrumentation-Operating 3.3.1 SURVEILLANCE REQUIREMENTS (continued)..- )
SURVEILLANCE FREQUENCY SR 3.3.1.9 -----------------NOTE---------------
Neutron detectors are excluded from RPS RESPONSE TIME testing.
Verify RPS RESPONSE TIME i~s within limits. ; mths on a STAGGERED) TES-
)
B CALVERT CLIFFS - UNIT 1 3.3.1-8 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
RPS Instrumentation-Shutdown 3.3.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.2.1 Perform a CHANNEL CHECK of each Wide Range Logarithmic Neutron Flux Monitor.
SR 3.3.2.2 Perform a CHANNEL FUNCTIONAL TEST on the Once within Rate of Change of Power trip instrument 7 days prior to channel. The allowable value shall be each reactor
< 2.6 dpm. startup 1-SR 3.3.2.3 Perform a CHANNEL FUNCTIONAL TEST on each automatic bypass removal feature.
SR 3.3.2.4 ----------------- -NOTE----------------
Neutron detectors are excluded from CHANNEL CALIBRATION.
Perform a CHANNEL CALIBRATION, including automatic bypass removal features.
CALVERT CLIFFS - UNIT 1 3.3.2-4 Amendment No. 276 CALVERT CLIFFS - UNIT 2 Amendment No. 253
RPS Logic and Trip Initiation 3.3.3 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME D. Two channels of RTCBs D.1 Open the affected Immediately or Trip Path Logic RTCBs.
affecting the same trip leg inoperable.
E. Required Action and E.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A, AND B, or D not met.
E.2 Open all RTCBs. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> OR One or more Functions with two or more Manual Trip, Matrix Logic, Trip Path Logic, or RTCB channels inoperable for reasons other than Condition A or D.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.3.1 Perform a CHANNEL FUNCTIONAL TEST on each 92 4aýy.
RTCB channel.
SR 3.3.3.2 Perform a CHANNEL FUNCTIONAL TEST on each RPS Logic channel.
CALVERT CLIFFS - UNIT 1 3.3.3-2 Amendment No. 275 CALVERT CLIFFS - UNIT 2 Amendment No. 252
ESFAS Instrumentation 3.3.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.4.1 Perform a CHANNEL CHECK of each ESFAS sensor 12-ei~
channel.
SR 3.3.4.2 Perform a CHANNEL
...... FUNCTIONAL TEST of each * \
ESFAS sensor channel.
-N SR 3.3.4.3 Perform a CHANNEL FUNCTIONAL TEST on each automatic block removal feature.
SR 3.3.4.4 Perform a CHANNEL CALIBRATION of each ESFAS cancnr rh~nnal inrl"IHinn A"lnmnfir hln~rL 11 removal feature.
SR 3.3.4.5 Verify ESF RESPONSE TIME is within limits. 24 -Anthg on a
ýS:ýAGGERED TEST, BAG+S +Eý CALVERT CLIFFS - UNIT 1 3.3.4-4 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
ESFAS Logic and Manual Actuation 3.3.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4
SR 3.3.5.1 -- NOTES---------------
- 1. Testing of Actuation Logic shall include verification of the proper relay driver output signal.
- 2. Relays associated with plant equipment that cannot be operated during plant operation are only required to be tested once per 24 months.
Perform a CHANNEL FUNCTIONAL TEST on each ESFAS Actuation Logic channel.
SR 3.3.5.2 Perform a CHANNEL FUNCTIONAL TEST on each ESFAS Manual Actuation channel.
CALVERT CLIFFS - UNIT 1 3.3.5-3 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
DG-LOVS 3.3.6 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Enter applicable Immediately associated Completion Conditions and Time not met. Required Actions for the associated DG made inoperable by DG-LOVS instrumentation.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.6.1 Perform CHANNEL FUNCTIONAL TEST.
= .94month MAU4~
SR 3.3.6.2 Perform CHANNEL CALIBRATION with setpoint 24-.e 4h Allowable Values as follows:
- 1. Transient Degraded Voltage Function
Ž 3630 V and
- 3790 V; Time Delay: Ž 7.6 seconds and
- 8.4 seconds;
- 2. Steady State Degraded Voltage Function
Ž 3820 V and
- 3980 V Time Delay: Ž 97.5 seconds and
- 104.5 seconds; and
- 3. Loss of voltage Function Ž 2345 V and
- 2555 V Time Delay: Ž 1.8 seconds and
- 2.2 seconds at 2450 V.
CALVERT CLIFFS - UNIT 1 3.3.6-3 Amendment No. 232 CALVERT CLIFFS - UNIT 2 Amendment No. 208
CRS 3.3.7 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME B. One required Manual B.1 Place and maintain Immediately Actuation channel or containment purge and Actuation Logic exhaust valves in channel inoperable, closed position.
OR OR More than one B.2 Enter applicable Immediately radiation monitor Conditions and sensor module or Required Actions for associated affected valves of measurement channel LCO 3.9.3, inoperable. "Containment Penetrations," made OR inoperable by isolation Required Action and instrumentation.
associated Completion Time of Condition A not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.7.1 Perform a CHANNEL CHECK on each containment radiation monitor sensor.
aýýE CALVERT CLIFFS - UNIT 1 3.3.7-2 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
CRS 3.3.7 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.3.7.2 -------------------NOTE---------------
Testing of Actuation Logic shall include verification of the proper relay driver output signal.
Perform a CHANNEL FUNCTIONAL TEST on each CRS Actuation Logic channel.
SR 3.3.7.3 Perform a CHANNEL FUNCTIONAL TEST on each containment radiation monitor sensor.
Verify CRS high radiation setpoint is less than or equal to the Allowable Value of 220 mR/hr.
SR 3.3.7.4 Perform a CHANNEL CALIBRATION on each containment radiation monitor instrument channel.
SR 3.3.7;5 SR 3.3.7.6 Perform a CHANNEL FUNCTIONAL TEST on each CRS Manual Actuation channel.
Verify CRS response time is within limits.
IýI 21A fnth1 On a BA-S-H CALVERT CLIFFS - UNIT 1 3.3.7-3 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
CRRS 3.3.8 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C. CRRS trip circuit or C.1 Place one Control Immediately measurement channel Room Emergency.
inoperable during Ventilation System movement of train in irradiated fuel recirculation mode assemblies, with post-loss-of-coolant incident filter fan in service.
OR C.2 Suspend movement of Immediately irradiated fuel assemblies.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.8.1 Perform a CHANNEL CHECK on the control room radiation monitor channel.
SR 3.3.8.2 Perform a CHANNEL FUNCTIONAL TEST on the CRRS radiation monitor trip circuit and measurement channel.
Verify CRRS high radiation setpoint is less than or equal to the Allowable Value of
'6E4 cpm above normal background.
CALVERT CLIFFS.- UNIT 1 3.3.8-2 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
CRRS 3.3.8 SURVEILLANCEREQUIREMENTS_(continued) ________
SURVEILLANCE FREQUENCY I
SR 3.3.8.3 Perform a CHANNEL CALIBRATION on the CRRS radiation monitor trip circuit and measurement channel.
CALVERT CLIFFS - UNIT 1 3.3.8-3 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
CVCS Isolation Signal 3.3.9 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME I
C. Two CVCS isolation C.1 Place one sensor 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> sensor modules or module in bypass and associated place the'other measurement channels sensor module in inoperable, trip.
AND C.2 Restore one sensor 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> module and associated measurement channel to OPERABLE status.
D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND D.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.9.1 Perform a CHANNEL CHECK of each sensor channel.
Lr~~eA x CALVERT CLIFFS - UNIT 1 3.3.9-2 Amendment No. 276 CALVERT CLIFFS - UNIT 2 Amendment No. 253
CVCS Isolation Signal.
3.3.9 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY i
SR 3.3.9.2 ------------------- NOTES---------------
- 1. Testing of Actuation Logic shall include the verification of the proper relay driver output signal.
- 2. Relays associated with plant equipment that cannot be operated during plant operation are only required to be tested once per 24 months.
Perform a CHANNEL FUNCTIONAL TEST on each CVCS sensor channel with setpoints in accordance with the following Allowable Values:
West Penetration Room Pressure-High
- 0.5 psig Letdown Heat Exchanger T :,er ý I Room Pressure-High
- 0.5 psig SR 3.3.9.3 Perform a CHANNEL CALIBRATION on each CVCS 2A--nieths sensor channel.
SR 3.3.9.4 Verify CVCS Isolation Signal response time 24 mnth on a is within limits. ST7AGGERED TI CALVERT CLIFFS.- UNIT 1 3.3.9-3 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
PAM Instrumentation 3.3.10 SURVEILLANCE REQUIREMENTS
NOTE -------------------------------------
These Surveillance Requirements apply to each PAM instrumentation Function in Table 3.3.10-1.
SURVEILLANCE FREQUENCY SR 3.3.10.1 Perform CHANNEL CHECK for each required indication channel that is normally energized.
SR 3.3.10.2 Deleted SR 3.3.10.3 ---------------
NOTE -----------------
Neutron detectors, Core Exit Thermocouples, and Reactor Vessel Level Monitoring System are excluded from CHANNEL CALIBRATION.
Perform CHANNEL CALIBRATION on each indication I
channel.
I _____________________________________________________________
CALVERT CLIFFS - UNIT 1 3.3.10-3 Amendment No. 262 CALVERT CLIFFS - UNIT 2 Amendment No. 239
Remote Shutdown Instrumentation 3.3.11 SURVEILLANCEREQUIREMENTS _________
SURVEILLANCE FREQUENCY I
SR 3.3.11.1 Perform CHANNEL CHECK for each required indication channel that is normally energized.
SR 3.3.11.2 --------------------NOTE---------------
Neutron detectors and Reactor Trip Breaker Indication are excluded from the CHANNEL CALIBRATION.
Perform CHANNEL CALIBRATION for each required indication channel.
CALVERT CLIFFS - UNIT 1 3.3.11-2 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
Wide Range Logarithmic Neutron Flux Monitor Channels 3.3.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.12.1 Perform CHANNEL CHECK.
SR 3.3.12.2 Perform CHANNEL FUNCTIONAL TEST. Once within 7 days prior t each reactor startup SR 3.3.12.3 ------------------- NOTE----------------
Neutron detectors are excluded from CHANNEL CALIBRATION.
Perform CHANNEL CALIBRATION.
CALVERT CLIFFS - UNIT 1 3.3.12-2 Amendment No. 266 CALVERT CLIFFS - UNIT 2 Amendment No. 243
RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 SURVEILLANCE REQUIREMENTS 0 SURVEILLANCE FREQUENCY t
SR 3.4.1.1 Verify pressurizer pressure is within the limits specified in the COLR.
SR 3.4.1.2 Verify RCS cold leg temperature is within the limits specified in the COLR.
SR 3.4.1.3 Verify RCS total flow rate is greater than or equal to the limits specified in the COLR.
SR 3.4.1.4 Verify measured RCS total flow rate is within the limits specified in the COLR.
CALVERT CLIFFS - UNIT 1 3.4.1-2 Amendment No. 301 CALVERT CLIFFS - UNIT 2 Amendment No. 278
RCS P/T Limits 3.4.3 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLET'ION TIME B. (Continued) B.2 Be in MODE 5 with 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> RCS pressure
< 300 psia.
C. --------- NOTE ------- C.1 Initiate action to Immediately Required Action C.2 restore parameter(s) shall be completed to within limits.
whenever this Condition is entered. AND C.2 Determine RCS is Prior to Requirements of acceptable for entering MODE 4 Limiting Condition continued operation.
for Operation not met any time in other than MODE 1, 2, 3, or 4.
SURVEILLANCE REQUIREMENTS SURVE I LLANCE FREQUENCY SR 3.4.3.1 ------------------ NOTE----------------
Only required to be performed during RCS heatup and cooldown operations and RCS CL-rs "*
inservice leak and hydrostatic testing.
Verify RCS pressure, RCS temperature, and RCS heatup and cooldown rates are within limits specified in Figures 3.4.3-1 and 3.4.3-2.
CALVERT CLIFFS - UNIT 1 3.4.3-2 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
RCS Loops - MODES 1 and 2 3.4.4 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 RCS Loops - MODES I and 2 LCO 3.4.4 Two RCS loops shall be OPERABLE and in operation.
APPLICABILITY: MODES 1 and 2.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of A.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Limiting Condition of Operation not met.
SURVEILLANCE REQUIREMENTSSURVEILLANCE___
SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify each RCS loop is in operation.
CALVERT CLIFFS - UNIT 1 3.4.4-1 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
RCS Loops - MODE 3 3.4.5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required RCS loop A.1 Restore required RCS 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperable, loop to OPERABLE status.
B. Required Action and B.1 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not met.
C. No RCS loop OPERABLE. C.1 Suspend operations Immediately that would cause OR introduction of coolant into the RCS No RCS loop in with boron operation. concentration less than required to meet the SDM of LCO 3.1.1.
AND C.2 Initiate action to Immediately restore one RCS loop to OPERABLE status and operation.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 Verify required RCS loop is in operation.
CALVERT CLIFFS - UNIT 1 3.4.5-2 Amendment No. 266 CALVERT CLIFFS - UNIT 2 Amendment No. 243
RCS Loops - MODE 3 3.4.5 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.4.5.2 Verify secondary side water level in each steam generator > -50 inches.
SR 3.4.5.3 Verify correct breaker alignment and indicated power available to the required pump that is not in operation.
CALVERT CLIFFS - UNIT 1 3.4.5-3 Amendment No. 266 CALVERT CLIFFS - UNIT 2 Amendment No. 243
RCS Loops - MODE 4 3.4.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.6.1 Verify one RCS or SDC loop is in operation. .......
SR 3.4.6.2 Verify secondary side water level in required steam generator(s) is > -50 inches.
SR 3.4.6.3 Verify correct breaker alignment and indicated power available to the required loop components that are not in operation.
CALVERT CLIFFS - UNIT 1 3.4.6-3 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
RCS Loops - MODE 5, Loops Filled 3.4.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.7.1 Verify one SDC loop is in operation.
SR 3.4.7.2 Verify required SG secondary side water level is > -50 inches.
SR 3.4.7.3 Verify correct breaker alignment and indicated power available to the required SDC loop components that are not in operation.
CALVERT CLIFFS - UNIT 1 3.4.7-3 Amendment No. 266 CALVERT CLIFFS - UNIT 2 Amendment No. 243
RCS Loops - MODE 5, Loops Not Filled 3.4.8 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME B. Required SDC loops B.1 Suspend operations Immediately inoperable, that would cause introduction of OR coolant into the RCS with boron No SDC loop in concentration less operation. than required to meet the SDM of LCO 3.1.1.
AND B.2 Initiate action to Immediately restore one SDC loop to OPERABLE status and operation.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.8.1 Verify one SDC loop is in operation.
SR 3.4.8.2 Verify correct breaker alignment and indicated power available to the required SDC loop components that are not in operation. ns- t CALVERT CLIFFS - UNIT 1 3.4.8-2 Amendment No. 266 CALVERT CLIFFS - UNIT 2 Amendment No. 243 1
Pressurizer 3.4.9 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition B AND not met.
C.2 Be in Mode 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 Verify pressurizer water level is 133 inches and
- 225 inches.
SR 3.4.9.2 Verify capacity of each required bank of pressurizer heaters Ž 150 kW.
CALVERT CLIFFS - UNIT 1 3.4.9-2 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
Pressurizer PORVs 3.4.11 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME E. Two block valves E.1 Place associated 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> inoperable. PORVs in override closed.
AND E.2 Restore one block 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> valve to OPERABLE status.
F. Required Action and F.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND F.2 Reduce any RCS cold 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> leg temperature 365 0 F (Unit 1),
3017F (Unit 2).
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.1 Perform a CHANNEL FUNCTIONAL TEST of each PORV.
CALVERT CLIFFS - UNIT 1 3.4.11-3 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
Pressurizer PORVs 3.4.11 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY
+
SR 3.4.11.2 NOTE----------------
Not required to be performed with block valve closed in accordance with the Required Actions of this Limiting Condition for Operation.
Perform a complete cycle of each block valve.
SR 3.4.11.3 Perform a complete cycle of each PORV.
Eý SR 3.4.11.4 Perform a CHANNEL CALIBRATION of each PORV. 24-menk CALVERT CLIFFS - UNIT 1 3.4.11-4 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
LTOP System 3.4.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.12.1 Verify a maximum of one HPSI pump is only capable of manually injecting into the RCS.
SR 3.4.12.2 Verify HPSI loop MOVs are only capable of manually aligning HPSI pump flow to the RCS.
SR 3.4.12.3 Verify required RCS vent is open.
SR. 3.4.12.4 Verify PORV block valve is open for each required PORV. -
SR 3.4.12.5 ------------------ NOTE----------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after decreasing any RCS cold leg temperature to : 3651F (Unit 1),
- 3011F (Unit 2).
Perform CHANNEL FUNCTIONAL TEST on each required PORV, excluding actuation.
SR 3.4.12.6 Perform CHANNEL CALIBRATION on each required PORV actuation channel.
CALVERT CLIFFS - UNIT 1 3.4.12-5 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
RCS Operational LEAKAGE 3.4.13 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
.associated Completion Time of Condition A AND not met.
B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR Pressure boundary LEAKAGE exists.
OR Primary to secondary LEAKAGE not within limit.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1 - NOTES----------------
- 1. Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.
- 2. Not applicable to primary to secondary LEAKAGE.
Verify RCS Operational LEAKAGE is within limits by performance of RCS water inventory balance.
CALVERT CLIFFS - UNIT 1 3.4.13-2 Amendment No. 278 CALVERT CLIFFS - UNIT 2 Amendment No. 255
RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.4.13.2 ------------------ NOTE---------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state*
operation.
Verify primary to secondary LEAKAGE is
< 100 gallons per day through any one SG.
CALVERT CLIFFS - UNIT 1 3.4.13-3 Amendment No. 278 CALVERT CLIFFS - UNIT 2 Amendment No. 255
RCS Leakage Detection Instrumentation 3.4.14 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND D.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> E. All required alarms E.1 Enter LCO 3.0.3. Immediately and monitors inoperable.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.1 Perform CHANNEL CHECK of the required 1h containment atmosphere radioactivity 2 moni tor.
SR 3.4.14.2 Perform CHANNEL FUNCTIONAL TEST of the . .
day required containment atmosphere radioactivity monitor.
SR 3.4.14.3 Perform CHANNEL CALIBRATION of the required 24-4tS-containment sump level alarm.
SR 3.4.14.4 Perform CHANNEL CALIBRATION of the required containment atmosphere radioactivity monitor.
CALVERT CLIFFS - UNIT I 3.4.14-3 Amendment No. 299 CALVERT CLIFFS - UNIT 2 Amendment No. 276
RCS Specific Activity 3.4.15 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 3 with 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Tavg < 500 0 F.
Time of Condition A not met.
OR DOSE EQUIVALENT 1-131 in the unacceptable region of Figure 3.4.15-1.
C. Gross activity of the C.1 Be in MODE 3 with 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> reactor coolant not Tavg < 500 0F.
within limit.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.15.1 Verify reactor coolant gross activity
_<10/0E [tCi/gm.
CALVERT CLIFFS - UNIT 1 3.4.15-2 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
RCS Specific Activity 3.4.15 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.4.15.2 -NOTE ----------
Only required to be performed in MODE 1.
Verify reactor coolant DOSE EQUIVALENT 1-131 specific activity
- 0.5 PCi/gm.
AND Between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after THERMAL POWER change of
Ž 15% RTP within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period I-SR 3.4.15.3 - -NOTE -------------
Not required to be performed until 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for Ž 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
Determine E from a sample taken in MODE 1 after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for Ž 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
CALVERT CLIFFS - UNIT 1 3.4.15-3 Amendment No. 281 CALVERT CLIFFS - UNIT 2 Amendment No. 258
STE-RCS Loops - MODE 2 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 Special Test Exception (STE) RCS Loops - MODE 2 LCO 3.4.16 The requirements of LCO 3.4.4, "RCS Loops-MODES 1 and 2," and the listed requirements of LCO 3.3.1, "Reactor Protective System (RPS) Instrumentation-Operating," for the Reactor Coolant Flow-Low, Thermal Margin/Low Pressure, and Asymmetric Steam Generator Transient Functions may be suspended provided:
- a. THERMAL POWER
- 5% RTP; and
- b. The reactor trip setpoints of the OPERABLE Power Level-High channels are set 5 15% RTP.
APPLICABILITY: MODE 2, during startup and PHYSICS TESTS.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. THERMAL POWER not A.1 Open reactor trip Immediately within limit, breakers.
CALVERT CLIFFS - UNIT 1 3.4.16-1 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
STE RCS Loops MODES 4 and 5 3.4.17 ACTIONS..... . .'
CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Suspend activities Immediately requirements of the being performed under Limiting Condition this Special Test for Operation not Exception.
met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.17.1 Verify xenon reactivity is within limits. Once within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> prior to suspending the reactor coolant circulation requirements of LCO 3.4.6, LCO*3.4.7, and LCO 3.4.8 SR 3.4.1 SR 3.4.17.3 Verify charging flow paths isolated.
SR 3.4.17.4 Perform SR 3.1.1.1.
CALVERT CLIFFS - UNIT 1 3.4.17-2 Amendment No. 266 CALVERT CLIFFS - UNIT 2 Amendment No. 243
SITs 3.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY K..
SR 3.5.1.1 Verify each SIT isolation valve is fully open.
SR 3.5.1.2 Verify borated water volume in each SIT is
> 1113 cubic feet.(187 inches) and
- 1179 cubic feet (199 inches).
SR 3.5.1.3 Verify nitrogen cover pressure in each SIT is Ž 200 psig and :9 250 psig.
SR 3.5.1.4 Verify boron concentration in each SIT is
> 2300 ppm and : 2700 ppm.
ffb AND
-'----NOTE----
Only required to be performed for affected SIT Once within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> prior to each solution volume increase of Ž 1% of tank volume CALVERT CLIFFS - UNIT 1 3.5.1-2 Amendment No., 255 CALVERT CLIFFS - UNIT 2 Amendment No. 232 I
SITs 3.5.1 SURVEILLANCE FREQUENCY SR 3.5.1.5 Verify power is removed from each SIT isolation valve operator when pressurizer pressure is
- 2000 psig.
CALVERT CLIFFS - UNIT 1 3.5.1-3 Amendment No. 255 CALVERT CLIFFS - UNIT 2 Amendment No. 232
ECCS - Operating 3.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify the following valves are in the listed position with power to the valve operator removed.
Valve Number Position Function MOV-659 Open Mini-flow Isolation MOV-660 Open Mini-flow Isolation CV-306 Open Low Pressure Safety Injection Flow Control SR 3.5.2.2 Verify each ECCS manual, power-operated, and =324-ý automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.
SR 3.5.2.3 Verify each high pressure safety injection - In accordance and low pressure safety injection pump's with the developed head at the test flow point is Inservice greater than or equal to the required Testing Program developed head.
SR 3.5.2.4 Deleted SR 3.5.2.5 Verify each ECCS automatic valve that is not 6,4___ýýý locked, sealed, or otherwise secured in position, in the flow path actuates to the correct position on an actual or simulated actuation signal.
CALVERT CLIFFS - UNIT 1 3.5.2-2 Amendment No. 260 CALVERT CLIFFS - UNIT 2 Amendment No. 237
ECCS - Operating 3.5.2 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.5.2.6 Verify each ECCS pump starts automatically on an.actual or simulated actuation signal.
SR 3.5.2.7 Verify each low pressure safety injection 24 me~ths pump stops on an actual or simulated actuation signal.
SR 3.5.2.8 Verify, by visual inspection, each ECCS train containment sump suction inlet is not restricted by debris and the suction inlet strainers show no evidence of structural d distress or abnormal corrosion.
SR 3.5.2.9 Verify the Shutdown Cooling System open-permissive interlock prevents the Shutdown Cooling Systemsuction isolation valves from being opened with a simulated or actual Reactor Coolant System pressure signal of
> 309 psia.
CALVERT CLIFFS - UNIT 1 3.5.2-3 Amendment No. 284 CALVERT CLIFFS - UNIT 2 Amendment No. 261
RWT 3.5.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.4.1 ----------------- -NOTE----------------
Only required to be performed when ambient air temperature is < 40 0 F.
Verify RWT borated water temperature is
Ž 40 0 F.
SR 3.5.4.2 NOTES---------------
- 1. Only required to be met in MODE 1.
- 2. Only required to be performed when ambient air temperature is > 1007F.
Verify RWT borated water temperature is O 0 F.
100 SR 3.5.4.3 Verify RWT borated water volume is
Ž 400,000 gallons.
SR 3.5.4.4 Verify RWT boron concentration is Ž 2300 ppm and
- 2700 ppm.
CALVERT CLIFFS - UNIT 1 3.5.4-2 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
STB 3.5.5 3.5 EMERGENCY CORE COOLING SYSTEM (ECCS) 3.5.5 Sodium Tetraborate (STB)
LCO 3.5.5 The STB baskets shall contain Ž 13,750 Ibm of STB.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. STB not within A.1 Restore STB to within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> limits, limits.
B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.5.1 Verify the STB baskets contain Ž 13,750 Ibm 4 of equivalent weight sodium tetraborate decahydrate.
SR 3.5.5.2 Verify that a sample from the STB baskets provides adequate pH adjustment of water borated to be representative of a post-loss-of-coolant accident sump condition.
CALVERT CLIFFS - UNIT 1 3.5.5-1 Amendment No. 290 CALVERT CLIFFS - UNIT 2 Amendment No. 266
Containment Air Locks 3.6.2 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND D.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.1 ----------------- NOTES---------------
- 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.
Perform required air lock leakage rate In accordance testing in accordance with the Containment with the Leakage Rate Testing Program. Containment Leakage Rate Testing Program SR 3.6.2.2 Verify only one door in the air lock can be 24-ifemth&
opened at a time.
CALVERT CLIFFS - UNIT 1 3.6.2-5 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.1 Verify each 4 inch containment vent valve is closed except when the 4 inch containment vent valves are open for pressure control, ALARA or air quality considerations for personnel entry, or for Surveillances that require the valves to be open.
SR 3.6.3.2 ----------------- NOTE ---------------
Valves and blind flanges in high radiation areas may be verified by use of administrative means.
Verify each containment isolation manual valve and blind flange that is located outside containment and not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed, except for containment isolation valves that are open under administrative controls.
CALVERT CLIFFS - UNIT 1 3.6.3-5 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY 4.
SR 3.6.3.3 - ------------------- NOTE---------------
Valves and blind flanges in high radiation areas may be verified by use of administrative means.
Verify each containment isolation manual Prior to valve and blind flange that is located entering MODE 4 inside containment and not locked, sealed, from MODE 5 if or otherwise secured and required to be not performed closed during accident conditions is closed, within the except for containment isolation valves that previous are open under administrative controls. 92 days SR 3.6.3.4 Verify the isolation time of each automatic In accordance power-operated containment isolation valve with the is within limits. Inservice Testing Program SR 3.6.3.5 Verify each automatic containment isolation 4-me n t valve that is not locked, sealed, or otherwise secured in position, actuates to the isolation position on an actual or simulated actuation signal.
CALVERT CLIFFS - UNIT 1 3.6.3-6 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
Containment Pressure 3.6.4 3.6 CONTAINMENT SYSTEMS 3.6.4 Containment Pressure LCO 3.6.4 Containment pressure shall be Ž -1.0 psig and
- 1.0 psig.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Containment pressure A.1 Restore containment 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> not within limits. pressure to within limits.
B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> CALVERT CLIFFS - UNIT 1 3.6.4-1 Amendment No. 303 CALVERT CLIFFS - UNIT 2 Amendment No. 281
Containment Air Temperature 3.6.5 3.6 CONTAINMENT SYSTEMS 3.6.5 Containment Air Temperature LCO 3.6.5 Containment average air temperature shall be
- 120 0 F.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Containment average A.1 Restore containment 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> air temperature not average air within limit, temperature to within limit.
B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> CALVERT CLIFFS - UNIT 1 3.6.5-1 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
Containment Spray and Cooling Systems 3.6.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE i FREQUENCY SR 3.6.6.1 Verify each containment spray manual, power- 4-4a-'g operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position.
SR 3.6.6.2 Operate each containment cooling train fan unit for Ž 15 minutes.
SR 3.6.6.3 Verify each containment cooling train cooling water flow rate is Ž 2000 gpm to each fan cooler.
SR 3.6.6.4 Verify each containment spray pump's In accordance developed head at the flow test point is with the greater than or equal to the required Inservice developed head. Testing Program SR 3.6.6.5 Verify each automatic containment spray mths valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.
SR 3.6.6.6 Verify each containment spray pump starts automatically on an actual or simulated actuation si-gnal.
SR 3.6.6.7 Verify each containment cooling train starts automatically on an actual or simulated actuation signal.
CALVERT CLIFFS - UNIT 1 3.6.6-3 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
IRS 3.6.8 3.6 CONTAINMENT SYSTEMS 3.6.8 Iodine Removal System (IRS)
LCO 3.6.8 Three IRS trains shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One IRS train A.1 Restore IRS train to 7 days inoperable. OPERABLE status.
B. Two IRS trains B.1 Restore one IRS train 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> inoperable, to OPERABLE status.
C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND C.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.8.1 Operate each IRS train for Ž 15 minutes.
JLýA CALVERT CLIFFS - UNIT 1 3.6.8-1 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
IRS 3.6.8 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.6.8.2 Perform required IRS filter testing in In accordance accordance with the Ventilation Filter with the Testing Program. Ventilation Filter Testing Program SR 3.6.8.3 Verify each IRS train actuates on an actual or simulated actuation signal.
---.I CK;ý CALVERT CLIFFS - UNIT 1 3.6.8-2 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
AFW System 3.7.3 SURVEILLANCE REQUIREMENTS ii SURVEILLANCE FREQUENCY SR 3.7.3.1 Verify each AFW manual, power-operated, and automatic valve in each water flowipath and in both steam supply flow paths tolthe steam J turbine-driven pumps, that is not locked, sealed, or otherwise secured in position, is in the correct position.
SR 3.7.3.2 Cycle each testable, remote-operated valve In accordance that is not in its operating position. with the Inservice Testing Program SR 3.7.3.3 ------------------ NOTE---------------
Not required to be performed for the turbine-driven AFW pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 800 psig in the steam generators.
Verify the developed head of each AFW pump In accordance at the flow test point is greater than or with the equal to the required developed head. Inservice Testing Program SR 3.7.3.4 --------------------NOTE---------------
Not required to be performed for the turbine-driven AFW pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 800 psig in the steam generators.
Verify each AFW automatic valve that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation' signal.
CALVERT CLIFFS - UNIT 1 3.7.3-4 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
AFW System 3.7.3 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY 4
SR 3.7.3.5 ----------------- NOTE---------------
Not required to be performed for the turbine-driven AFW pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 800 psig in the steam generators.
Verify each AFW pump starts 'automatically on an actual or simulated actuation signal.
SR 3.7.3.6 -------------------NOTE---------------
Not required to be performed for the AFW train with the turbine-driven AFW pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 800 psig in the steam generators.
Verify the AFW system is capable of f* n m;n~mm im A A nnm n m;n I fl n F1 WV I "IIIVU11MII U IIIIIIssIIUII UI .JU UW QI IVU to each flow leg.
SR 3.7.3.7 Verify the proper alignment of the required Prior to AFW flow paths by verifying flow from the entering MODE 2 condensate storage tank to each steam whenever unit generator. has been in MODE 5 or 6 for
> 30 days CALVERT CLIFFS - UNIT 1 3.7.3-5 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
CST 3.7.4 SURVEILLANCE REQUIREMENTS SURVE I LLANCE FREQUENCY SR 3.7.4.1 Verify CST usable volume is
Ž_150,000 gallons per Unit.
I (Tý--AA CALVERT CLIFFS - UNIT 1 3.7.4-2 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
CC System 3.7.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 ---------------------NOTE---------------
Isolation of CC flow to individual components does not render the CC System inoperable.
Verify each CC manual, power-operated, and automatic valve in the flow path servicing safety-related equipment, that is not locked, sealed, or otherwise secured in position, is in the correct position.
SR 3.7.5.2 Verify each CC automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or I simulated actuation signal.
SR 3.7.5.3 Verify each CC pump starts automatically on an actual or simulated actuation signal.
CALVERT CLIFFS - UNIT 1 3.7.5-2 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
SRW 3.7.6 ACTIQNS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME B. One SRW subsystem B.1 ---------NOTE------ I inoperable. Enter applicable Conditions and Required Actions of LCO 3.8.1, "AC Sources--Operating,"
for diesel generator made inoperable by SRW.
Restore SRW subsystem 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to OPERABLE status.
C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A AND or B not met.
C.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.6.1 -----------------NOTE ---------------
Isolation of SRW flow to individual components does not render SRW inoperable.
Verify each SRW manual, power-operated, and day&-
automatic valve in the flow path servicing safety-related equipment, that is not locked, sealed, or otherwise secured in --LA-*r-position, is in the correct position.
CALVERT CLIFFS - UNIT 1 3.7.6-2 Amendment No. 230 CALVERT CLIFFS - UNIT 2 Amendment No. 206
SRW 3.7.6 SURVEILLANCE REQUIREMENTS (continued) If SURVEILLANCE FREQUENCY SR 3.7.6.2 Verify each SRW automatic valve in the flow 24 MAfl+9 path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.
SR 3.7.6.3 Verify each SRW pump starts automatically on an actual or simulated actuation signal.
CALVERT CLIFFS - UNIT 1 3.7.6-3 Amendment No. 230 CALVERT CLIFFS - UNIT 2 Amendment No. 206
SW 3.7.7 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A AND not met.
B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.7.1 --------------------NOTE------ --------
Isolation of SW System flow to individual components does not render SW inoperable.
Verify each SW System manual, power-operated, and automatic valve in the flow path servicing safety-related equipment, that is not locked, sealed, or otherwise secured in position, is in the correct position. i i
SR 3.7.7.2 Verify each SW System automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuiates to the correct position on an actual or simulated actuation signal.
/
SR 3.7.7.3 Verify each SW System pump starts automatically on an actual or simulated actuation signal.
CALVERT CLIFFS - UNIT 1 3.7.7-2 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
CREVS 3.7.8 SURVEILLANCEREQUIREMENTS _________
SR 3.7.8.1 SURVEILLANCE Operate each required CREVS filter train for I. FREQUENCY
> 15 minutes.
SR 3.7.8.2 Perform required CREVS filter testing in In accordance accordance with the Ventilation Filter with the VFTP Testing Program (VFTP).
SR 3.7.8.3 Verify each CREVS train actuates on an actual or simulated actuation signal.
SR 3.7.8.4 Perform required CRE unfiltered air In accordance inleakage testing in accordance with the with the Control Room Envelope Habitability Program. Control Room Envelope Habitability Program CALVERT CLIFFS - UNIT 1 3.7.8-5 Amendment No. 287 CALVERT CLIFFS - UNIT 2 Amendment No. 264
CRETS 3.7.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.9.1 Verify each required CRETS train has the capability to maintain control room temperature within limits.
CALVERT CLIFFS - UNIT 1 3.7.9-2 Amendment No. 250 CALVERT CLIFFS - UNIT 2 Amendment No. 226
SFPEVS 3.7.11 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.11.1 Verify an OPERABLE SFPEVS train is in operation.
SR 3.7.11.2 Deleted.
SR 3.7.11.3 Verify each SFPEVS fan can maintain a measurable negative pressure with respect to adjacent areas.
CALVERT CLIFFS - UNIT 1 3.7.11-2 Amendment No. 281 CALVERT CLIFFS - UNIT 2 Amendment No. 258
PREVS 3.7.12 3.7 PLANT SYSTEMS 3.7.12 Penetration Room Exhaust Ventilation System (PREVS)
LCO 3.7.12 Two PREVS trains shall be OPERABLE.
APPLICABILITY: MODES 1, 2, and 3.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One PREVS train A.1 Restore PREVS train 7 days inoperable, to OPERABLE status.
B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.12.1 Operate each PREVS train for Ž 15 minutes.
SR 3.7.12.2 Verify required PREVS filter testing in In accordance accordance with the Ventilation Filter with the VFTP Testing Program (VFTP).
CALVERT CLIFFS - UNIT 1 3.7.12-1 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
PREVS 3.7.12 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.7.12.3 Verify each PREVS train actuates on an actual or simulated actuation signal.
CALVERT CLIFFS - UNIT 1 3.7.12-2 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
SFP Water Level 3.7.13 3.7 PLANT SYSTEMS 3.7.13 Spent Fuel Pool (SFP) Water Level LCO 3.7.13 The SFP water level shall be Ž 21.5 ft over the top of irradiated fuel assemblies seated in the storage racks, and 19.8 ft over the top of fuel assemblies seated on rack spacers in the storage racks for reconstitution activities.
APPLICABILITY: During movement of irradiated fuel assemblies in the SFP.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. SFP water level not A.1 -------- NOTE------
within limits. LCO 3.0.3 is not applicable.
Suspend movement of Immediately irradiated fuel assemblies in SFP and suspend reconstitution activities.
SURVEILLANCE REQUIREMENTS ,
SURVEILLANCE FREQUENCY SR 3.7.13.1 Verify the SFP water level is Ž 21.5 ft 4-4&ys-above the top of irradiated fuel assemblies seated in the storage racks. ar-. I CALVERT CLIFFS - UNIT 1 3.7.13-1 Amendment No. 233 CALVERT CLIFFS - UNIT 2 Amendment No. 209
Secondary Specific Activity 3.7.14 3.7 PLANT SYSTEMS 3.7.14 Secondary Specific Activity LCO 3.7.14 The specific activity of the secondary coolant shall be
- 0.10 pCi/gm DOSE EQUIVALENT 1-131.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Specific activity not A.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> within limit.
AND A.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.14.1 Verify the specific activity of the secondary coolant is within limit.
CALVERT CLIFFS - UNIT 1 3.7.14-1 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
SFP Boron Concentration 3.7.16 3.7 PLANT SYSTEMS 3.7.16 Spent Fuel Pool (SFP) Boron Concentration LCO 3.7.16 Boron concentration of the SFP shall be Ž 2000 ppm.
APPLICABILITY: When fuel assemblies are stored in the SFPs.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent Fuel Pool boron ------------ NOTE----------
concentration not LCO 3.0.3 is not applicable.
within limit.
A.1 Suspend movement of Immediately fuel assemblies in the SFPs.
AND A.2 Initiate action to Immediately restore boron concentration to within limit.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.16.1 Verify boron concentration is greater than dayg 2000 ppm.
CALVERt CLIFFS - UNIT 1 3.7.16-1 Amendment No. 267 CALVERT CLIFFS - UNIT 2 Amendment No. 246
AC Sources-Operating 3.8.1 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME K. Three or more K.1 Enter LCO 3.0.3. Immediately required LCO 3.8.1.a and LCO 3.8.1.b AC sources inoperable.
SURVEILLANCE REQUIREMENTS
NOTE-- -------------------------
SR 3.8.1.1 through SR 3.8.1.15 are only applicable to LCO 3.8.1.a and LCO 3.8.1.b AC sources. SR 3.8.1.16 is only applicable to LCO 3.8.1.c AC sources.
SURVEILLANCE FREQUENCY SR 3.8.1.1 ------------------ NOTE ----------------
Only required to be performed when SMECO is being credited for an offsite source.
Verify correct breaker alignment and Once within indicated power availability for the 69 kV 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after SMECO offsite circuit. substitution for a 500 kV offsite circuit AND SR 3.8.1.2 Verify correct breaker alignment and indicated power availability for each required 500 kV offsite circuit.
CALVERT CLIFFS - UNIT 1 3.8.1-10 Amendment No. 265 CALVERT CLIFFS - UNIT 2 Amendment No. 242
AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.8.1.3 NOTES---------------
- 1. Performance of SR 3.8.1.9 satisfies this Surveillance Requirement.
- 2. All DG starts may be preceded by an engine prelube period and followed by a warmup period prior to-loading.
- 3. A modified DG start involving idling and gradual acceleration to synchronous speed may be used for this Surveillance Requirement as recommended;by the manufacturer. When modified start procedures are not used, the voltage and frequency tolerances of SR 3.8.1.9 must be met.
Verify each DG starts and achieves steady:
state voltage Ž 4060 V and
- 4400 V, and frequency Ž 58.8 Hz and
- 61.2 Hz.
CALVERT CLIFFS - UNIT 1 3.8.1-11 Amendment No. 265 CALVERT.CLIFFS - UNIT 2 Amendment No. 242
AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
SURVE I LLANCE FREQUENCY SR 3.8.1.4 -------------------NOTES---------------
- 1. DG loadings may include gradual loading as recommended by the manufacturer.
- 2. Momentary transients below the load limit do not invalidate this test.
- 3. This Surveillance shall be conducted on only one DG at a time.
- 4. This Surveillance Requirement shall be preceded by and immediately follow without shutdown a successful performance of SR 3.8.1.3 or SR 3.8.1.9.
Verify each DG is-syfithronized and loaded, and operates for Ž 60 minutes at a load
Ž 4000 kW for DG 1A and Ž 2700 kW for DGs 1B, 2A, and 2B.
SR 3.8.1.5 Verify each day tank contains Ž 325 gallons of fuel oil for DG 1A and 2 275 gallons of fuel oil for DGs 1B, 2A, and 2B.
I SR 3.8.1.6 Check for and remove accumulated water from cat. uy Ila" SR 3.8.1.7 Verify the fuel oil transfer system operates to automatically transfer fuel oil from storage tank[s] to the day tank.
CALVERT CLIFFS - UNIT 1 3.8.1-12 Amendment No. 265 CALVERT CLIFFS - UNIT 2 Amendment No. 242
AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY I
SR 3.8.1.8 Verify interval between each sequenced load block is within +/- 10% of design interval for the load sequencer.
SR 3.8.1.9 NOTE---------------
All DG starts may be preceded by an engine prelube period.
Verify each DG starts from standby condition and achieves, in
- 10 seconds, voltage
> 4060 V and frequency > 58.8 Hz, and after steady state conditions are reached, maintains voltage Ž 4060 V and
- 4400 V and frequency of > 58.8 Hz and
- 61.2 Hz.
SR 3.8.1.10 Verify manual transfer of AC power sources from the normal offsite circuit to the alternate offsite circuit.
CALVERT CLIFFS - UNIT 1 3.8.1-13 Amendment No. 302 CALVERT CLIFFS - UNIT 2 Amendment No. 279
AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.8.1.11 - ------------------- NOTE---------------
- 1. Momentary transients outside the load and power factor limits do not invalidate this test.
- 2. If performed with the DG synchronized with offsite power, the surveillance test shall be performed at the required power factor. However, if grid conditions do not permit, the power factor limit is not required to be met.
Under this condition, the power factor shall be maintained as close to the limit as practicable.
Verify each DG, operating at a power factor of
- 0.84 for DG 1A and
- 0.83 for DGs 1B, 2A, and 2B, operates for Ž 60 minutes while loaded to Ž 4000 kW for DG 1A and : 3000 kW for DGs 1B, 2A, and 2B.
SR 3.8.1.12 Verify each DG rejects a load Ž 500 hp without tripping.
CALVERT CLIFFS - UNIT 1 3.8.1-14 Amendment No. 302 CALVERT CLIFFS - UNIT 2 Amendment No. 279
AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.8.1.13 Verify that automatically bypassed DG trips are automatically bypassed on an actual or simulated required actuation signal.
SR 3.8.1.14 Verify each DG:
- a. Synchronizes with offsite power source while loaded upon a simulated restoration of offsite power;
- b. Manually transfers loads to offsite power source; and
- c. Returns to ready-to-load operation.
CALVERT CLIFFS - UNIT 1 3.8.1-15 Amendment No. 302 CALVERT CLIFFS - UNIT 2 Amendment No. 279
AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE E FREQUENCY SR 3.8.1.15 ----------------- NOTE--
All DG starts may be preceded by an engine prelube period.
Verify on an actual or simulated loss of offsite power signal in conjunction with an actual or simulated Engineered Safety Feature actuation signal: zh~ I
- a. De-energization of emergency buses;
- b. Load shedding from emergency buses;
- c. DG auto-starts from standby condition and:
- 1. energizes permanently connected loads in
- 10 seconds,
- 2. energizes auto-connected emergency loads through load sequencer,
- 3. maintains steady state voltage
Ž 4060 V and
- 4400 V,
- 4. maintains steady state frequency of Ž 58.8 Hz and
- 61.2 Hz, and
- 5. supplies permanently connected and auto-connected emergency loads for
Ž 5 minutes.
CALVERT CLIFFS - UNIT I 3.8.1-16 Amendment No. 302 CALVERT CLIFFS - UNIT 2 Amendment No. 279
Diesel Fuel Oil 3.8.3
.1:
ACTIONS (continued) .. . . ........ ... . . .
CONDITION REQUIRED ACTION COMPLETION TIME E. One or more DGs with E.1 Restore stored fuel 30 days new fuel oil oil properties to properties not within within limits.
limits.
F. Required Action and F.1 Declare associated Immediately associated Completion DG(s) inoperable.
Time not met.
OR One or more DGs with diesel fuel oil not within limits for reasons other than Condition A, B, C, D, or E. .. I
-SURVEILLANCE.REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.3.1 Verify fuel oil volume of:
- a. FOST 1A Ž 49,500 gallons, and
- b. FOST 21 > 85_000 aallonn.
SR 3.8.3.2 Verify fuel oil properties.of new and stored In accordance fuel oil ate tested in accordance with, and with the Diesel maintained within the limits of, the Diesel Fuel Oil Fuel Oil Testing Program. Testing Program CALVERT CLIFFS - UNIT 1 3.8.3-4 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
Diesel Fuel.Oil
. 3.8.3 SURVEILLANCE REQUIREMENTS (continued)
SURVE I LLANCE T FREQUENCY SR 3.8.3.3 Check for and remove accumulated water from each FOST. 9~Mlse~l CALVERT CLIFFS - UNIT 1 3.8.3-5 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
DC Sources-Operating 3.8.4
-AII SURVEILLANCE REQUIREMENTS-SURVEILLANCE FREQUENCY I
t SR 3.8.4.1 Verify battery terminal voltage is
- 125 V on float charge.
SR 3.8.4.2 Verify no visible corrosion at battery terminals and connectors.
OR Verify battery connection resistance'is within limits.
SR 3.8.4.3 Verify battery cells, cell plates, and racks show no visual indication of physical damage or abnormal deterioration that degrades performance.
SR 3.8.4.4 Remove visible terminal corrosion and verify battery cell to cell and terminal connections are coated with anti-corrosion material.
SR 3.8.4.5 Verify battery connection resistance is within limits.
SR 3.8.4.6 Verify each battery charger supplies
- 400 amps at Ž 125 V for Ž 30 minutes.
-I CALVERT CLIFFS - UNIT 1 3.8.4-2 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
DC Sources-Operating 3.8.4 SIJRVFTII ANCF PFOIIIRFMFNTS (rnnfinii~d~
SURVEILLANCE FREQUENCY SR 3.8.4.3 -------------------- NOTE---------------
The modified performance discharge test in SR 3.8.4.8 may be performed in lieu of the, service test in SR 3.8.4.7.
Verify battery capacity is adequate to supply, and maintain in OPERABLE status, the required emergency loads for the design duty cycle when subjected to a battery service test. I CALVERT CLIFFS - UNIT 1 3.8.4-3 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
DC Sources-Operating 3.8.4
-I ciiovrv i IammW DrAIITDCr~McTC (4-4 .. Ai
. . . .F. I, S.UR.V....A
" , SURVEILLANCE -I FREQUENCY SR 3.8.4.8 Verify- battery capacity is - 80% of the manufacturer.'s rating-:when: subjected to a performance di scharge test ora :modified AND performance discharge'test.
12 months when battery shows degradation or has reached 85%
of the expected life with capacity
< 100% of manufacturer's rating I
AND 24 months-.when battery has reached 85% of the.expected life with capacity
Ž 100% of manufacturer's rating i
CALVERT CLIFFS - UNIT 1 3.8.4-4 Amendment No. 227 CALVERt. CLIFFS - UNIT 2 .Amendment No. 201
.Battery Cell Parameters 3.8.6 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Declare associated Immediately associated Completion battery inoperable.
Time of. Condition A not met, OR One or more batteries with average electrolyte temperature of the representative cells
- < 69 0 F.
OR One "br io-re b-atteri is with one or more battery cell parameters not within Category C limits.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.6.1 Verify battery cell parameters meet Table 3.8.6-1 Category A limits.
SR 3.8.6.2 Verify battery cell parameters meet Table 3.8.6-1 Category B limits.
I
.CALVERT CLIFFS - UNIT 1 3.8.6-3 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
Battery Cell Parameters 3.8.6 SURVEILLANCE REQUIREMENTS (continued) ....... __
SURVEILLANCE FREQUENCY SR 3.8.6i3 Verify average electrolyte temperature of 92 days representativecells is ý 69 0 F.
III II lII I J r4 CALVERT CLIFFS -. UNIT 1 3.8.6-4 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
Inverters-Operating 3.8.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.7.1 Verify correct inverter voltage and alignment to required AC vital. buses.
CALVERT CLIFFS - UNIT 1 3.8.7-2 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
Inverters-Shutdown 3.8.8 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.2 Suspend operations Immediately involving positive reactivity additions that could result in loss of required SDM or boron concentration.
AND A.2.3 Initiate action to Immediately restore required inverters to OPERABLE status.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct inverter voltage and alignment to required AC vital buses.
CALVERT CLIFFS - UNIT 1 3.8.8-2 Amendment No. 279 CALVERT CLIFFS - UNIT 2 Amendment No. 256
Distribution Systems-Operating 3.8.9 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME E. Two or more E.1 Enter LCO 3.0.3. Immediately electrical power distribution subsystems inoperable that result in a loss of function.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.9.1 Verify correct breaker alignments and voltage to AC, DC, and AC vital bus electrical power distribution subsystems.
CALVERT CLIFFS - UNIT 1 3.8.9-2 Amendment No. 304 CALVERT CLIFFS - UNIT 2 Amendment No. 282
Distribution Systems-Shutdown 3.8.10 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.2 Suspend operations Immediately I involving positive reactivity additions that could result in loss of required SDM or boron concentration.
AND A.2.3 Initiate actions to Immediately restore required AC, DC, and AC vital bus electrical power distribution subsystems to OPERABLE status.
AND A.2.4 Declare associated Immediately required shutdown cooling subsystem(s) inoperable and not in operation.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.10.1 Verify correct breaker alignments and voltage to required AC, DC, and AC vital bus -7 day electrical power distribution subsystems.
CALVERT CLIFFS - UNIT 1 3.8.10-2 Amendment No. 279 CALVERT CLIFFS - UNIT 2 Amendment No. 256
Boron Concentration 3.9.1 3.9 REFUELING OPERATIONS 3.9.1 Boron Concentration LCO 3.9.1 Boron concentrations of the Reactor Coolant System and the refueling pool shall be maintained within the limit specified in the COLR.
APPLICABILITY: MODE 6.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Boron concentration not within limit. A.1 Suspend positive Immediately reactivity additions.
AND A.2 Initiate action to Immediately restore boron concentration to within limit.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.1.1 Verify boron concentration is within the limit specified in the COLR.
CALVERT CLIFFS - UNIT 1 3.9.1-1 Amendment No. 279 CALVERT CLIFFS - UNIT 2 Amendment No. 256
Nuclear Instrumentation 3.9.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.2.1 Perform CHANNEL CHECK.
SR 3.9.2.2 ------------------- NOTE----------------
I Neutron detectors are excluded from CHANNEL CALIBRATION.
Perform CHANNEL CALIBRATION.
CALVERT CLIFFS - UNIT 1 3.9.2-2 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201
Containment Penetrations
3.9.3 APPLICABILITY
During movement of irradiated fuel assemblies within containment.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Suspend movement of Immediately containment irradiated fuel penetrations not in assemblies within required status. containment.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.3.1 Verify each required containment penetration is in the required status.
SR 3.9.3.2 Verify each required containment purge and exhaust valve actuates to the isolation position on an actual or simulated actuation signal.
CALVERT CLIFFS - UNIT 1 3.9.3-2 Amendment No. 281 CALVERT CLIFFS - UNIT 2 Amendment No. 258
SDC and Coolant Circulation-High Water Level 3.9.4 ACTIONS (continued) I CONDITION REQUIRED ACTION COMPLETION TIME A. (Continued) A.5 Close one door in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> each air lock.
AND A.6.1 Close each 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> penetration providing direct access from the containment atmosphere to the outside atmosphere with a manual or automatic isolation valve, blind flange, or equivalent.
OR A.6.2 Verify each 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> penetration is capable of being closed by an OPERABLE Containment Purge Valve Isolation System.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.4.1 Verify one SDC loop is in operation and circulating reactor coolant at a flow rate of Ž 1500 gpm.
Tr~s~A~
CALVERT CLIFFS - UNIT I 3.9.4-3 Amendment No. 268 CALVERT CLIFFS - UNIT 2 Amendment No. 244
SDC and Coolant Circulation-Low Water Level 3.9.5 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME I I B. (Continued) B.5.2 Verify each 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> penetration is capable of being closed by an OPERABLE Containment Purge Valve Isolation System.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.5.1 Verify required SDC loops are OPERABLE and one SDC loop is in operation.
SR 3.9.5.2 Verify SDC loop in operation is circulating reactor coolant at a flow rate of
Ž 1500 gpm.
SR 3.9.5.3 Verify correct breaker alignment and indicated power available to the required SDC loop components that are not in operation.
CALVERT CLIFFS - UNIT 1 3.9.5-4 Amendment No. 268 CALVERT CLIFFS - UNIT 2 Amendment No. 244
Refueling Pool Water Level 3.9.6 3.9 REFUELING OPERATIONS 3.9.6 Refueling Pool Water Level LCO 3.9.6 Refueling pool water level shall be maintained Ž 23 ft above the top of the irradiated fuel assemblies seated in the reactor vessel.
APPLICABILITY: During movement of irradiated fuel assemblies within I containment.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Refueling pool water level not within A.1 Suspend movement of Immediately limit, irradiated fuel assemblies within containment.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.6.1 Verify refueling pool water level is Ž 23 ft K 4-I
ýux-above the top of the irradiated fuel assemblies seated in the reactor vessel.
CALVERT CLIFFS - UNIT 1 3.9.6-1 Amendment No. 279 CALVERT CLIFFS - UNIT 2 Amendment No. 256
Programs and Manuals 5.5 5.5 Programs and Manuals Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.
- d. License controlled programs will be used to verify the integrity of the CRE boundary. Conditions that generate relevant information from those programs will be entered into the corrective action process and shall be trended and used as part of the 36 month assessments of the CRE boundary.
- e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.
- f. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and assessing the CRE boundary as required by paragraphs c and d respectively.
CALVERT CLIFFS - UNIT 1 5.5-19 Amendment No. 287 CALVERT CLIFFS - UNIT 2 Amendment No. 264
ATTACHMENT (4)
MARKED-UP TECHNICAL SPECIFICATION BASES PAGES Calvert Cliffs Nuclear Power Plant, LLC May 1, 2014
ATTACHMENT (4)
MARKED-UP TECHNICAL SPECIFICATION BASES PAGES Insert 3 The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
1
SDM B 3.1.1 BASES operator should borate with the best source available for the plant conditions. However, as a minimum, the boration flow rate shall be Ž 40 gpm and the boron concentration shall be Ž 2300 ppm boric acid solution or equivalent.
Assuming that a value of 1% Ak/k must be recovered and a boration flow rate of 40 gpm from the boric acid storage tank, it is possible to increase the boron concentration of the RCS by 100 ppm in approximately 15 minutes. If an inverse boron worth of 100 ppm/% Ak/k is assumed, this combination of parameters will increase the SDM by 1% Ak/k.
These boration parameters of 40 gpm and 100 ppm represent typical values and are provided for the purpose of offering a specific example.
SURVEILLANCE SR 3.1.1.1 REQUIREMENTS SHUTDOWN MARGIN is verified by performing a reactivity balance calculation, considering the listed reactivity effects:
- b. CEA positions;
- c. RCS average temperature;
- d. Fuel burnup based on gross thermal energy generation;
- e. Xenon concentration;
- f. Samarium concentration; and
- g. Isothermal temperature coefficient.
Using the isothermal temperature coefficient accounts for Doppler reactivity in this calculation because the reactor is subcritical and the fuel temperature will be changing at the same rate as the RCS.
_T-ePeunyo 4housi ae onthe--ge-Rra!!y slel
- UNITS 11 & 2 B 3.1.1-6 Revision 27 CALVERT CLIFFS - UNITS CALVERT CLIFFS -
& 2 B 3.1.1-6 Revision 27
CEA Alignment B 3.1.4 BASES Required Action D.2.2 is modified by a Note indicating that performing this Required Action is not required when in conflict with Required Actions A.1, B.1, C.2, or E.1.
E.1 When the CEA deviation circuit is inoperable, performing SR 3.1.4.1 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter ensures improper CEA alignments are identified before unacceptable flux distributions occur. The specified Completion Times take into account other information continuously available to the operator in the Control Room, so that during CEA movement, deviations can be detected, and the protection provided by the CEA inhibit and deviation circuit is not required.
F.1 If any Required Action and associated Completion Time of Condition C, Condition D, or Condition E is not met, one or more regulating or shutdown CEAs are untrippable, two or more CEAs are misaligned by > 15 inches, the unit is required to be brought to MODE 3. By being brought to MODE 3, the unit is brought outside the MODE of applicability. Continued operation is not allowed in the case of more than one CEA misaligned from any other CEA in its group by > 15 inches, or one or more CEAs untrippable.
This is because these cases could result in a loss of SDM and power distribution and a loss of safety function, respectively.
When a Required Action cannot be completed within the required Completion Time, a controlled shutdown should be commenced. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.1.4.1 REQUIREMENTS Verification individual CEA positions are within 7.5 inches (indicated reed switch positions) of all other CEAs in the group Pe - t rAp i'PFowithin 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of any CEA movement of >7.5 i*
CALVERT CLIFFS - UNITS 1 & 2 B 3.1.4-7 Revision 37
CEA Alignment B 3.1.4 BASES
(,. The CEA position verification after each movement of > 7.5 inches ensure that the CEAs in that group are properly aligned at the time when CEA misalignments are most lduring tE movement, dc"iaticrz 2hn bc ......
t , a-nd SR 3.1.4.2 Demonstrating the CEA motion inhibit OPERABLE verifies that the CEA motion inhibit is functional, even if it is not regularly operated. The verification shall ensure that the motion inhibit circuit maintains the CEA group overlap and sequencing requirements of LCO 3.1.6, and prevents any regulating CEA from being misaligned from all other CEAs in pregulrl inerted.
C~The vrifc, otatio durilngurmzte that SR 3.1.4.3 Demonstrating the CEA deviation circuit is OPERABLE verifies SR 3.1.4.4 Verifying each CEA is trippable would require that each CEA be tripped. In MODEs 1 and 2, tripping each CEA would result in radial or axial power tilts or oscillations.
Therefore, individual CEAs are exercised a to provide increased confidence that all CEAs continue to be trippable, even if they are not regularly tripped. A movement of 7.5 inches is adequate to demonstrate motion without exceeding the alignment limit when only one CEA is CALVERT CLIFFS - UNITS 1 & 2 B 3. 1.4-8 Revision 37
CEA Alignment B 3.1.4 BASES being moved. For the purposes of performing the CEA operability test, if the CEA has an inoperable position indicator channel, the alternate indication system (pulse counter or voltage divdin network will be used to monitor Between required performances of SR 3.1.4.5, i a CEA(s)is discovered to be immovable, but remains trippable and aligned, the CEA is considered to be OPERABLE. At any time, if a CEA(s) is immovable, a determination of the trippability (OPERABILITY) of the CEA(s) must be made, and appropriate action taken.
SR 3.1.4.5 Performance of a CHANNEL FUNCTIONAL TEST of each reed switch position transmitter channel ensures the channel is OPERABLE and capable of indicating CEA position over the entire length of the CEA's travel. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specification tests at least once per GOMPononts 'Ii"l; 1 1 y pagg this SR when pc-frfcrmcd at a FI* ueIy *. U e-zr Cvh.oII1 24 . FuI theIIU *e, -the Freq.'.'ezý-rytakes imto accoun1t .. M* . , p ..
shorter Fr-j 1l~es, whi.c,,*. detemin the...
... .. ..... L TY- of th, SR 3.1.4.6 Verification of CEA drop times determined that the maximum CEA drop time permitted is consistent with the assumed drop time used in that safety analysis (Reference 1, Chapter 14).
Control element assembly drop time is measured from the time when electrical power is interrupted to the CEDM until the CALVERT CLIFFS - UNITS 1 & 2 B 3.1.4-9 Revision 37
Shutdown CEA Insertion Limits B 3.1.5 BASES reactor, and the required SDM will be maintained following a reactor trip. This SR and Frequency ensure that the shutdown CEAs are withdrawn before the regulating CEAs are withdrawn during a unit startup.
4ent1 aleRoo tpe prator,Yeiinte ioof 9hutdcm fCE thc ct REFERENCES 1. UFSAR
- 2. 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Plants" CALVERT CLIFFS - UNITS 1 & 2 B 3.1.5-5 Revision 38
Regulating CEA Insertion Limits B 3.1.6 BASES ensures improper CEA alignments are identified before unacceptable flux distributions occur.
E.1 When a Required Action cannot be completed within the required Completion Time, a controlled shutdown should be commenced. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.1.6.1 REQUIREMENTS With the PDIL alarm circuit OPERABLE, verification of each regulating CEA group positionG is sufficient to detect CEA positions that may approach the acceptable limits, and to provide the operator with time to undertake the Required Action(s) should the sequence or insertion limits be found to be exceeded. /'f a.
Verification of the accumulated time of CEA group insertion between the long-term steady state insertion limits and the transient insertion limits ensures the cumulative time limits are not exceeded. T-echrf Fhe REFERENCES 1. UFSAR
- 2. 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Plants" 10 CFR 50.46 CALVERT CLIFFS - UNITS 1 & 2 B 3.1.6-7 Revision 43
STE-SDM B 3.1.7 BASES SURVEILLANCE SR 3.1.7.1 REQUIREMENTS Verification of the position of each partially or fully withdrawn full-length or part-length CEA is necessary to ensure that the minimum negative reactivity requirements for SR 3.1.7.2 Prior demonstration that each CEA to be withdrawn from the core during PHYSICS TESTS is capable of full insertion, when tripped from at least a 50% withdrawn position, ensures that the CEA will insert on a trip signal. The Frequency ensures that the CEAs are OPERABLE prior to reducing SDM to less than the limits of LCO 3.1.1.
The SR is modified by a Note that allows the SR to not be performed during initial power escalation following a refueling outage if SR 3.1.4.6 has been met during that refueling outage. This allows the CEA drop time test, which also proves the CEAs are trippable, to be credited for this SR.
REFERENCES 1. 10 CFR Part 50
- 2. Regulatory Guide 1.68, Revision 2, "Initial Test Programs for Water-Cooled Nuclear Power Plants,"
August 1978
- 3. UFSAR 3.1.7-5 B 3.1.7-5 Revision 11 UNITS 1 CALVERT CLIFFS - UNITS 1&& 22 B Revision 11
STE-MODEs 1 and 2 B 3.1.8 BASES SURVEILLANCE SR 3.1.8.1 REQUIREMENTS Verifying that THERMAL POWER is equal to or less than that allowed by the test power plateau, as specified in the PHYSICS TESTS procedure and required by the safety analysis, ensures that adequate LHR and DNB paramet r mar ins are maintained while LCOs are suspended. h -hu rg~c IhA~
rnci Uf C opn rt ion-9a_-d tr l1- in1 o durin g PHYSTCS REFERENCES 1. 10 CFR Part 50
- 2. Regulatory Guide 1.68, Revision 2, "Initial Test Programs for Water-Cooled Nuclear Power Plants,"
August 1978
- 3. UFSAR CALVERT CLIFFS - UNITS 1 & 2 B 3.1.8-5 Revision 2
LHR B 3.2.1 BASES is not within limits. Therefore, this SR is only applicable when the Excore Detector Monitoring System is being used to The SR is modified by a Note that states that the SR is only applicable when the Excore Detection Monitoring System is being used to determine LHR. The reason for the Note is that the excore detectors input neutron flux information into the ASI calculation.
SR 3.2.1.3 and SR 3.2.1.4 Continuous monitoring of the LHR is provided by the Incore Detector Monitoring System and the Excore Detector Monitoring System. Either of these two core power distribution monitoring systems provides adequate monitoring of the core power distribution and is capable of verifying that the LHR does not exceed its specified limits.
Performance of these SRs verifies that the Incore Detector Monitoring System can accurately monitor LHR. Therefore, they are only applicable when the Incore Detector Monitoring System is being used to determine the LHR.
are modified by two Notes. Note 1 allows the SRs to be performed only when the Incore Detector Monitoring System is being used to determine LHR. Note 2 states that the SRs are not required to be performed when THERMAL POWER is
< 20% RTP. The accuracy of the neutron flux information from the incore detectors is not reliable at THERMAL POWER
< 20% RTP.
REFERENCES 1. Updated Final Safety Analysis Report (UFSAR)
- 2. 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Plants" Revision 43 UNITS 1 CALVERT CLIFFS - UNITS
& 2 1 & 2 B 3.2.1-6 B 3.2.1-6 Revision 43
Fri B 3.2.3 BASES SURVEILLANCE SR 3.2.3.1 REQUIREMENTS The periodic SR to determine the calculated Fr ensures that FT remains within the range assumed in the analysis throughout the fuel cycle. Determining the measured FT once after each fuel loading prior to exceeding 70% RTP ensures that the core is properly loaded.
The power distribution map can only be obtained after THERMAL POWER exceeds 20% RTP because the incore detectors are not reliable below 20% RTP.
The SR is modified by a Note that requires the incore detectors to be used to determine Fr by using them to obtain a power distribution map with all full length CEAs above the long-term steady state insertion limits, as specified in the COLR.
REFERENCES 1. UFSAR
- 2. 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Plants" Revision 43 UNITS 1 CLIFFS - UNITS CALVERT CLIFFS -
1&& 2 2 B 3.2.3-5 B 3.2.3-5 Revision 43
Tq B 3.2.4 BASES is necessary to account explicitly for power asymmetries because the radial power peaking factor used in core power distribution calculations is based on an untilted power distribution.
If Tq is not restored to within its limits, the reactor continues to operate with an axial power distribution mismatch. Continued operation in this configuration may induce an axial xenon oscillation that causes increased LHRs when the xenon redistributes. If Tq cannot be restored to within its limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, reactor power must be reduced.
C.1 If Required Actions and associated Completion Times of Condition A or B are not met, THERMAL POWER must be reduced to
- 50% RTP. This requirement provides conservative protection from increased peaking due to potential xenon redistribution and provides reasonable assurance that the core is operating within its thermal limits and places the core in a conservative condition. Four hours is a reasonable time to reach 50% RTP in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.2.4.1 REQUIREMENTS at ý2-+ e'-'
kk imervs-' Tq is determined using the incore and excore detectors. When one excore channel is inoperable and THERMAL POWER is > 75%
REFERENCES 1. UFSAR
- 2. 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Plants" CALVERT CLIFFS - UNITS 1 & 2 B 3.2.4-5 Revision 43
ASI B 3.2.5 BASES ACTIONS A.1 Operating the core within ASI limits specified in the COLR and within the limits of LCO 3.3.1 ensures an acceptable margin for DNB and for maintaining local power density in the event of an AOO. Maintaining ASI within limits also ensures that the limits of Reference 2 are not exceeded during accidents. The Required Actions to restore ASI must be completed within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to limit the duration the plant is operated outside the initial conditions assumed in the accident analyses. In addition, this Completion Time is sufficiently short that the xenon distribution in the core cannot change significantly.
B.1 If the ASI cannot be restored to within its specified limits, or ASI cannot be determined because of Excore Detector Monitoring System inoperability, core power must be reduced. Reducing THERMAL POWER to
- 20% RTP provides reasonable assurance that the core is operating farther from thermal limits and places the core in a conservative condition. Four hours is a reasonable amount of time, based on operating experience, to reduce THERMAL POWER to
< 20% RTP in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.2.5.1 REQUIREMENTS Verifying that the ASI is within the specified limits provides reasonable assurance that the core is n t apprechaninq ta affcct thc rdztiut efI32chcxcc s4 eowly and shewul d be diseee befei e the 1imits ai CALVERT CLIFFS - UNITS 1 & 2 B 3.2.5-5 Revision 11
RPS Instrumentation-Operating B 3.3.1 BASES SURVEILLANCE The SRs for any particular RPS Function are found in the SR REQUIREMENTS column of Table 3.3.1-1 for that Function. Most Functions are subject to CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION.
SR 3.3.1.1 Performance of the CHANNEL CHECK h ensures that gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one instrument channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument channel drift in one of the channels or of something even more serious. CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined by the plant staff based on a qualitative assessment of the instrument channel combined with the instrument channel uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the transmitter or the signal processing equipment has drifted outside its limits. CHANNEL CHECKS are performed on the wide range logarithmic neutron flux monitor for the Rate of Change of Power-High trip Function.
losslof R in retimd nt nneis in L fnrcdourantechannel TameC s foal, but ofo
.... of th* dipayl...
2 B 3.3.1-27 Revision 34 CALVERT CALVERT CLIFFS -
UNITS 1 CLIFFS - UNITS 1&
& 2 B 3.3.1-27 Revision 34
RPS Instrumentation-Operating B 3.3.1 BASES SR 3.3.1.2 A daily calibration (heat balance) is performed when THERMAL POWER is Ž 15%. The daily calibration shall consist of adjusting the "nuclear power calibrate" potentiometers to agree with the calorimetric calculation if the absolute difference is > 1.5%. The "AT power calibrate" potentiometers are then used to null the "nuclear power-AT power" indicators on the RPS Calibration and Indication Panel. Performance of the daily calibration ensures that the two inputs to the Q power measurement are indicating accurately with respect to the much more accurate secondary calorimetric calculation. The heat balance addresses overall gain of the instruments and does not include ASI.
tha once the unit reaches 15% RTP, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is the maximum time allowed for completing this Surveillance. The secondary calorimetric is inaccurate at lower power levels.
The 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allows time for plant stabilization, data-taking, and instrument calibration.
A second Note indicates the daily calibration may be suspended during PHYSICS TESTS. This ensures that calibration is proper both preceding and following physics testing at each plateau, recognizing that during testing, changes in power distribution and RCS temperature may render the calibration inaccurate.
SR 3.3.1.3 It is necessary to calibrate the excore power range channel upper and lower subchannel amplifiers such that the internal ASI used in the TM/LP trip and APD-High trip Functions reflects the true core power distribution as determined by the incore detectors. A Note indicates that once the unit reaches 20% RTP, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is the maximum time allowed for completion of this Surveillance. The Surveillance is required to be performed prior to operation above 90% RTP.
Uncertainties in the excore and incore measurement process make it impractical to calibrate when THERMAL POWER is CALVERT CLIFFS - UNITS 1 & 2 B 3.3. 1-28 Revision 34
RPS Instrumentation-Operating B 3.3.1 BASES
< 20% RTP. The Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allows time for plant stabilization data-taking, and instrument
ý-5 equiring te Sprior to operations above 90% RTP is because of the increased uncertainties associated with using uncalibrated excore detectors. If the excore channels are not properly calibrated to agree with the incore detectors, power is restricted during subsequent operations because of increased uncertainty associated with using uncalibrated excore SR 3.3.1.4 A CHANNEL FUNCTIONAL TEST is performed on each RPS instrument channel, except Loss of Load and Rate of Change of Power, e to ensure the entire channel will perform its intended function when needed. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specification tests at least once per refueling interval with applicable extensions.
In addition to reference voltage power supply tests, the RPS CHANNEL FUNCTIONAL TEST consists of three overlapping tests as described in Reference 1, Section 7.2. These tests verify that the RPS is capable of performing its intended function, from bistable input through the RTCBs. They include:
Bistable Tests The bistable setpoint must be found to trip within the Allowable Values specified in the LCO and left set CALVERT CLIFFS - UNITS 1 & 2 B 3.3. 1-29 Revision 36
RPS Instrumentation-Operating B 3.3.1 BASES consistent with the assumptions of Reference 4. As-found values must also be recorded and reviewed for consistency with the assumptions of the frequency extension analysis.
The requirements for this review are outlined in Reference 8.
A test signal is substituted as the input in one instrument channel at a time to verify that the bistable trip unit trips within the specified tolerance around the setpoint.
This is done with the affected RPS channel bistable trip unit bypassed. Any setpoint adjustment shall be consistent with the assumptions of Reference 4.
Matrix Loqic Tests Matrix logic tests are addressed in LCO 3.3.3. This test is performed one matrix at a time. It verifies that a coincidence in the two instrument channels for each Function removes power from the matrix relays. During testing, power is applied to the matrix relay test coils and prevents the matrix relay contacts from assuming their de-energized state. This test will detect any short circuits around the bistable contacts in the coincidence logic, such as may be caused by faulty bistable relay or trip bypass contacts.
Trip Path Tests Trip path logic tests are addressed in LCO 3.3.3. These tests are similar to the matrix logic tests, except that test power is withheld from one matrix relay at a time, allowing the trip path circuit to de-energize, opening the affected set of RTCBs. The RTCBs must then be closed prior to testing the other three trip path circuits, or a reactor trip may result.
SR 3.3.1.5 A CHANNEL CALIBRATION of the excore power range channels K ensures that the channels are reading accurately and within tolerance. The SR verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves CALVERT CLIFFS - UNITS 1 & 2 B 3.3. 1-30 Revision 36
RPS Instrumentation-Operating B 3.3.1 BASES the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive tests. CHANNEL CALIBRATIONS must be performed consistent with the plant-specific SRs.
The as-found and as-left values must also be recorded and reviewed for consistency with the assumptions of the Frequency extension analysis. The requirements for this review are outlined in Reference 8.
A Note is added stating that the neutron detectors are excluded from CHANNEL CALIBRATION because they are passive devices with minimal drift and because of the difficulty of simulating a meaningful signal (Reference 7). Slow changes in detector sensitivity are compensated for by performing the daily calorimetric calibration (SR 3.3.1.2) and the monthly linear subchannel gain check (SR 3.3.1.3). In addition, associated control room indications are continuously monitored by the operators.
SR 3.3.1.6 A CHANNEL FUNCTIONAL TEST on the Loss of Load, and Rate of Change of Power channels is performed prior to a reactor startup to ensure the entire channel will perform its intended function if required. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specification tests at least once per refueling interval with applicable extensions. The Loss of Load sensor cannot be tested during reactor operation without causing reactor trip. The Power Rate of Change-High trip Function is required during startup operation and is bypassed when shut down or > 12% RTP.
CALVERT CLIFFS - UNITS 1 & 2 B 3.3. 1-31 Revision 36
RPS Instrumentation-Operating B 3.3.1 BASES SR 3.3.1.7 Surveillance Requirement 3.3.1.7 is a CHANNEL FUNCTIONAL TEST similar to SR 3.3.1.4, except SR 3.3.1.7 is applicable only to Functions with automatic bypass removal features. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specification tests at least once per refueling interval with applicable extensions. Proper operation of operating bypasses are critical during plant startup because the bypasses must be in place to allow startup operation and must be removed at the appropriate *nts dun po er ascent to enable certain
-'ýypasses are-removed, the bypasses must not fail in such a way that the associated trip Function gets inadvertently bypassed. This feature is verified by the trip Function CHANNEL FUNCTIONAL TEST, SR 3.3.1.4. Therefore, further testing of the automatic bypass removal feature after startup is unnecessary.
SR 3.3.1.8 Surveillance Requireis the performance of a CHANNEL CALIBRATION CHANNEL CALIBRATION is a check of the instrument channel, including the sensor. The SR verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument channel drift between successive calibrations to ensure that the channel remains operational between successive tests. CHANNEL CALIBRATIONS must be performed consistent with Reference 4.
The as-found and as-left values must also be recorded and reviewed for consistency with the assumptions of the frequency extension analysis. The requirements for this review are outlined in Reference 6.
CALVERT CLIFFS - UNITS I & 2 B 3.3.1-32 Revision 36
RPS Instrumentation-Operating B 3.3.1 BASES The SR is modified by a Note to indicate that the neutron detectors are excluded from CHANNEL CALIBRATION because they are passive devices with minimal drift, and because of the difficulty of simulating a meaningful signal. Slow changes in detector sensitivity are compensated for by performing the calorimetric calibration (SR 3.3.1.2) and the Y i near subchannel gain check (SR 3.3.1.3).
SR 3.3.1.9 This SR ensures that the RPS RESPONSE TIMES are verified to be less than or equal to the maximum values assumed in the safety analysis. Individual component response times are not modeled in the analyses. The analyses model the overall or total elapsed time from the point at which the parameter exceeds the trip setpoint value at he sens to the oint acceptance criteria are inc ded in Reference i, Section 7.2.Tes me pertre inoer ma s at hocn thetrndr fail~s opn ;trmnarencopra I-ailurc, ar inrgun ~ ~
uiur~c~ R:ý Alzo, esp t iimnz aluothe iso rbasedsuporns uprtimg eperuiencenwhich htm 24
- h ...he - . &n-failures 01 ............ t -at e eoepn) nt is.verified. oans ef eAl enraraesponseut notiesaye obtain....e dfrom red e r s; ultsA- d'A-.a Avend t-r or reu etn a eprom noemaueetor in over apping segments, with verification that all components are tested.
Response time may be verified by any series of sequential, overlapping or total channel measurements, including allocated sensor response time, such that the response time is verified. Allocations for sensor response times may be obtained from records of test results, vendor test data, or vendor engineering specifications. Reference 9 provides the basis and methodology for using allocated sensor response times in the overall verification of the channel response CALVERT CLIFFS - UNITS 1 & 2 B 3.3. 1-33 Revision 36
RPS Instrumentation-Shutdown B 3.3.2 BASES D.1. D.2.1. and D.2.2 Condition D applies to two inoperable automatic bypass removal features. If the automatic bypass removal features cannot be restored to OPERABLE status, the associated Rate of Change of Power-High trip RPS channel may be considered OPERABLE only if the bypasses are not in effect. Otherwise, the affected RPS channels must be declared inoperable, as in Condition B, and the bypasses either removed or the automatic bypass removal features repaired. Also, Required Action D.2.2 provides for the restoration of the one affected automatic trip channel to OPERABLE status within the rules of Completion Time specified under Condition B.
Completion Times are consistent with Condition B.
E.1 Condition E is entered when the Required Actions and associated Completion Times of Condition A, B, C, or D are not met.
If Required Actions associated with these Conditions cannot be completed within the required Completion Time, opening the RTCBs brings the reactor to a MODE where the LCO does not apply and ensures no CEA withdrawal will occur. The basis for the Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is that it is adequate to complete the Required Actions without challenging plant systems.
SURVEILLANCE SR 3.3.2.1 REQUIREMENTS Performance of the CHANNEL CHECK on each wide range channel l ensures that gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one instrument channel to a similar parameter on another channel. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument channel drift in one of the channels or of something even more serious. CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying Revision 31 UNITS 1 CALVERT CLIFFS - UNITS 1& & 2 2 B 3.3.2-7 B 3.3.2-7 Revision 31
RPS Instrumentation-Shutdown B 3.3.2 BASES that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined by the plant staff based on a qualitative assessment of the instrument channel that considers instrument channel uncertainties, including isolation, indication, and readability. If a channel is outside the criteria, it may be an indication that the transmitter or the signal processing equipment has drifted outside its limits.
SR 3.3.2.2 A CHANNEL FUNCTIONAL TEST on the power rate of change channels is performed once within 7 days prior to each reactor startup to ensure the entire instrument channel will perform its intended function if required. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specification tests at least once per refueling interval with applicable extensions. The Rate of Change of Power-High trip Function is required during startup operation and is bypassed when shut down or > 12% RTP. Additionally, operating experience has shown that these components usually pass the SR when performed at a Frequency of once within 7 days prior to each reactor startup.
Only the Allowable Values are specified for each RPS trip Function in the SR. Nominal trip setpoints are established for the Functions via the plant-specific procedures. The CALVERT CLIFFS - UNITS 1 & 2 B 3.3.2-8 Revision 36
RPS Instrumentation-Shutdown B 3.3.2 BASES nominal setpoints are selected to ensure the plant parameters do not exceed the Allowable Value if the bistable trip unit is performing as required. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable, provided that operation and testing are consistent with the assumptions of the plant-specific setpoint calculations. Each nominal trip setpoint is more conservative than the analytical limit assumed in the safety analysis in order to account for instrument channel uncertainties appropriate to the trip Function. These uncertainties are defined in Reference 3.
SR 3.3.2.3 Surveillance Requirement 3.3.2.3 is a CHANNEL FUNCTIONAL TEST similar to SR 3.3.2.2, except SR 3.3.2.3 is applicable on9JYtnbypass Functions i A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specification tests at least once per refueling interval with applicable extensions.
Proper operation of operating bypasses is critical during plant startup because the bypasses must be in place to allow startup operation and must be removed at the appropriate points during power ascent to enable certain reactor trips.
bypasses must not fail in such a way that the associated trip Function gets inadvertently bypassed. This feature is verified by SR 3.3.2.2. Therefore, further testing of the automatic bypass removal feature after startup is unnecessary.
SR 3.3.2.4 Surveillance Requirement 3.3.2.4 is the performance of a CHANNEL CALIBRATION_
CALVERT CLIFFS - UNITS 1 & 2 B 3.3.2-9 Revision 36
RPS Instrumentation-Shutdown B 3.3.2 BASES CHANNEL CALIBRATION is a check of the instrument channel including the sensor. The SR verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive tests. CHANNEL CALIBRATIONS must be performed consistent with Reference 3.
The as-found and as-left values must also be recorded and reviewed for consistency with the assumptions of the SR interval extension analysis. The requirements for this review are outlined in Reference 4.
The SR is modified by a Note to indicate that the neutron detectors are excluded from CHANNEL CALIBRATION because they are passive devices with minimal drift (Reference 5).
REFERENCES 1. 10 CFR Parts 50, "Domestic Licensing of Production and Utilization Facilities," and 100, "Reactor Site Criteria"
Chapter 14, "Safety Analysis"
- 3. CCNPP Setpoint File
- 4. Combustion Engineering Topical Report CEN-327, "RPS/ESFAS Extended Test Interval Evaluation" dated June 2, 1986, including Supplement 1, March 3, 1989
- 5. Letter from Mr. R. E. Denton (BGE) to NRC Document Control Desk, dated June 6, 1995, "License Amendment Request; Extension of Instrument Surveillance Intervals" CALVERT CLIFFS - UNITS 1 & 2 B 3.3.2-10 Revision 36
RPS Logic and Trip Initiation B 3.3.3 BASES one manual trip, matrix logic, trip path logic, or RTCB channel is inoperable for reasons other than Condition A or D.
If the RTCBs associated with the inoperable channel cannot be opened, the reactor must be shut down within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and all the RTCBs opened. A Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, to reach the required MODE from full power conditions in an orderly manner, without challenging plant systems, and to open RTCBs. All RTCBs should then be opened, placing the plant in a MODE where the LCO does not apply and ensuring no CEA withdrawal occurs.
SURVEILLANCE SR 3.3.3.1 REQUIREMENTS A CHANNEL FUNCTIONAL TEST is performed on each RTCB channel
.This verifies proper operation of each RTCB.
A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specification tests at least once per refueling interval with applicable extensions. The RTCB must then be closed prior to testing SR 3.3.3.2 A CHANNEL FUNCTIONAL TEST on each RPS logic channel is performed teto ensure the entire channel will perform its in end function when needed. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specification tests at B 3.3.3-11 Revision 36 UNITS 1 CLIFFS - UNITS CALVERT CLIFFS & 2 1& 2 B 3.3.3-11 Revision 36
RPS Logic and Trip Initiation B 3.3.3 BASES least once per refueling interval with applicable extensions.
In addition to reference voltage tests, the RPS CHANNEL FUNCTIONAL TEST consists of three overlapping tests as described in Reference 1, Section 7.2. These tests verify that the RPS is capable of performing its intended function, from bistable input through the RTCBs. The first test, the instrument channel test, is addressed by SR 3.3.1.4 in LCO 3.3.1.
This SR addresses the two tests associated with the RPS logic: matrix logic and trip path logic.
Matrix Logic Tests These tests are performed one matrix at a time. They verify that a coincidence in the two instrument channels for each Function removes power from the matrix relays. During testing, power is applied to the matrix relay test coils and prevents the matrix relay contacts from assuming their de-energized state. The matrix logic tests will detect any short circuits around the bistable contacts in the coincidence logic such as may be caused by faulty bistable relay or trip bypass contacts.
Trip Path Tests These tests are similar to the matrix logic tests, except that test power is withheld from one matrix relay at a time, allowing the trip path circuit to de-energize, opening the affected set of RTCBs. The RTCBs must then be closed prior to testing the other three trip path circuits, or a reactor trip may result.
affece setunc of 92~s The3 isbasc mun then rbe closed pro B 3.3.3-12 Revision 36 CALVERT CLIFFS - UNITS 1 CLIFFS - UNITS 1&& 2 2 B 3.3.3-12 Revision 36
ESFAS Instrumentation B 3.3.4 BASES SURVEILLANCE The SRs for any particular ESFAS Function are found in the REQUIREMENTS SRs column of Table 3.3.4-1 for that Function. Most Functions are subject to CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, CHANNEL CALIBRATION, and response time testing.
SR 3.3.4.1 Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one sensor channel to a similar parameter on other sensor channels. It is based on the assumption that sensor channels monitoring the same parameter should read approximately the same value. Significant deviations between sensor channels could be an indication of excessive sensor channel drift in one of the channels or of something even more serious. CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined by the plant staff based on a qualitative assessment of the sensor channel, which considers sensor channel uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. If the channels are within the criteria, it is an indication that the channels are OPERABLE. If the channels are normally off-scale during times when surveillance testing is required, the CHANNEL CHECK will only verify that they are off-scale in the same direction. Off-scale low current loop channels are verified to be reading at the bottom of the range and not failed down-scale.
ft channclur*s.-..n ufureqentur f u n T-he CHANNEL CII[2( sup.eflemets less fermal, but mr-e frrequent, cheeks f h canc drngnoml pý ýioa u-se of displays CALVERT CLIFFS - UNITS 1 & 2 B 3.3.4-19 Revision 26
ESFAS Instrumentation B 3.3.4 BASES SR 3.3.4.2 A CHANNEL FUNCTIONAL TEST is performed to ensure the entire sensor channel will perform its intended function when needed. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specification tests at least once per refueling interval with applicable extensions.
The CHANNEL FUNCTIONAL TEST tests the individual sensor channels using an analog or level switch test input to each bistable.
A test signal is substituted for the input in one sensor channel at a time to verify that the bistable trips within the specified tolerance around the setpoint. Any setpoint adjustment shall be consistent with the assumptions of the Reference 5.
SR 3.3.4.3 Surveillance Requirement 3.3.4.3 is a CHANNEL FUCNCIONAL TEST similar to SR 3.3.4.2, except 3.3.4.3 is 2 only applicable to automatic block removW features ofthe sensor block modules. These include the Pressurizer Pressure-Low trip block and the SGIS Steam Generator Pressure-Low trip block. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specification tests at least once per refueling interval with applicable extensions.
The CHANNEL FUNCTIONAL TEST for proper operation of the automatic block removal features is critical during plant heatups because the blocks may be in place prior to entering MODE 3, but must be removed at the appropriate oints during plant startup to enable the ESFAS Function.
CALVERT CLIFFS - UNITS 1 & 2 B 3.3.4-20 Revision 36
ESFAS Instrumentation B 3.3.4 BASES (Frequeney- iz,adequwate toenur prpepr automfat ie bleel) mduc ocrtin~a cci
~erovl in ifrnc3 Once the blocks are removed, the blocks must not fail in such a way that the associated ESFAS Function is inappropriately blocked. This feature is verified by the appropriate ESFAS Function CHANNEL FUNCTIONAL TEST.
SR 3.3.4.4 CHANNEL CALIBRATION is a check of the sensor channel, including the automatic block removal feature of the sensor block module and the sensor. The SR verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for sensor channel drift between successive calibrations to ensure that the channel remains operational between successive surveillance tests.
CHANNEL CALIBRATIONS must be performed consistent with Reference 5.
The as-found and as-left values must also be recorded and reviewed for consistency with the assumptions of the extension analysis. The requirements for this review are outlined in Reference 6.
of cal ibra tiodin tera for cr Im w ic ,f.
nh the filiaugnitu SR 3.3.4.5 This SR ensures that the train actuation response times are the maximum values assumed in the safety analyses.
Individual component response times are not modeled in the analyses. The analysis models the overall or total elapsed time, from the point at which the parameter exceeds the trip setpoint value at the sensor to the point at which the equipment in both trains reaches the required functional state (e.g., pumps are rated discharge pressure, valves in full open or closed position). Response time testing CALVERT CLIFFS - UNITS 1 & 2 B 3.3.4-21 Revision 36
ESFAS Instrumentation B 3.3.4 BASES acceptance criteria are included in Reference 1, Section 7.3. The test may be performed in one measurement or in overlapping segments, which verification that all components are measured.
Response time may be verified by any series of sequential, overlapping or total channel measurements, including allocated sensor response time, such that the response time is verified. Allocations for sensor response times may be obtained from records of test results, vendor test data, or vendor engineering specifications. Reference 7 provides the basis and methodology for using allocated sensor response times in the overall verification of the channel response time for specific sensors identified in the reference.
Response time verification for other sensor types must be demonstrated by test. The allocation of sensor response times must be verified prior to placing a new component in operation and reverified after maintenance that may adversely affect the sensor response time.
Instrument loop or test cables and wiring add an insignificant response time and can be ignored.
Eptpaegred tzatity geaur Resons Timetess reification 3f LhcPadSTAGERE ET-.S1S iuili.Te2-Meths 4
every -e 4TC[-DTS resbiltse~eI~ in
- Wfe-ef1 4nots "ir b etween supur m is 3 the rtests siv nmbenn*vmrn ofer~n copntentow. c~auzing ho . S-iu- nr.. h i tm urnatT crdin ofu u REFERENCES 1. UFSAR
- 2. IEEE No. 279, "Proposed IEEE Criteria for Nuclear Power Plant Protection Systems," August 1968 CALVERTNCLIFS UNI
- 1. S 1&2 B3. -2Reiin3 CALVERT CLIFFS - UNITS 1 & 2 B 3.3.4-22 Revision 36
ESFAS Logic and Manual Actuation B 3.3.5 BASES which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. 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 3.3.5.1 REQUIREMENTS A CHANNEL FUNCTIONAL TEST is performed to ensure the entire actuation logic channel will perform its intended function when needed. Sensor channel tests are addressed in LCO 3.3.4. This SR addresses actuation logic tests. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specification tests at least once per refueling interval with applicable extensions.
Actuation Logic Tests Actuation logic channel testing includes injecting one actuation signal into each two-out-of-four logic actuation modules in each ESFAS Function, and using a bistable trip input to satisfy the actuation logic. Testing includes block logic modules.
Note 1 requires that actuation logic tests include operation of actuation relays. Note 2 allows deferred at power testing of certain subchannel relays to allow for the fact that operating certain relays during power operation could cause plant transients or equipment damage. Those subchannel relays that cannot be tested at power must be tested in accordance with Note 2. These include SIAS No. 5, SIAS No. 10, CIS No. 5, SGIS No. 1, and CSAS No. 3.
These subchannel relays actuate the following components, which cannot be tested at power:
RCP seal bleedoff isolation valves; Service water isolation valves; CALVERT CLIFFS - UNITS 1 & 2 B 3.3.5-13 Revision 36
ESFAS Logic and Manual Actuation B 3.3.5 BASES Volume control tank discharge valves; Letdown stop valves; Component Cooling to and from the RCPs; MSIVs and feedwater isolation valves; Instrument air CIVs; Heater drain pumps; Main feedwater pumps; and Condensate booster pumps.
The reasons each of the above cannot be fully tested at power are stated in Reference 1.
Actuation logic tests verify that the ESFAS is capable of performing its intended function, from bistable input through the actuated components.
SR 3.3.5.2 A CHANNEL FUNCTIONAL TEST is performed on the manual ESFAS actuation circuitry, de-energizing relays and providing manual actuation of the Function. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specification tests at least once per refueling interval with applicableextensions.
This surveillance test verifies that the actuation push buttons are capable of opening contacts in the actuation logic as designed, de-energizing the actuati elays and providin manual tri of the Function. Th24 -mi94i
... u the .
ditio's that apply-during a plnt outagc, and the potertial! nupandtrnin ft~ts were-to be pe Formed wi th the reactor at -pewer. Opeprating Revision 36 UNITS 1 CLIFFS - UNITS CALVERT CLIFFS - 1& & 2 2 B 3.3.5-14 B 3.3.5-14 Revision 36
ESFAS Logic and Manual Actuation B 3.3.5 BASES expere...
he cm . cct; mh~
eh~ usu al 4y pas; the stryeillanee test when pcrformcled at ea Fire ueIILJ uoF iL ery24 monthg.
REFERENCES 1. UFSAR, Section 7.3, "Engineered Safety Features Actuation Systems"
- 2. Letter from Mr. R. E. Denton (BGE) to NRC Document Control Desk, dated June 5, 1995, "Response to NRC Request for Review & Comment on Review of Preliminary Accident Precursor Analysis of Trip; Loss of 13.8 kV Bus; Short-Term Saltwater Cooling System Unavailability, CCNPP Unit 2" CALVERT CLIFFS - UNITS 1 & 2 B 3.3.5-15 Revision 36
DG-LOVS B 3.3.6 BASES degraded condition in an orderly manner and takes into account the low probability of an event requiring LOVS occurring during this interval.
D.1 Condition D applies if the Required Actions and associated Completion Times are not met.
Required Action D.1 ensures that Required Actions for the affected DG inoperabilities are initiated. The actions specified in LCO 3.8.1 are required immediately.
SURVEILLANCE The following SRs apply to each DG-LOVS Function.
REQUIREMENTS SR 3.3.6.1 A CHANNEL FUNCTIONAL TEST is performed to ensure that the entire sensor channel will perform its intended function when needed. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specification tests at least once per refueling interval with applicable extensions.
which ... + *othat .. therm.. eme emse
'rare Ye...
. Aniy- s t÷i ig adj us.......s.. ...... .2. ............ ...i tlog t-he assumptions of the current plant specific setting SR 3.3.6.2 Surveillance Requirement 3.3.6.2 is the performance of a CHANNEL CALIBRATION . The CHANNEL CALIBRATION verifies the accuracy of each component within the sensor channel, except stepdown transformers, which are not calibrated. This includes calibration of the undervoltage relays and demonstrates that the equipment CALVERT CLIFFS - UNITS I & 2 B 3.3.6-7 Revision 36
DG-LOVS B 3.3.6 BASES falls within the specified operating characteristics defined by the manufacturer.
The SR verifies that the sensor channel responds to a measured parameter within the necessary range and accuracy.
CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive tests. CHANNEL CALIBRATIONS must be performed consistent with the plant-specific setting analysis.
The as-found and as-left values must also be recorded and reviewed for consistency with the assumptions of the SR interval extension analysis. The requirements for this review are outlined in Reference 4.
The settings, as well as the response to Loss of Voltage and Degraded Voltage tests, shall include a single point verification that the trip occurs within the required dela REFERENCES 1. UFSAR
- 2. CCNPP Setpoint File
- 3. IEEE No. 279, "Proposed IEEE Criteria for Nuclear Power Plant Protection Systems," August 1968
- 4. Calvert Cliffs Procedure EN-4-104, "Surveillance Testing" CALVERT CLIFFS - UNITS 1 & 2 B 3.3.6-8 Revision 36
CRS B 3.3.7 BASES SURVEILLANCE SR 3.3.7.1 REQUIREMENTS Performance of the CHANNEL CHECK or-u ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one sensor channel to a similar parameter on other channels. It is based on the assumption that sensor channels monitoring the same parameter should read approximately the same value.
Significant deviations between the two sensor channels could be an indication of excessive sensor channel drift in one of the channels or of something more serious. CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined by the plant staff, based on a qualitative assessment of the sensor channel that considers sensor channel uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the transmitter or the signal processing equipment has drifted outside its limits.
vna,ppt-Al
- foHt
,f aoutg,,, ,cf z of the axparnz tuhin denoral cpraticalso the rart e eseP eay SR 3.3.7.2 Proper operation of the actuation relays is verified by verification of the relay driver output signal. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specification Revision 36 UNITS 1 CLIFFS - UNITS CALVERT CLIFFS -
1&& 22 B 3.3.7-5 B 3.3.7-5 Revision 36
CRS B 3.3.7 BASES tests at least once per refueling interval with applicable extensions.
The Frequeic of 92 days i3,4ae on pi pat opmatraLI cx~ontac th rega to chalrd a yrfo r ABILIme, byh
-hr. de Tis clares ailut wthat is anuie Lhan uae CHAiNmEL agiven Funtio ; 19-J teiLr val u, e e.. nt.,
SR 3.3.7.3 A CHANNEL FUNCTIONAL TEST is performed on each containment radiation sensor channel to ensure the entire channel, except for sensor and initiating relays, will perform its intended function. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specification tests at least once per C *refueling interval with applicable extensions.
CHANNE with rAINrs ce of the 9se.nsrAchannel
......... CA LIBRATIbON~ is aQ chek IQ lotesfII;ensor ch nne indin i t he14.seeidmnsor. ThSuvilnethaterfes h te~oeofmae+
e-9 channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for sensor channel drift between successive calibrations to ensure that the channel remains operational between successive tests. CHANNEL CALIBRATIONS must be performed consistent with Reference 2.
instruF~mnts met beoing insri-- v uigpwroprto u 1-art of preparat'-o f0-1 -b-Cing placed im seri eic i-s a CHAMLEI CALVERT CLIFFS - UNITS 1 & 2 B 3.3.7-6 Revision 36
CRS B 3.3.7 BASES SR 3.3.7.5 GN CHANNEL FUNCTIONAL TEST is performed on tatuation circuitry. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specification tests at least once per refueling interval with applicable extensions.
This surveillance test verifies that the actuation push buttons are capable of opening contacts in the actuation logic as designed, de-energizing the actuation rela s and SR 3.3.7.6 This surveillance test ensures that the train actuation respoise times are less than or equal to the maximum times assumed in the analyses. Response times are defined in the same manner as ESF RESPONSE TIME. Response time testing acceptance criteria are included in Reference 1, -
,.*e~Y-ýting of the final act-uating devices, which make up the bulk of the response time, is included. Testing of the final actuating device is one channel is included in the testing of each actuation logic channel.
REFERENCES 1. UFSAR
- 2. CCNPP Setpoint File CALVERT CLIFFS - UNITS 1 & 2 B 3.3.7-7 Revision 36
CRRS B 3.3.8 BASES A.1, B.1, B.2. C.1, C.2.1. and C.2.2 Conditions A, B, and C are applicable to the CRRS trip circuit and measurement channel. Condition A applies to the failure of the CRRS trip circuit or measurement channel in MODE 1, 2, 3, or 4. Entry into this Condition requires action to either restore the failed channel or manually perform the CREVS function (Required Action A.1). The Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is sufficient to complete the Required Actions. If the channel cannot be restored to OPERABLE status, 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The Completion Times of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> for reaching MODEs 3 and 5 from MODE 1 are reasonable, based on operating experience and normal cooldown rates, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant safety systems or operators.
Condition C applies to the failure of the CRRS trip circuit or measurement channel when moving irradiated assemblies.
The Required Actions are immediately taken to place one OPERABLE CREVS train in the recirculation mode with post-LOCA fans in service or to suspend movement of irradiated fuel'assemblies. The Completion Time recognizes the fact that the radiation signal is the only Function available to initiate control room isolation in the event of a Fuel Handling Accident.
SURVEILLANCE SR 3.3.8.1 REQUIREMENTS Performance of the CHANNEL CHECKi * - .cur ensures that a gross failure of instrumentation has not occurred.
CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Acceptance criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the CALVERT CLIFFS - UNITS 1 & 2 B 3.3.8-3 Revision 2
CRRS B 3.3.8 BASES transmitter or the signal processing equipment has drifted outside its limit.
rcequiredehannel.. In adftii, a down-scal alarmi and up-SR 3.3.8.2 A CHANNEL FUNCTIONAL TEST is performed on the control room radiation monitoring channel to ensure the entire channel will perform its intended function. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specification tests at least once per refueling interval with applicable extensions.
including the sensor. The surveillance test verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for channel drift between successive calibrations to ensure that the channel remains operational between successive surveillance tests. CHANNE L CALIBRATIONS must be performed consistent with Reference 2 REFERENCES 1. UFSAR
- 2. CCNPP Setpoint File CALVERT CLIFFS - UNITS 1 & 2 B 3.3.8-4 Revision 36
CVCS Isolation Signal B 3.3.9 BASES Restoring at least one sensor channel to OPERABLE status is the preferred Required Action. If this cannot be accomplished, one channel should be placed in bypass and the other channel in trip. The allowed Completion Time of I hour is sufficient time to perform the Required Actions.
Once the Required Action to trip or bypass the channel has been complied with, Required Action C.2 provides for restoring one channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
The justification of the 48-hour Completion Time is the same as for Condition B.
After one channel is restored to OPERABLE status, the provisions of Condition C still apply to the remaining inoperable channel.
D.1 and D.2 Condition D specifies the shutdown track to be followed if the Required Actions and associated Completion Times of Condition A, B, or C are not met. If the Required Actions cannot be met 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
The Completion Times are reasonable, based on operating experience, to reach the required MODE from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.3.9.1 REQUIREMENTS Performance of the CHANNEL CHECK on each CVCS isolation pressure indicating sensor channel n erJ 1-2 houpr ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that sensor channels monitoring the same parameter should read approximately the same value.
Significant deviations between the two sensor channels could be an indication of excessive sensor channel drift in one of the channels or of something more serious. CHANNEL CHECK CALVERT CLIFFS - UNITS 1 & 2 B 3.3.9-5 Revision 26
CVCS Isolation Signal B 3.3.9 BASES will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined by the plant staff, based on a qualitative assessment of the sensor channel that considers sensor channel uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the transmitter or the signal processing equipment has drifted outside its limit.
ahctuatinyrlabu is v erf y vehif ic atioed on properany edperieree that dsinalstrate2 the rarity ehannl rf failur canneothe testedli~o at p wo ancd failuros inhredwunant chanelsin any 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per-iod is~ the CHIANNEL-CHCK mnizes the ehanac of less of pret. tive functon due to.
railture of redund~ant ehannel s. lt:%Th 1 CHANNLZCHC supplefmcnts less formal, b.t. Fflre frequent, eheeki; of SR 3.3.9.2 A CHANNEL FUNCTIONAL TEST is performed on each sensor channel to ensure the entire channel, except for sensor and initiation logic, will perform its intended function. A 3 successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of tests at least once per refueling interval with applicable Note 1 indicates proper operation of the individual actuation relays is verified by verification of proper relay driver output signal. Note 2 indicates that relays that cannot be tested at power are excepted from the SR while at CALVERT CLIFFS - UNITS I & 2 B 3.3.9-6 Revision 36
CVCS Isolation Signal B 3.3.9 BASES power. These relays must, however, be tested once per 24 months.
SR 3.3.9.3 CHANNEL CALIBRATION is a check of the sensor channel including the sensor. The surveillance test verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for sensor channel drift between successive calibrations to ensure that the channel remains operational between successive tests. CHANNEL CALIBRATIONS must be performed consistent with Reference 2.
The as-found and as-left values must also be recorded and reviewed for consistency with the assumptions of the SR interval extension analysis. The requirements for this review are outlined in Reference 4.
Radiation detectors may be removed and calibrated in a laboratory, calibrated in place using a transfer source, or replaced with an equivalent laboratory calibrated unit.
Tevices, wich make up the bulk of the response time, is included. Testing of the final actuating device in one channel is included in the testing of each actuation logic channel.
CALVERT CLIFFS - UNITS 1 & 2 B 3.3.9-7 Revision 36
PAM Instrumentation B 3.3.10 BASES SR 3.3.10.1 Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one indication channel to a similar parameter on other channels. It is based on the assumption that indication channels monitoring the same parameter should read approximately the same value. Significant deviations between the two indication channels could be an indication of excessive instrument drift in one of the channels or of something more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined by the plant staff, based on a qualitative assessment of the indication channel that considers indication channel uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit.
If the channels are within the criteria, it is an indication that the channels are OPERABLE. If the channels are normally off-scale during times when surveillance testing is required, the CHANNEL CHECK will only verify that they are off-scale in the same direction. Off-scale low current loop channels are verified to be reading at the bottom of the range and not failed down-scale.
- h~ requncyof 3 das i -ba-d upon plant op&ratinyt nxPezrietnccp ar toch eith
...l OPERABILITY am' 'rift, whic-h dcmenstrate that failure of muie than.- ncindcaio
....... a a given Function in anmy 31 day intervalis rp SR 3.3.10.2 Deleted.
Revision 32 UNITS 1 CLIFFS - UNITS CALVERT CLIFFS -
& 2 1 & 2 B 3.3.10-15 B 3.3.10-15 Revision 32
PAM Instrumentation B 3.3.10 BASES SR 3.3.10.3 CHANNEL CALIBRATION is a cc of the indication channel including the sensor. The SR verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION of the CIV position indication channels will consist of verification that the position indication changes from not-closed to closed when the valve is exercised to the isolation position as required by Technical Specification 5.5.8, Inservice Testing Program. The position switch is the sensor for the CIV position indication channels. A Note allows exclusion of neutron detectors, CETs, and reactor vessel level (HJTC) from the CHANNEL CALIBRATION.
REFERENCES 1. Letter from Mr. R. E. Denton (BGE) to NRC Document Control Desk, dated June 6, 1995, "License Amendment Request; Extension of Instrument Surveillance Intervals"
- 2. Letter from Mr. J. A. Tiernan (BGE) to NRC Document Control Desk, dated August 9, 1988, "Regulatory Guide 1.97 Review Update"
- 3. Regulatory Guide 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants To Assess Plant and Environs Conditions During and Following an Accident (Errata Published July 1981), December 1975
- 4. NUREG-0737, Supplement 1, Requirements for Emergency Response Capabilities (Generic Letter 82-33),
December 17, 1982
- 5. UFSAR, Chapter 7, "Instrumentation and Control" CALVERT CLIFFS - UNITS 1 & 2 B 3.3.10-16 Revision 32
Remote Shutdown Instrumentation B 3.3.11 BASES SURVEILLANCE SR 3.3.11.1 REQUIREMENTS Performance of the CHANNEL CHECK cn* wayensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one indication channel to a similar parameter on other channels. It is based on the assumption that indication channels monitoring the same parameter should read approximately the same value. Significant deviations between the indication channels could be an indication of excessive instrument drift in one of the channels or of something more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff, based on a qualitative assessment of the indication channel that considers indication channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. As specified in the SR, a CHANNEL CHECK is only required for those channels that are normally energized. If the channels are within the criteria, it is an indication that the channels are OPERABLE. If the channels are normally off-scale during times when surveillance testing is required, the CHANNEL CHECK will only verify that they are off-scale in the same direction. Off-scale low current loop channels are verified to be reading at the bottom of the range and not failed down-scale.
'~~ ,D ,, *T j, , ,,* , * ..... . .. ...... i m e,,pr , , ,a CALVERT CLIFFS - UNITS I & 2 B 3.3.11-4 Revision 26
Remote Shutdown Instrumentation B 3.3.11 BASES The SR is modified by a Note, which excludes neutron detectors and reactor trip breaker indication from the CHANNEL CALIBRATION.
REFERENCES 1. Updated Final Safety Analysis Report, Appendix 1C, "AEC Proposed General Design Criteria for Nuclear Power Plants" Revision 26 CALVERT CALVERT CLIFFS - 1&
UNITS I CLIFFS - UNITS &22 B 3.3.11-5 B 3.3.11-5 Revision 26
Wide Range Logarithmic Neutron Flux Monitor Channels B 3.3.12 BASES reduced. These Completion Times are also based on operating experience in performing the Required Actions and the fact that plant conditions will change slowly.
SURVEILLANCE SR 3.3.12.1 REQUIREMENTS Surveillance Requirement 3.3.12.1 is the performance of a CHANNEL CHECK on each required channel, A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based upon the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or of something more serious. CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined by the plant staff and should be based on a qualitative assessment of the indication channel that considers indication channel uncertainties, including control isolation, indication, and readability. If a channel is outside the criteria, it may be an indication that the transmitter or the signal processing equipment has drifted outside its limits. If the channels are within the criteria, it is an indication that the channels are OPERABLE.
G-ha~nnls in any 12 hlely in, eMA~NU I raim1re of i edumdamt ehammels. ChANNEL eC:I[CK sapplheut3 less formal, but more frequent, eheeks of channol EPRADiLITY -dut "yg noniiuiii l et~iivai ub-af d 1s pta~
B 3.3.12-3 Revision 2 UNITS 1 CLIFFS - UNITS CALVERT CLIFFS 1& & 22 B 3.3. 12-3 Revision 2
Wide Range Logarithmic Neutron Flux Monitor Channels B 3.3.12 BASES SR 3.3.12.2 A CHANNEL FUNCTIONAL TEST is performed once within 7 days prior to each reactor startup. This SR ensures that the entire channel is capable of properly indicating neutron flux. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specification tests at least once per refueling interval with applicable extensions. Internal test circuitry is used to feed pre-adjusted test signals into the preamplifier to verify channel alignment. It is not necessary to test the detector, because generating a meaningful test signal is difficult; the detectors are of simple construction, and any failures in the detectors will be apparent as change in channel output. This Frequency is the same as that employed for the same channels in the other applicable MODEs.
SR 3.3.12.3 Surveillance Requirement 3.3.12.3 is the erfor c of a CHANNEL CALIBRATION. etf 2The surveillance test is a complete check and rea justment of the wide range logarithmic neutron flux monitor channel from the preamplifier input through to the remote indicators. The surveillance test verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive surveillance tests. CHANNEL CALIBRATIONS must be performed consistent with the plant-specific setpoint analysis.
This SR is modified by a Note to indicate that it is not necessary to test the detector because generating a meaningful test signal is difficult; the detectors are of simple construction, and any failures in the detectors will i~n chnne u pt
~~ be apparent as ch ange CALVERT CLIFFS - UNITS 1 & 2 B 3.3.12-4 Revision 36
RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES a violation of this LCO occur, the operator should check whether or not an SL may have been exceeded.
ACTIONS A.1 Pressurizer pressure and RCS cold leg temperature are controllable and measurable parameters. Reactor Coolant System flow rate is not a controllable parameter and is not expected to vary during steady-state operation. With any parameter not within its LCO limit, action must be taken to restore the parameter.
The two hour Completion Time for restoration of the parameters provides sufficient time to adjust plant parameters, to determine the cause of the off normal condition, and to restore the readings within limits. The Completion Time is based on plant operating experience that shows the parameter can be restored in this time period.
B.1 If Required Action A.1 is not met 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 2 within six hours.
In MODE 2, the reduced power condition eliminates the potential for violation of the accident analysis bounds.
Six hours is a reasonable time that permits the plant power to be reduced at an orderly rate in conjunction with even control of steam generator (SG) heat removal.
SURVEILLANCE SR 3.4.1.1 REQUIREMENTS th-12 h~
1 iteur SRlFrqume fo -p-res brueratn pracitice b suf-flieitallUto ULth*I changes, tat thetliaiisieil prstieta regsulrleeAeLteU U*QIJ Thc dcrainand7stre-t-veifhopcut, i-feld lidsbethow 0 ~y b op-ratyimg paeutiee tý.
CALVERT CLIFFS - UNITS 1 & 2 B 3.4. 1-3 Revision 45
RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES SR 3.4.1.3 cd uasigthce iaanetersc flowtitremcutti.
,Arfours te1-2 ho-ur rpcc haz bneme fhor eln le eper atune x~isnet vfmifytortoceistirithat tafety analysis temei-atio.eIanb Measurem oentnof Ctoal flowtm ratead iste perfrmed 4-4fe~+=--This2 u veiiesthat theacta RCS flow rate i wierthin th bons te ofthe anlstf-ces.-t euail sesf Siys gfS col water leg~iniaion muay bheusdfre thiscoRe has described, in~
plan proceures UseJ of lthemaximum3 folod REFRECE 1.nUdiateFiona SafetyiAnayis Rcepotal anUFosArvaiva Secio 1412."latChrcersiconierdi R
Safety Analyis" o-RSttalFl CALfoERT CLIFFS UNITSle 1la&ntuf 2 th B-3..14heviin4 i
RCS P/T Limits B 3.4.3 BASES Besides restoring operation to within limits, an evaluation is required to determine if RCS operation can continue. The evaluation must verify that the RCPB integrity remains acceptable and must be completed before continuing operation. Several methods may be used, including comparison with pre-analyzed transients in the stress analyses, new analyses, or inspection of the components.
Reference 2,Section XI, Appendix E, may be used to support the evaluation. However, its use is restricted to evaluation of the vessel beltline.
The Completion Time of prior to entering MODE 4 forces the evaluation prior to entering a MODE where temperature and pressure can be significantly increased. The evaluation for a mild violation is possible within several days, but more severe violations may require special, event specific stress analyses or inspections.
Condition C is modified by a Note requiring Required Action C.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.
SURVEILLANCE SR 3.4.3.1 REQUIREMENTS V that operation is within limits is required
~when RCS P/T conditions are undergoing The SR for heatup, cooldown, or ISLH testing may be discontinued when the definition given in the relevant plant procedure for ending the activity is satisfied.
CALVERT CLIFFS - UNITS 1 & 2 B 3.4.3-7 Revision 47
RCS Loops - MODEs 1 and 2 B 3.4.4 BASES ACTIONS A._1 If the requirements of the LCO are not met, the Required Action is to reduce power and bring the plant to MODE 3.
This lowers power level and thus reduces the core heat removal needs, and minimizes the possibility of violating DNB limits. It should be noted that the reactor will trip and place the plant in MODE 3 as soon as the RPS senses less than 370,000 gpm** RCS flow.
The Completion Time of six hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging safety systems.
SURVEILLANCE SR 3.4.4.1 RFOIITRFMFNTS This SR requires verification & -of the required number of loops in operation. Verification includes flow rate, temperature, or pump status monitoring, which help to ensure that forced flow is providing heat removal while ma n ann g th ma***gi "-.L -
to DNB.*=,,*1
-* .. * -_* -.* .- .-. ' ,2FJ REFERENCES 1. UFSAR, Chapter 14, "Safety Analysis"
- The Rnater GCool*an System Flow Rate limit shall be Ž 340,000 gpm through Unit 2, Cycle 14.
Revision 13 CALVERT CLIFFS CALVERT -
UNITS 1 CLIFFS - UNITS 1&& 22 B 3.4.4-3 B 3.4.4-3 Revision 13
RCS Loops - MODE 3 B 3.4.5 BASES into the RCS with a boron concentration less than that required to meet the minimum SDM of LCO 3.1.1 is required to assure continued safe operation. When water is added without forced circulation, unmixed coolant could be introduced to the core, however water added with a boron concentration meeting the minimum SDM maintains an acceptable margin to subcritical operations. The immediate Completion Times reflect the importance of maintaining operation for decay heat removal.
SURVEILLANCE SR 3.4.5.1 REQUIREMENTS This SR requires verification -that the required number of RCS loops are in operation. Verification includes flow rate, temperature, and pump status monitoring, which help e re hat forced flow is providing heat SR 3.4.5.2 This SR requires verification that the secondary side water level in each SG is > -50 inches. An adequate SG water level is required in order to have a heat sink for emoval of the core decay heat from the reactor
~SR 3.4.5.3 Verification that the required number of RCPs are OPERABLE ensures that the single failure criterion is met and that an additional RCS loop can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment andpower availability to the reu d Revision 28 UNITS 1 CALVERT CLIFFS - UNITS CALVERT -
& 2 1 & 2 B 3.4.5-4 B 3.4.5-4 Revision 28
RCS Loops - MODE 4 B 3.4.6 BASES C.1 and C.2 If no RCS or SDC loops are OPERABLE or in operation, except during conditions permitted by Note 1 in the LCO section, all operations involving introduction of water into the RCS with a boron concentration less than that required to meet the minimum SDM of LCO 3.1.1, must be suspended and action to restore one RCS or SDC loop to OPERABLE status and operation must be initiated. The required margin to criticality must not be reduced in this type of operation.
Suspending the introduction of water into the RCS with a boron concentration less than that required to meet the minimum SDM of LCO 3.1.1 is required to assure continued safe operation. When water is added without forced circulation, unmixed coolant could be introduced to the core, however water added with a boron concentration meeting the minimum SDM maintains an acceptable margin to subcritical operations. The immediate Completion Times reflect the importance of decay heat removal. The action to restore must continue until one loop is restored to operation.
SURVEILLANCE SR 3.4.6.1 REQUIREMENTS This SR requires verification eth,,rs; hat one required loop is in operation. This ensures forced flow is providing heat removal. Verification inludes flow rate, temerature, or pump status " "tri
.T 1- er
&o"-feiet +rcIa ssyoess .... leap sa4 In-SR 3.4.6.2 This SR requires verification- of secondary side water level in the required SG(s) > -50 inches. An adequate SG water level is required in order to have a heat sink for removal of the core deca heat from the reactor colant. h orýf-ylha .
and verify operation wit-hin safety analyses assum~ption:.
CALVERT CLIFFS - UNITS 1 & 2 B 3.4.6-4 Revision 19
RCS Loops - MODE 4 B 3.4.6 BASES SR 3.4.6.3 Verification that the required pump is OPERABLE ensures that an additional RCS or SDC loop can be placed in operation, if needed to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the required loop components that are not in operation. For an RCS loop, the required component is a pump. For an SDC loop, the required comDonents are the pump and valves. /hz Frz~ucncv of REFERENCES None CALVERT CLIFFS - UNITS 1 & 2 B 3.4.6-5 Revision 19
RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES safe operation. When water is added without forced circulation, unmixed coolant could be introduced to the core, however water added with a boron concentration meeting the minimum SDM maintains an acceptable margin to subcritical operations. The immediate Completion Times reflect the importance of maintaining operation for decay heat removal.
SURVEILLANCE SR 3.4.7.1 REQUIREMENTS This SR requires verification ] that one SDC loop is in operation. Verification includes flow rate, temperature, or pump status monitoring, which help ensure imdieate leep statu.
The SOC flow is established to ensure that core outlet temperature is maintained sufficiently below saturation to allow time for swapover to the standby SOC loop should the operating loop be lost.
SR 3.4.7.2 Verifying the SGs are OPERABLE by ensuring their secondary side water levels are Ž -50 inches ensures that redundant heat removal paths are available if the second SDC loop is inoperable. This surveillance test is required to be performed when the LCO requirement is being met by use of the SGs. If both SDC loo s are OPERABLE, this SR is not needejd.ý4, 12 "h horezu~,h cnzonb peratn pract+ic- tc he 3suffi ient to rcgularly assess degradatieH and "rifyoper-ateion w4itti saFety anetl~yseq asufmpti~nz SR 3.4.7.3 Verification that the second SDC loop is OPERABLE ensures that redundant paths for decay heat removal are available.
The requirement also ensures that the additional loop can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is CALVERT CLIFFS - UNITS 1 & 2 B 3.4.7-5 Revision 42
RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES performed by verifying proper breaker alignment and power available to the required pumps and valves that are not in operation. This surveillance test is required to be performed when the LCO requirement is being met by one of two SDC loops, e.g., both SGs have < -50 inches water REFERENCES None CALVERT CLIFFS - UNITS 1 & 2 B 3.4.7-6 Revision 42
RCS Loops - MODE 5, Loops Not Filled B 3.4.8 BASES status and place it in operation must be initiated immediately. The required margin to criticality must not be reduced in this type of operation. Suspending the introduction of water into the RCS with a boron concentration less than that required to meet the minimum SDM of LCO 3.1.1 is required to assure continued safe operation. When water is added without forced circulation, unmixed coolant could be introduced to the core, however water added with a boron concentration meeting the minimum SDM maintains an acceptable margin to subcritical operations. The immediate Completion Time reflects the importance of maintaining operation for decay heat removal.
SURVEILLANCE SR 3.4.8.1 REQUIREMENTS This SR requires verification that one SDC loop is in operation. Verification includes flow rate, temperature, or pump status monitoring, which help ensure SVerification that the required number of loops are OPERABLE ensures that redundant paths for heat removal are available and that additional loops can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and indicated power available to the REFERENCES None CALVERT CLIFFS - UNITS 1 & 2 B 3.4.8-3 Revision 42
Pressuri zer B 3.4.9 BASES SURVEILLANCE SR 3.4.9.1 REQUIREMENTS This SR ensures that during steady-state operation, pressurizer water level is maintained below the nominal upper limit to provide a minimum space for a steam bubble.
The **indicated surveillance level test is. performed
.1u by observing the 1týIEIIU
(~1ee fer aydy I**~,
on a ndu,,a lver fi tdi tin.
SR 3.4.9.2 The SR is satisfied when the power supplies are demonstrated to be capable of producing the minimum power and the associated pressurizer heaters are verified to be at their design rating. (This may be done by testing the power supply output and by performing an electrical check on heater element continuity and resistance.) .e rqeGo 21~~ ~~ i3 cosidao adqut ta ..et-4tehat-"
month dgadfttieR-and as been shown by eprating cxpicrince to be REFERENCES 1. NUREG-0737, II.E.3.1, "Clarification of TMI Action Plan Requirements," November 1980 CALVERT CLIFFS - UNITS 1 & 2 B 3.4.9-5 Revision 2
Pressurizer PORVs B 3.4.11 BASES of one hour or place the associated PORVs in override closed and restore at least one block valve to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and the remaining block valve in five days, per Required Action C.2. The Completion Time of one hour to either restore the block valves or place the associated PORVs in override closed is reasonable based on the small potential for challenges to the system during this time and provides the operator time to correct the situation.
F.1 and F.2 If the Required Actions and associated Completion Times are not met, then the plant must be brought to a MODE in which the LCO does not apply. The plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce any RCS cold leg temperature
- 365 0 F (Unit 1), : 301°F (Unit 2) within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The Completion Time of six hours is reasonable, based on operating experience, to reach MODE 3 from full power in an orderly manner and without challenging safety systems. Similarly, the Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to reduce any RCS cold leg temperature
- 365 0 F (Unit 1),
< 301'F (Unit 2) is reasonable considering that a plant can cool down within that time frame. In MODE 3 with any RCS 0
cold leg temperature 5 365 F (Unit 1), ! 301OF (Unit 2) and in MODEs 4, 5, and 6, maintaining PORV OPERABILITY is required per LCO 3.4.12.
SURVEILLANCE SR 3.4.11.1 REQUIREMENTS A CHANNEL FUNCTIONA L is erformed on each PORV instrument channel 9to ensure the entire channel will perform its in en ed function when needed.
iIf the-block valve is cTo-sed to isolatea at is capable of being manually cycled, the OPERABILITY of the block valve is of importance because opening the block valve is necessary to permit the PORV to be used for manual control of RCS pressure. If the block valve is closed to isolate an otherwise inoperable PORV, the maximum Completion Time to restore the PORV and open the Revision 26 UNITS 1 CALVERT CLIFFS - UNITS CALVERT -
I& 2
& 2 B 3.4.11-6 B 3.4.11-6 Revision 26
Pressurizer PORVs B 3.4.11 BASES block valve is 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br />, which is well within the allowable limits (25%) to extend the block valve surveillance interval
- . Furthermore, these test requirements would be completed by the reopening of a recently closed block valve upon restoration of the PORV to OPERABLE status (i.e., completion of the Required Action fulfills the SR).
The Note modifies this SR by stating that this SR is not required to be performed with the block valve closed in accordance with the Required Actions of this LCO.
SR 3.4.11.3 Surveillance Requirement 3.4.11.3 requires complete cycling of each PORV. Power-operated relief valve cyclin demonstrates its function. 's--~em" ip n r l biSR q 3.4.a1.4
~Performance of a CHANNEL CALIBRATION on each required PORV hc2mon channel actuation thpra F rcquzraingzxcrince is required to adjust the whole channel so that it responds, and the valve opens REFERENCES 1. NUREG-0737, Paragraph II, G.I, "Clarification of TMI Action Plan Requirements," November 1980
- 2. Inspection and Enforcement Bulletin 79-05B, "Nuclear Incident at Three Mile Island - Supplement,"
April 21, 1979
- 3. ASME Code for Operation and Maintenance of Nuclear Power Plants CALVERT CLIFFS - UNITS 1 & 2 B 3.4.11-7 Revision 38
LTOP System B 3.4.12 BASES transient reasonable during the applicable MODEs. This action protects the RCPB from a low temperature overpressure event and a possible brittle failure of the reactor vessel.
The Completion Time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to depressurize and vent the RCS is based on the time required to place the plant in this condition and the relatively low probability of an overpressure event during this time period due to increased operator awareness of administrative control requirements.
SURVEILLANCE SR 3.4.12.1 and SR 3.4.12.2 REQUIREMENTS To minimize the potential for a low temperature overpressure event by limiting the mass input capability, verification that a maximum of one HPSI pump is only capable of manually injecting into the RCS, and automatic alignment of the HPSI loop MOVs, is prevented (by disabling the automatic opening features of the HPSI loop MOVs) is required. The HPSI pumps are rendered incapable of injecting into the RCS through removing the power from the pumps by racking the breakers out under administrative control or by verifying their discharge valves are locked shut.
The passive vent arrangement must only be open to be OPERABLE. This SR need only be performed if the vent is being used to satisfy-thejrqgjiements of this LCO. The mispositioning of wunlockcd and leeked vent valvc.s,
. espeetiYl ,
Revision 2 UNITS 1 CLIFFS - UNITS CALVERT CLIFFS -
I&& 2 2 B 3.4.12-11 B 3.4.12-11 Revision 2
LTOP System B 3.4.12 BASES SR 3.4.12.4 The PORV block valve must be verified open to provide the flow path for each required PORVtorperorm its function when actuated. The valve can be remotely verified open in the main Control Room.
The block valve is a remotely controlled MOV. The power to the valve motor operator is not required to be removed, and the manual actuator is not required locked in the inactive position. Thus, the block valve can be closed in the event the PORV develops excessive leakage or does not close (sticks open) after relieving an overpressure event.
The 7-2 In~ Fretqueney oprtigitciec- 4eeicrý*
accidental movcmcnt of le;h ngrmtcorlad pozitien indicatian ea-abilities aalbewcecsl moioe.These eonsziderationz include the adR4Mirýistre v 094nt=15 ve main Control Reom access and cquipmen Performance of a CHANNEL FUNCTIONAL TEST is requiredEE A Not o verify and, as necessary, adjust the PORV open setpoints. The CHANNEL FUNCTIONAL TEST will verify on a pefmonthly basis that the PORV lift setpoints are within the LCO limit. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required ASNeihiastben tssaddledniastionge this Srefeisgrqintredaolb performed 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after decreasing RCS cold leg temperature to *ý365 0F (Unit 1), *5301OF (Unit 2). The test cannot be performed until the RCS is in the LTOP MODEs when the PORV lift setpoint can be reduced to the LTOP setting. The test CALVERT CLIFFS - UNITS 1 & 2 B 3.4.12-12 Revision 36
LTOP System B 3.4.12 BASES must be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering the LTOP MODEs.
SR 3.4.12.6 Performance of a CHANNEL CALIBRATION on each required PORV actuation channel is required .y -P. to adjust the whole channel so that it responds and the valve opens within the required LTOP range and with accuracy to known input.
The 24 111ft Pi tquemy ClUisid, ope I at i~I. tj-- p~ m-t equipfflet reliahiit ýarmd miateheq the typpicalreuln REFERENCES 1. 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities
- 2. Generic Letter 88-11, NRC Position on Radiation Embrittlement of Reactor Vessel Materials and its Impact on Plant Operations, July 12, 1988
- 3. UFSAR, Section 4.2.2, Low Temperature Overpressure Protection
- 4. Generic Letter 90-06, Resolution of Generic Issues 70, "PORV and Block Valve Reliability," and 94, "Additional LTOP Protection for PWRs," June 25, 1990 Revision 36 CLIFFS - UNITS CALVERT CLIFFS -
UNITS 1 1& & 2 2 B 3.4.12-13 B 3.4.12-13 Revision 36
RCS Operational LEAKAGE B 3.4.13 BASES SURVEILLANCE SR 3.4.13.1 REQUIREMENTS Verifying RCS LEAKAGE to be within the LCO limits ensures the integrity of the RCPB is maintained. Pressure boundary LEAKAGE would first appear as unidentified LEAKAGE and can only be positively identified by inspection. Unidentified LEAKAGE and identified LEAKAGE are determined by performance of an RCS water inventory balance.
The RCS water inventory balance must be performed with the reactor at steady-state operating conditions (stable temperature, power level, pressurizer and makeup tank levels, and makeup and letdown). The surveillance is modified by two Notes. Note 1 states that this SR is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishing steady state operation. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.
Steady-state operation is required to perform a proper water inventory balance; calculations during maneuvering are not useful. For RCS operational LEAKAGE determination by water inventory balance, steady-state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal leakoff flows.
An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the automatic systems that monitor the containment atmosphere radioactivity and the containment sump level. These leakage detection systems are specified in LCO 3.4.14.
Note 2 states that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 100 gpd cannot be measured accurately by an RCS water inventory balance.
SR 3.4.13.2 This SR verifies that primary to secondary LEAKAGE is less or equal to 100 gpd through any one SG. Satisfying the Revision 46 UNITS 1 CLIFFS - UNITS CALVERT CLIFFS -
1&& 2 2 B 3.4.13-5 B 3.4. 13-5 Revision 46
RCS Operational LEAKAGE B 3.4.13 BASES primary to secondary LEAKAGE limi't ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this SR is not met, compliance with LCO 3.4.18, "Steam Generator Tube Integrity," should be evaluated. The 100 gpd limit is measured at hot plant conditions as described in Reference 4. The operational LEAKAGE rate limit applies to LEAKAGE through any one SG.
If it is not practical to assign the LEAKAGE to an individual SG, all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.
The Surveillance is modified by a Note which states that the Surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady-state operation. For the RCS primary to secondary LEAKAGE determination, steady-state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, and makeup and letdown.
REFERENCES 1. UFSAR
- 2. Regulatory Guide 1.45, Reactor Coolant Pressure Boundary Leakage Detection Systems, May 1973
- 3. Nuclear Energy Institute (NEI) 97-06, Steam Generator Program Guidelines
- 4. Electric Power Research Institute, Pressurized Water Reactor Primary-to-Secondary Leakage Guidelines
- 5. Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000 Revision 46 UNITS 1 CLIFFS - UNITS CALVERT CLIFFS -
1&& 2 2 B 3.4.13-6 B 3.4. 13-6 Revision 46
RCS Leakage Detection Instrumentation B 3.4.14 BASES the plant will not be operated in a degraded configuration for a lengthy time period.
D.1 and D.2 If any required Action of Conditions A, B, or C cannot be met 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 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.
E.1 If all required alarms and monitors are inoperable, no automatic means of monitoring leakage are available, an immediate plant shutdown in accordance with LCO 3.0.3 is required.
SURVEILLANCE SR 3.4.14.1 REQUIREMENTS Surveillance Requirement 3.4.14.1 requires the performance of a CHANNEL CHECK of the required containment atmosphere radioactivity monitors. The check gives reason le SR 3.4.14.2 Surveillance Requirement 3.4.14.2 requires the performance of a CHANNEL FUNCTIONAL TEST of the required containment atmosphere radioactivity monitors. The test ensures that the monitor can perform its function in the desired manner.
The test verifies the alarm setpoint and relative accuracy of the instrument string. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by CALVERT CLIFFS - UNITS 1 & 2 B 3.4.14-6 Revision 44
RCS Leakage Detection Instrumentation B 3.4.14 BASES other Technical Specification tests at least once per 3 These SRs require the performance of a CHANNEL CALIBRATION for each of the RCS leakage detection instrumentation channels. The calibration verifies the accuracy of the instrument string, includin the instruments located inside REFERENCES 1. UFSAR
- 2. Regulatory Guide 1.45, Revision 0, Reactor Coolant Pressure Boundary Leakage Detection Systems, May 1973 CALVERT CLIFFS - UNITS I & 2 B 3.4. 14-7 Revision 44
RCS Specific Activity B 3.4.15 BASES C.1 With the gross activity in excess of the allowed limit, the unit must be placed in a MODE in which the requirement does not apply.
The change within six hours to MODE 3 and RCS average temperature < 5007F lowers the saturation pressure of the reactor coolant below the setpoints of the main steam safety valves and prevents venting the SG to the environment in an SGTR event. The allowed Completion Time of six hours is required to reach MODE 3 below 500OF from full power conditions and without challenging plant systems.
SURVEILLANCE SR 3.4.15.1 REQUIREMENTS The SR requires performing a gamma isotopic analysis, as a measure of the gross activity of the reactor coolant e IWhile E is basically a quantitative measure of radionuclides with half lives longer than 15 minutes, excluding iodines, this gamma isotopic measurement is the sum of the degassed gamma activities and the gaseous gamma activities in the sample taken. This SR provides an indication of any increase in gross activity.
Trending the results of this SR allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions. The SR is applicable in MODEs 1 and 2, and in MODE 3 with RCS average temperature at
~SR 3.4.15.2 This SR is performed to ensure iodine remains within limits during normal operation and following fas pow changes after a power change of Ž 15% RTP within a one hour period, is established because the iodine levels peak during this time following fuel failure; samples at other times would provide inaccurate results.
CALVERT CLIFFS - UNITS 1 & 2 B 3.4.15-4 Revision 41
RCS Specific Activity B 3.4.15 BASES The SR is modified by a Note which requires the surveillance test to only be performed in MODE 1. This is required because the level of fission products generated in other MODEs is much less. Also, fuel failures associated with fast power changes is more apt to occur in MODE 1 than in MODEs 2 and 3.
SR 3.4.15.3 A radiochemical analysis for E determination is required
.with the plant operating in MODE 1 equilibrium conditions. The E determination directly relates to the LCO and is required to verify plant operation within the specified gross activity LCO limit.
The analysis for E is a measurement of the average energies per disintegration for isotopes with half lives longer than This SR has been modified by a Note that indicates sampling is not required to be performed until 31 days after 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. This ensures the radioactive materials are at equilibrium so that analysis for E is representative and not skewed by a crud burst or other similar abnormal event.
REFERENCES 1. UFSAR
- 2. Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000 CALVERT CLIFFS - UNITS 1 & 2 B 3.4. 15-5 Revision 41
STE-RCS Loops - MODE 2 B 3.4.16 BASES LCO This LCO is provided to allow for the performance of PHYSICS TESTS in MODE 2 (after a refueling), where the core cooling requirements are significantly different than after the core has been operating. Without this LCO, plant operations would be held bound to the normal operating LCOs for reactor coolant loops and circulation (MODEs 1 and 2), and the appropriate tests could not be performed.
In MODE 2, where core power level is considerably lower and the associated PHYSICS TESTS must be performed, operation is allowed under no flow conditions provided THERMAL POWER is
< 5% RTP and the reactor trip setpoints of the OPERABLE power level channels are set ! 15% RTP. These limits ensure no SLs or fuel design limits will be violated.
The exception is allowed even though there are no bounding safety analyses. These tests are allowed since they are performed under close supervision during the test program and provide valuable information on the plant's capability to cool down without offsite power available to the RCPs.
APPLICABILITY This LCO ensures that the plant will not be operated in MODE 1 without forced circulation. It only allows testing under these conditions while in MODE 2. This testing establishes that heat input from nuclear heat does not exceed the natural circulation heat removal capabilities.
Therefore, no safety or fuel design limits will be violated as a result of the associated tests.
ACTIONS A.1 If THERMAL POWER increases to > 5% RTP, the reactor must be tripped immediately. This ensures the plant is not placed in an unanalyzed condition and prevents exceeding the specified acceptable fuel design limits.
SURVEILLANCE SR 3.4.16.1 REQUIREMENTS THERMAL POWER must be verified to be within limits -S
( to ensure that the fuel design criteria are not during the etrformance
- '. of-the PHYSICS TESýTS tlviolated
- -4 CALVERT CLIFFS NT .. 62revso2 aýpten~tial ý 3**
-*on uf*.*n CALVERT CLIFFS - UNITS 1 & 2 B 3.4. 16-2 Revision 2
STE-RCS Loops - MODE 2 B 3.4.16 BASES ddQeogrzad ation And - cýrif4y@oper-a ti iz iIi the LCO lmt~
P-lan~t--epe-rationAs arc@ conductcd slow;ly dtuifirg thcpýfomac ofPHYSICS TESTS, And monito4rig trhi P~e level cne~ P-.
hourt iz suf-..ficrieQRto,*11' F'Cnzu'Jjre thet tile!V~Ilec doe-cz no cXcced thbe limit.
SR 3.4.16.2 Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of initiating startup or PHYSICS TESTS, a CHANNEL FUNCTIONAL TEST must be performed on each logarithmic power level neutron flux monitoring channel to verify OPERABILITY and adjust setpoints to proper values.
This will ensure that the RPS is properly aligned to provide the required degree of core protection during startup or the performance of the PHYSICS TESTS. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specification tests at least once per refueling interval with applicable extensions. The interval is adequate to ensure that the appropriate equipment is OPERABLE prior to the tests to aid the monitoring and protection of the plant during these tests.
REFERENCES 1. 10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,Section XI Revision 36 UNITS 1 CLIFFS - UNITS CALVERT CLIFFS -
I& & 2 2 B 3.4.16-3 B 3.4. 16-3 Revision 36
STE RCS Loops - MODES 4 and 5 B 3.4.17 BASES APPLICABILITY The LCO ensures that while within this LCO the plant will not be operated in any other MODE besides MODEs 4 and 5 without forced circulation. This is because the MODEs of Applicability for this Specification are MODEs 4 and 5.
This Specification allows testing and maintenance to be performed on the SDC System while SDC is required to be OPERABLE.
ACTIONS A.1 If one or more requirements of the LCO are not met, all activities being performed under this STE must be immediately suspended. These activities are local leak rate testing of the SDC penetration and maintenance on valves in the SDC System. The Completion Time to suspend these activities immediately ensures the plant is not placed in an unanalyzed condition and prevents exceeding the specified acceptable fuel design limits.
SURVEILLANCE SR 3.4.17.1 REQUIREMENTS Xenon reactivity must be verified to be within limits once within one hour prior to suspending the reactor coolant circulation requirements of LCO 3.4.6, LCO 3.4.7, and LCO 3.4.8. The frequency of once within one hour prior to suspending the applicable RCS Loops LCO will ensure that the xenon reactivity is within limits and trending toward stability prior to suspending forced flow cooling. This will ensure no SLs or fuel design limits will be violated while testing or maintenance are being conducted.
SR 3.4.17.2 and SR 3.4.17.3 Verifying the charging pumps are de-energized and the charging flow paths are isolated, ensures that the major source of a boron reduction is not available. These two SRs support the requirement that no source be available tha_
CALVERT CLIFFS - UNITS I & 2 B 3.4. 17-2 Revision 34
STE RCS Loops - MODES 4 and 5 B 3.4.17 BASES Subsequent performance of these SRs after the initial verification that the charging pumps are de-energized and the charging flow paths are isolated, may be performed administratively.
SR 3.4.17.4 This SR requires that a SHUTDOWN MARGIN verification be prmediin accordance wih. uR 3.r 1 .1 RF CN eonei.te MARG1 l REFERENCES None CALVERT CLIFFS - UNITS 1 & 2 B 3.4. 17-3 Revision 34
SITs B 3.5.1 BASES conditions in an orderly manner and without challenging plant systems.
D.1 If more than one SIT is inoperable, the unit is in a condition outside the accident analyses. Therefore, LCO 3.0.3 must be entered immediately.
SURVEILLANCE SR 3.5.1.1 D*nIITPPMPNT',
l\ L. l,(l,# L i
- m=
Verification Wthat each SIT isolation valve is fully open, as -indicated in the Control Room, ensures that SITs are available for injection and ensures timely discovery if a valve should be partially closed. If an isolation valve is not fully open, the rate of injection to the RCS would be reduced. Although a motor-operated valve should not change position with power removed, a closed valve could result in not meeting accident analysis SR 3.5.1.2 and SR 3.5.1.3 Safety injection tank borated water volume and nitrogen cover pressure should be verified to be within specified limits-e a dequate in .ection
\
SR 3.5.1.4 Six months is reasonable for verification by sampling to determine that each SIT's boron concentration is within the required limits, because the static design of the SITs limits the ways in which the concentration can be changed.
f-q - _Arom 6-ccir mcai ly s, qu~ urW~~~~~L ICL iieakg CALVERT CLIFFS - UNITS 1 & 2 B 3.5.1-7 Revision 43
SITs B 3.5.1 BASES Verification consists of monitorJin9 inleakage or sampling.
The inleakage is monitored '(by monitoring tank level. Sampling of each tank is done . All intentional sources of level increase are maintaine administratively to ensure SIT boron concentrations are within technical specification limits. The boron concentration of each tank is verified prior to startup from outages. A sample of the SIT is required, to verify boron concentration, if 10 inches or greater of inleakage has occurred since last sampled.
Sampling the affected SIT (by taking the sample at the discharge of the operating HPSI pump) within one hour prior to a 1% volume increase of normal tank volume, will ensure the boron concentration of the fluid to be added to the SIT is within the required limit prior to adding inventory to the SIT(s).
SR 3.5.1.5 Verification that power is removed from each SIT isolation valve operator, by maintaining the feeder breaker open under administrative control, when the pressurizer pressure is Ž 2000 psig ensures that an active failure could not result in the undetected closure of an SIT motor-operated isolation valve. If this were to occur, only two SITs would be available for inection given a single failure coinien ih *-
.A.altnad This SR allows power to be supplied to the motor-operated isolation valves when RCS pressure is < 2000 psig, thus allowing operational flexibility by avoiding unnecessary delays to manipulate the breakers during unit startups or shutdowns. Even with power supplied to the valves, inadvertent closure is prevented by the RCS pressure interlock associated with the valves. Should closure of a valve occur in spite of the interlock, the safety injection signal provided to the valves would open a closed valve in the event of a LOCA.
CALVERT CLIFFS - UNITS 1 & 2 B 3.5. 1-8 Revision 43
ECCS - Operating B 3.5.2 BASES An event accompanied by a loss of offsite power and the failure of an emergency diesel generator can disable one ECCS train until power is restored. A reliability analysis (Reference 3) has shown that the impact with one full ECCS train inoperable is sufficiently small to justify continued operation for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Reference 4 describes situations in which one component, such as a SDC total flow control valve, can disable both ECCS trains. With one or more components inoperable, such that 100% of the equivalent flow to a single OPERABLE ECCS train is not available, the facility is in a condition outside the accident analyses. Therefore, LCO 3.0.3 must be immediately entered.
B.1 and B.2 If the inoperable train 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and pressurizer pressure reduced to
< 1750 psia within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power in an orderly manner and without challenging unit systems.
SURVEILLANCE SR 3.5.2.1 REQUIREMENTS Verification of proper valve position ensures that the flow path from the ECCS pumps to the RCS is maintained.
Misalignment of these valves could render both ECCS trains inoperable. MOV-659 and MOV-660 are secured in position by interrupting the control signal to the valve operator via a key switch in the Control Room. Power is removed from the valve operator for CV-306 by isolating the air supply to the valve positioner. These actions ensure that the valves cannot be inadvertently misaligned. These valves are of the type described in Reference 4, which can disable the function of both ECCS trains and invalidate the accident ans'is:. A 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Fregueney is Lul=deved reSU i CALVERT CLIFFS - UNITS 1 & 2 B 3.5.2-6 Revision 46
ECCS - Operating B 3.5.2 BASES SR 3.5.2.2 Verifying the correct alignment for manual, power-operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS 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 that receives an actuation signal is allowed to be in a non-accident position provided the valve automatically repositions within the proper stroke time. 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.
SR 3.5.2.3 Periodic surveillance testing of the HPSI and LPSI pumps to detect gross degradation caused by impeller structural damage or other hydraulic component problems is required by the American Society of Mechanical Engineers Code. This type of testing may be accomplished by measuring the pump developed head at only one point of the pump characteristic curve. This verifies both that the measured performance is within an acceptable tolerance of the original pump baseline performance and that the performance at the test flow is greater than or equal to the performance assumed in the unit safety analysis. Surveillance Requirements are specified in the Inservice Testing Program, which encompasses American Society of Mechanical Engineers Code. American Society of Mechanical Engineers Code provides the activities and Frequencies necessary to satisfy the requirements.
SR 3.5.2.4 The Surveillance Requirement was deleted in Amendment Nos. 260/237.
CALVERT CLIFFS - UNITS 1 & 2 B 3.5.2-7 Revision 38
ECCS - Operating B 3.5.2 BASES SR 3.5.2.5. SR 3.5.2.6. and SR 3.5.2.7 These SRs demonstrate that each automatic ECCS valve actuates to the required position on an actual, or simulated SIAS, and on a recirculation actuation signal; that each ECCS pump starts on receipt of an actual or simulated SIAS; and that the LPSI pumps stop on receipt of an actual or simulated recirculation actuation signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. In order to assure the results of the low temperature overpressure protection analysis remain bounding, whenever flow testing into the RCS is required at RCS temperatures _<365°F (Unit 1), *_ 301°F (Unit 2), the HPSI pump shall recirculate RCS water (suction from the RWT isolated) or the requirements of LCO 3.4.12, shall be aRydrn -,a lm outage... and the potent*al for unplaned
.... :i ,.tq if the ...... -'I anee te÷ ........
t ......
ne yv ,,
)*" F r* gtoratwi..... Thc 24 ,,,u,,t,, Feque,,cy is ctiso n-utpttle actu ionogiciseted spats one the Eengineere Sfty atainoiisetdapatoteEgneedog Safety Feature Actuation System testing, and equipment performance is monitored as part of the Inservice Testing Program.
~SR 3.5.2.8 thz Petriode inec peteitanee test ed th onundere twi hetheacesn tn oit spcwcr cndintzifn that zurvcly ands
- en lanc
". confirmed by eperatnrg eXpel'i-nco.
SR 3.5.2.9 Verifying that the SDC System open-permissive interlock is OPERABLE ensures that the SDC suction isolation valves are prevented from being remotely opened when RCS pressure, is at or above, the SDC System design suction pressure of CALVERT CLIFFS - UNITS 1 & 2 B 3.5.2-8 Revision 38
ECCS - Operating B 3.5.2 BASES 350 psia. The suction piping of the LPSI pumps, is the SDC component with the limiting design pressure rating. The interlock provides assurance that double isolation of the SDC System from the RCS is preserved whenever RCS pressure, is at or above, the design pressure. The 309 psia value specified in the Surveillance is the actual pressurizer pressure at the instrument tap elevation for PT-103 and PT-103-1 when the SDC System suction pressure is 350 psia.
The procedure for this surveillance test contains the required compensation to be applied to this value to account for instrument uncertainties. This surveillance test is normally performed usin a simulated RCS pressurei REFERENCES 1. UFSAR
- 2. 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Plants"
- 3. Nuclear Regulatory Commission Memorandum to V. Stello, Jr., from R. L. Baer, "Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975
- 4. Inspection and Enforcement Information Notice No. 87-01, "RHR Valve Misalignment Causes Degradation of ECCS in PWRs," January 6, 1987 CALVERT CLIFFS - UNITS 1 & 2 B 3.5.2-9 Revision 38
RWT B 3.5.4 BASES restore the RWT to OPERABLE status is based on this condition simultaneously affecting multiple redundant trains.
C.1 and C.2 If the RWT 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 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 3.5.4.1 and SR 3.5.4.2 REQUIREMENTS Refueling water tank borated water temperature shall be verified to be within the limits assumed in The SRs are modified by a Note that eliminates the requirement to perform this surveillance test when ambient air temperatures are within the operating temperature limits of the RWT. With ambient temperatures within this range, the RWT temperature should not exceed the limits.
Surveillance Requirement 3.5.4.2 is modified by an additional Note which requires the SR to be met in MODE 1 only. A SR is "met" only when the acceptance criteria are satisfied. Known failure of the requirements of a SR, even without a surveillance test specifically being "performed,"
constitutes a SR not "met." This reflects the maximum coolant temperature assumptions in the LOCA analysis.
SR 3.5.4.3 Above minimum RWT water volume level shall be verifiedk )I This Frequency ensures that a sufficient initial water supply is available for injection and to support continued ESF pump operation on recirculation.
Revision 2 CALVERT CLIFFS CALVERT UNITS 1 CLIFFS - UNITS
& 2 1 & 2 B 3.5.4-5 B 3.5.4-5 Revision 2
RWT B 3.5.4 BASES Boron concentration of the RWT shall be verified<6E to be within the required range. This Frequency ensures that the reactor will remain subcritical following a LOCA. Further, it ensures that the resulting sump pH will be maintained in an acceptable range such that boron precipitation in the core will not occur earlier than predicted, and the effect of chloride and caustic stress corrosion on mechanical systems and components will be REFERENCES 1. UFSAR, Chapters 6, "Engineered Safety Features," and 14, "Safety Analysis" B 3.5.4-6 Revision 2 1&
UNITS 1 CLIFFS - UNITS CALVERT CLIFFS & 2 2 B 3.5.4-6 Revision 2
STB B 3.5.5 BASES 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.5.5.1 REQUIREMENTS Periodic determination of the equivalent weight of STB in Containment must be performed due to the possibility of leaking valves and components in the Containment Building that could cause dissolution of the STB during normal
% operation. A is required to determine visually that a minimum of 13,750 lbm is contained in the STB baskets. This requirement ensures that there is an adequate mass of STB to adjust the pH of the post-LOCA sump solution to a value Ž 7.0.
access to the STB baskets is only feasible during adtages, SR 3.5.5.2 Testing must be performed to ensure the solubility and buffering ability of the STB after exposure to the containment environment. A representative sample of 2.74 +/- 0.05 grams of STB, from one of the baskets in Containment is submerged in 1.0 +/- 0.01 liters of water at a boron concentration of 3074 +/- 50 ppm, and at the standard temperature of 120 +/- 5°F. Within four hours without agitation, the solution is decanted and mixed, the temperature adjusted to 77 +/- 20 F, and the pH measured. The solution pH should be Ž 7.0. The representative sample weight is based on the minimum required STB equivalent weight of 13,750 lbm, and maximum possible post-LOCA sump water mass of 4,608,356 Ibm, normalized to buffer a 1.0 +/- 0.01 liter sample. The boron concentration of the CALVERT CLIFFS - UNITS 1 & 2 B 3.5.5-4 Revision 37
STB B 3.5.5 BASES test water is representative of the maximum possible boron concentration corresponding to the maximum possible post-LOCA sump volume. Agitation of the test solution is prohibited, since an adequate standard for the agitation intensity cannot be specified. A test time of four hours would allow time for the dissolved STB to naturally diffuse through the sample solution. A test time of less than four hours is more conservative than a test time of longer than four hours because the longer time could permit additional STB to dissolve, if excess STB was available. In the post-LOCA containment sump, rapid mixing would occur, significantly decreasing the actual amount of time before the required pH is achieved. This would ensure compliance with the Standard Review Plan requirement of a pH Ž 7.0 by he onset of recirculation after a LOCA.
REFERENCES None Revision 37 UNITS 1 CLIFFS - UNITS CALVERT CLIFFS -
& 2 1 & 2 B 3.5.5-5 B 3.5.5-5 Revision 37
Containment Air Locks B 3.6.2 BASES designed and that simultaneous opening of the inner and outer doors 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 air lock is used for entry and exit (procedures require strict adherence to single door opening), this test iso er 24y m n scP . Th E
re to b aped qulenhir pdoal n+f f or l oss -* - I.t4,,.-G-I 4.n *.j
-d t 21~~ tpee.Te2 month Fregqny juger h nt and f uhedonegneern i n a nmsid-Pre-d -adcquate given that -the ifttrerlck REFERENCES 1. UFSAR CALVERT CLIFFS - UNITS I & 2 B 3.6.2-8 Revision 2
Containment Isolation Valves B 3.6.3 BASES Required Action C.2 is modified by a Note that applies to valves and blind flanges, 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. Therefore, the probability of misalignment of these valves, once they have been verified to be in the proper position, is small.
D.1 and D.2 If the Required Actions and associated Completion Times are not 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 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 3.6.3.1 REQUIREMENTS This SR ensures that the containment vent valves are closed as required, or, if open, open for an allowable reason. If a containment vent valve is 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 SR is not required to be met when the containment vent valves are open for pressure control, ALARA or air quality considerations for personnel entry, or for surveillance tests that require the valves to be open. The containment vent valves are capable of closing in the environment, following a LOCA. Therefore, these valves are valvorcquiemets discussed in SR 3.6.3.2-.
SR 3.6.3.2 This SR requires verification that each containment isolation manual valve and blind flange located outside the Containment Structure, and not locked, sealed, or otherwise CALVERT CLIFFS - UNITS 1 & 2 B 3.6.3-8 Revision 32
Containment Isolation Valves B 3.6.3 BASES secured, and 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 containment boundary is within design limits. This SR does not require any testing or valve manipulation. Rather, it involves verification, through a system walkdown, that those containment isolation valves outside the Containment corr ct. . . . n..no Structure and capable of being mis ositioned are in the 11 ....... ... ........pres.i.. Gr Ctruettire~1 isireitelyiay'--3Pa a~.. u,... *..*
.... . ........ ation valves that are open under administrative controls are not required to meet the SR during the time the valves are open.
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.
The Note applies to valves and blind flanges 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 during MODEs 1, 2, 3, 4 and for ALARA reasons.
Therefore, the probability of misalignment of these containment isolation valves, once they have been verified to be in the proper position, is small.
SR 3.6.3.3 This SR requires verification that each containment isolation manual valve and blind flange located inside the Containment Structure, and not locked, sealed, or otherwise secured, and 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 containment boundary is within design limits. For containment isolation valves inside the Containment Structure, the Frequency of "prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days" is appropriate, since these containment isolation valves are operated under administrative controls and the probability CALVERT CLIFFS - UNITS 1 & 2 B 3.6.3-9 Revision 32
Containment Isolation Valves B 3.6.3 BASES of their misalignment is low. Containment isolation valves that are open under administrative controls are not required to meet the SR during the time that they are open. 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.
The Note allows valves and blind flanges located in high radiation areas 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 during MODEs 1, 2, and 3 for ALARA reasons. Therefore, the probability of misalignment of these containment isolation valves, once they have been verified to be in their proper position, is small.
SR 3.6.3.4 Verifying that the isolation time of each automatic power operated containment isolation valve is within limits is required to demonstrate OPERABILITY. The isolation time test, ensures the valve will isolate in a time period less than or equal to that assumed in the safety analysis. The isolation time and Frequency of this SR are in accordance with the Inservice Testing Program. The isolation time limits are contained in Reference 2.
SR 3.6.3.5 Automatic containment isolation valves close on an isolation signal [containment isolation signal Channels A or B, or safety injection actuation signal (SIAS) Channels A or B] to prevent leakage of radioactive material from the Containment Structure following a DBA. This SR ensures each automatic containment isolation valve will actuate to its isolation position on a containment isolation actuation signal. This surveillance test is not required for valves that are locked, sealed, or otherwise secured in the r *rd position under administrative controls.
4 2
rnnuunt. I ,,n , n-1 ^-n ra, n4 , + 4- ..- J-J L- 4k SR be performed only during a unit outage, inc sIelattio,,
of-pecnctratiens would eliminate coin - - A.a wAate r fI -- an d-CALVERT CLIFFS - UNITS 1 & 2 B 3.6.3-10 Revision 32
Containment Isolation Valves B 3.6.3 BASES disrupt nor-mal operation ef many crtQ lc~pnnz 0per-atina Pxppripnrp cshn that these eefmpenRt4. 11 IT7-pass this R hi fumdu I243t Freueicy Therefor~e, the Frcgucr-cy was conc~luded to beF acctal ur REFERENCES 1. UFSAR, Chapter 5, "Structures", Figure 5-10
- 2. UFSAR, Chapter 5, "Structures", Table 5-3 CALVERT CLIFFS - UNITS 1 & 2 B 3.6.3-11 Revision 32
Containment Pressure B 3.6.4 BASES B.1 and B.2 If containment pressure 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 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 3.6.4.1 REQUIREMENTS Verifying that containment pressure is within limits ensures that operation remains *thin th imi as med in the
~ariaions urin the applicable MD. Furt erffl ,th indic~tic*.aapk tEFEENCt thneiale i LIeCnrl can Roiii1atn lri pres-d REFERENCES None Revision 2 UNITS 1 CLIFFS - UNITS CALVERT CLIFFS - 1& & 2 2 B 3.6.4-3 B 3.6.4-3 Revision 2
Containment Air Temperature B 3.6.5 BASES within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 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 3.6.5.1 REQUIREMENTS Verifying that containment average air temperature is within the LCO limit ensures that containment operation remains within the limit assumed for the containment analyses. In order to determine the containment average air temperature, an arithmetic average is calculated using measurements taken from the containment dome [1(2)-TI-5309] and the containment reactor cavity [1(2)-TI-5311] temperature indicators selected to provide a representative sample of the overall REEENE . UFSAR, FA, Sect REEhCur i~onz14.20, etin1.2,"ontainment " tina R iesp n of the dCount Response" RJ CALVERT CLIFFS - UNITS 1 & 2 B 3.6.5-3 Revision 3
Containment Spray and Cooling Systems B 3.6.6 BASES E.1 and E.2 If the Required Actions and associated Completion Times of Conditions C or D of this LCO are not met, the plant must be I 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. 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.
F.1 With two containment spray trains or any combination of three or more Containment Spray and Cooling Systems trains inoperable, the unit is in a condition outside the accident analysis. Therefore, LCO 3.0.3 must be entered immediately.
SURVEILLANCE SR 3.6.6.1 REQUIREMENTS Verifying the correct alignment for manual, power-operated, and automatic valves in the containment spray flow path provides assurance that the proper flow paths will exist for Containment Spray System operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these were verified to be in the correct position prior to being secured. This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require any testing or valve manipulation. Rather, it involves verifying, through a system walkdown, that those valves outside the Containment Structure and capable of potentially being mispositioned are in the correct position.
SR 3.6.6.2 Starting each containment cooling train fan unit from the Control Room and operating it for __15 minutes ensures that all trains are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or motor cretve failure, actioor excessive akn vibration can be detected and CALVERT CLIFFS - UNITS 1 & 2 B 3.6.6-7 Revision 2
Containment Spray and Cooling Systems B 3.6.6 BASES Jeecpdcasdering VT.V- the knawm reliability of the Fail
'Init a~tios,'tetw a~d traim redundancy ayailabic, ad t~he low prgbabili-ty of-a 9igmifieant degradation ofth eentainmment. eeeling tri ocrimg between surve~iancc3 SR 3.6.6.3 Verifying a service water flow rate of Ž 2000 gpm to each cooling unit when the full flow service water outlet valves are fully open provides assurance that the design flow rate assumed in the safety analyses will be achieved (Reference 1, Chapter 7). Also considered in selecting this Frequenc were the known reliability of the Service Water
- f 31 doo-*j* System, the two train redundancy, and the low probability of a significant degradation of flow occurring between surveillance tests.
SR 3.6.6.4 Verifying that each containment spray pump's developed head at the flow test point is greater than or equal to the required developed head ensures that spray pump performance has not degraded during the cycle. Flow and differential pressure are normal tests of centrifugal pump performance required by Reference 2. Since the containment spray pumps cannot be tested with flow through the spray headers, they are tested on recirculation flow. This test confirms one point on the pump design curve and is indicative of overall performance. Such inservice inspections confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. The Frequency of this SR is in accordance with the Inservice Testing Program.
SR 3.6.6.5 and SR 3.6.6.6 These SRs verify that each automatic containment spray valve actuates to its correct position and that each containment spray pump starts upon receipt of an actual or simulated actuation signal (i.e., the appropriate Engineered Safety Feature Actuation System signal). This SR is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. The CALVERT CLIFFS - UNITS 1 & 2 B 3.6.6-8 Revision 2
Containment Spray and Cooling Systems B 3.6.6 BASES 24 molnth Frcfeq ~ Dsd nde,~ o efr hs r..illa... e test&....dcr.the etios tat apply d-urig a plaHt outage and the p*te.t.al for an unpla1 etr i Lh^ 3ur;'illoncc +e*+z w:ere prformed ,with thz ..... o a^t i power. Operating cxpcricnce has l that these cmmnennn ually pss;, the zurveillance tests whe..perf d
..... .at-te 24 .... h Fr-equ-e..ey. Therefore, the-k F10ucq-*y was cmt,,u.,
The surveillance test of containment sump isolation valves is also required by SR 3.5.2.5. A single surveillance test may be used to satisfy both requirements.
SR 3.6.6.7 This SR verifies that each containment cooling train actuates upon receipt of an actual or simulated actuation signal (i.e., the appropriate Engineered Safety Feature Actuation Svstem signal).Gba
~nenirerrgjdger ad -bkn hon o e ceptablo through operatin; expei i-enL. See SR .G5an Sn- 3.6.6.61 above, frfrhrd~uzc ft~b~f the 21 month Froguency.
SR 3.6.6.8 With the containment spray inlet valves closed and the spray header drained of any solution, low pressure air or smoke can be blown through check valve bonnets. Performance of this SR demonstrates that each spray nozzle is unobstructed and provides assurance that spray coverage of the Containment Structure during an accident is not degraded.
Due to the passive design of the nozzle, a test after maintenance that could result in nozzle blockage is considered adequate. Maintenance that could result in nozzle blockage is generally loss of foreign material control or a flow of borated water through a nozzle. Should either of these events occur, a supervisory evaluation will be required to determine whether nozzle blockage is a possible result of the event.
B 3.6.6-9 Revision 18 CALVERT CLIFFS - UNITS 1 CLIFFS - UNITS & 2 1 & 2 B 3.6.6-9 Revision 18
IRS B 3.6.8 BASES
- b. The fact that, even with no IRS train in operation, almost the same amount of iodine would be removed from the containment atmosphere through absorption by the Containment Spray System; and
- c. The fact that the Completion Time is adequate to make most repairs.
B.1 If two IRS trains are inoperable, one must be restored to OPERABLE status within one hour. The one hour Completion Time allows the swing train to be aligned to the appropriate bus to ensure each of the two remaining trains are powered from separate and independent buses. The one hour, also allows time to restore one train to OPERABLE status prior to initiating a plant shutdown. This is reasonable considering that a plant shutdown is a plant transient.
C.1 and C.2 If the IRS train 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 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 3.6.8.1 REQUIREMENTS Initiating each IRS train from the Control Room and operating it for Ž 15 minutes ensures that all trains are OPERABLE and that all associated controls are functioning properly. It also ensures that motor failure can be detected for corrective action.
~3 and- cotoH h w ri cudnyaaabl, anid the iodipndentmya aabl of the IRSanfe Revision 41 CALVERT CLIFFS UNITS 11 &
CLIFFS - UNITS
& 2 2 B 3.6.8-3 B 3.6.8-3 Revision 41
IRS B 3.6.8 BASES SR 3.6.8.2 This SR verifies that the required IRS filter testing is performed in accordance with the Ventilation Filter Testing Program. The IRS filter tests are in accordance with portions of Reference 2. The Ventilation Filter Testing Program includes testing high efficiency particulate air 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 Ventilation Filter Testing Program.
SR 3.6.8.3 The automatic startup test verifies that both trains of equipment start upon receipt of an actual or simulated test u.400l tV,s -A 1 ance v,=,1 the 1ei test -asd9 e nuvilac wei-e pe, ,1U...... hen et performeda
...... .. ....... h-*t month
.. Frequency Therefore, the Fr.qu...* was concluded tcbe acceptable fr9im a reliatbiliy tandpoint.
Furthcr4mrc, the Frcquincy Was developed coisei--,gtihatt th 3ystom ecjipmrt OPE[AHILITY is dem,,,u,,sated onr a 31 da-REFERENCES 1. UFSAR
- 2. Regulatory Guide 1.52, Revision 2, "Design, Testing, and Maintenance Criteria for Postaccident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants," March 1978 B 3.6.8-4 Revision 41 CALVERT CLIFFS - UNITS 1 CLIFFS - UNITS 1 && 22 B 3.6.8-4 Revision 41
AFW System B 3.7.3 BASES prior to locking, sealing, or securing. This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require any testing or valve manipulations; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position.
f~w--* eerat*,and emsures cc.,eft v_`ve pesitienr&.
SR 3.7.3.2 Cycling each testable, remote-operated valve that is not in its operating position, provides assurance that the valves will perform as required. Operating position is the position that the valve is in during normal plant operation.
This is accomplished by cycling each valve at least one cycle. This SR ensures that valves required to function during certain scenarios, will be capable of being properly positioned. The Frequency is based on engineering judgment that when cycled in accordance with the Inservice Testing Program, these valves can be placed in the desired position when required.
SR 3.7.3.3 Verifying that each AFW pump's developed head at the flow test point is greater than or equal to the required developed head (Ž 2800 ft for the steam-driven pump and
> 3100 ft for the motor-driven pump), ensures that AFW pump performance has not degraded during the cycle. Flow and differential head are normal tests of pump performance required by Reference 2. Because it is undesirable to introduce cold AFW into the steam generators while they are operating, this testing is performed on recirculation flow.
This test confirms one point on the pump design curve and is indicative of overall performance. Such inservice tests confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance.
Performance of inservice testing, discussed in Reference 2, at three month intervals satisfies this requirement.
Revision 26 CALVERT CALVERT CLIFFS -
UNITS 1 CLIFFS - UNITS 1&& 2 2 B 3.7.3-8 B 3.7.3-8 Revision 26
AFW System B 3.7.3 BASES This SR is modified by a Note indicating that the SR should be deferred up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> until suitable test conditions are established. This deferral is required because there is an insufficient steam pressure to perform the test.
SR 3.7.3.4 This SR ensures that AFW can be delivered to the appropriate steam generator, in the event of any accident or transient that generates an AFAS signal, by demonstrating that each automatic valve in the flow path actuates to its correct position on an actual or simulated actuation signal (verification of flow-modulating characteristics is not required). This SR is not required for valves that are locked, sealed, or otherwise secured in the re uquired ThsR e r h e p s stat in te evnt Thi 21 mont Fndreqdmincistrtvaccoptrolbsd cn _thedei ofhanye aecentor transient tnf by dem 40onstratingthatweeac AefWopmpd sv.taertsautomaticlyo This SR is modified by a Note. The Note indicates that the SR should be deferred up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> until suitable test conditions are established.
SR 3.7.3.6 This SR ensures that the AFW system is capable of providing a minimum nominal flow to each flow leg. This ensures that the minimum required flow is capable of feeding each flow CALVERT CLIFFS - UNITS 1 & 2 B 3.7.3-9 Revision 26
AFW System B 3.7.3 BASES leg. The test may be performed on one flow leg at a time.
The SR is modified by a Note which states, the SR is not required to be performed for the AFW train with the turbine-driven AFW pump until up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 800 psig in the steam generators. The Note ensures that proper test conditions exist prior to performing the test usin the SR 3.7.3.7 This SR ensures that the AFW System is properly aligned by verifying the flow path to each steam generator prior to entering MODE 2 operation, after 30 days in MODEs 5 or 6.
OPERABILITY of AFW flow paths must be verified before sufficient core heat is generated that would require the operation of the AFW System during a subsequent shutdown.
The Frequency is reasonable, based on engineering judgment, and other administrative controls to ensure that flow paths remain OPERABLE. To further ensure AFW System alignment, the OPERABILITY of the flow paths is verified following extended outages to determine that no misalignment of valves has occurred. This SR ensures that the flow path from the CST to the steam generators is properly aligned. Minimum nominal flow to each flow leg is ensured by performance of SR 3.7.3.6.
REFERENCES 1. UFSAR, Section 10.3
- 2. ASME Code for Operation and Maintenance of Nuclear Power Plants B 3.7.3-10 Revision 38 CALVERT CLIFFS - UNITS 1 CLIFFS - UNITS & 2 1 & 2 B 3.7.3-10 Revision 38
CST B 3.7.4 BASES B.1 and B.2 If the CST cannot be restored to OPERABLE status within the associated Completion Time, the affected unit(s) must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit(s) must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. 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 plant systems.
SURVEILLANCE SR 3.7.4.1 REQUIREMENTS This SR verifies that the CST contains the required usable volume of cooling water. (This volume ý 150 000 gallons Although the volume in the CST for each unit is required to be 150,000 gallons, the total combined volume for both units is 300,000 gallons.
REFERENCES 1. UFSAR CALVERT CLIFFS - UNITS 1 & 2 B 3.7.4-4 Revision 41
CC System B 3.7.5 BASES and the low probability of a DBA occurring during this period.
B.1 and B.2 If the CC loop cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
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.
SURVEILLANCE SR 3.7.5.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the CC flow path provides assurance that the proper flow paths exist for CC operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves are verified to be in the correct position prior to locking, sealing, or securing. This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in their correct position.
This SR is modified by a Note indicating that the isolation of the CC components or systems may render those components inoperable but does not affect the OPERABILITY of the CC System.
SR 3.7.5.2 This SR verifies proper automatic operation of the CC valves on an actual or simulated safety injection actuation signal CALVERT CLIFFS - UNITS 1 & 2 B 3.7.5-4 Revision 2
CC System B 3.7.5 BASES (SIAS). The CC System is a normally operating system that cannot be fully actuated as part of routine testing during normal operation. This SR is not required for valves that are locked, sealed, or otherwise secured in the reuired pass th s n ............ pe when pfe foit d at thr r 24 th
~SR 3.7.5.3 3This SR verifies proper automatic operation of the CC pumps on an actual or simulated SIAS. The CC System is a normally operating system that cannot be fully actuated as part of routine testhnnt normal operation.p:
REFEREurNCESlace test were perform.dwith thc 21 m th FrequeIgey. Therezfore, the F1reguryi aeeeptable f-rom a REFERENCES 1. UFSAR CALVERT CLIFFS - UNITS 1 & 2 B 3.7.5-5 Revision 2
SRW System B 3.7.6 BASES achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
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.
SURVEILLANCE SR 3.7.6.1 REQUIREMENTS Verifying the correct alignment for manual, power-operated, and automatic valves in the SRW flow path ensures that the proper flow paths exist for SRW operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since they are verified to be in the correct position prior to locking, sealing, or securing.
This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position.
This SR is modified by a Note indicating that the isolation of the SRW components or systems may render those components inoperable but does not affect the OPERABILITY of the SRW System.
SR 3.7.6.2 This SR verifies proper automatic operation of the SRW System valves on an actual or simulated actuation signal (SIAS or CSAS). The SRW System is a normally operating system that cannot be fully actuated as part of normal testing. This surveillance test is not required for valves that are locked, sealed, or otherwise secured in the re uired posillion under administrative controls. Te su-rvogillanoc test umder the eemd~itins that apply dwring a unit autago, and the potential for an unlnndt..c f rt-h gurveillamee tu-ezt w~e:crc -per-foermed wVvitlh the reaeto-r-a CALVERT CLIFFS - UNITS 1 & 2 B 3.7.6-4 Revision 5
SRW System B 3.7.6 BASES po~cr.Opeatig epcrcnc ha shwn hatthese componcnts UuatmLlj--
ypa-s- the su-,ryeillanee test wben perf redath 2ý4 munthl FIequeIIcy. Th~ei-Fo.-, the Freueeyisacpbl frrff a ro iablt stanpoin SR 3.7.6.3 The SR verifies proper automatic operation of the SRW System pumps on an actual or simulated actuation signal (SIAS or CSAS). The SRW System is a normally operating system that cannot be fully actuated as part of the normal testing dRFrENi n1UFnoSaRrie he REFERENCES 1. UFSAR CALVERT CLIFFS - UNITS 1 & 2 B 3.7.6-5 Revision 5
SW System B 3.7.7 BASES Thb 1 day 3b~ regn igjdmni 4~gec consistefnt with the preeedural eentrals goei'ing valv SR 3.7.7.2 This SR verifies proper automatic operation of the SW Syste m valves on an actual or simulated actuation signal (SIAS).
The SW System is a normally operating system that cannot be fully actuated as part of the normal testing. This surveillance test is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. Tr-4e u m t1 t Lesunerthe eonditionsthtapydrnauitutg and the petentia4 for an unplanned transient if the i
va ves with an Engineered Safety Feature Actuation System
~signal since automatic system reconfiguration during a LOCA
~is not requi red.
~SR 3.7.7.3
~The SR verifies proper automatic operation of the SW System
~pumps on an actual or simulated actuation signal (SIAS).
~The SW System is a normally operating system that cannot be fully actuate spr fte omltsigdrling norrmal REFERENCES 1. UFSAR, Section 9.5.2.3, "Saltwater System" CALVERT CLIFFS - UNITS 1 & 2 B 3.7.7-4 Revision 12
CREVS B 3.7.8 BASES normal operating conditions on this system are not severe, testing each required CREVS filter train once every month provides an adequate check on this system.
This SR verifies that the CREVSrequired testing is performed ~scssdwith infrmtio inaeaccordance i dta the lfinemtheliabiity.
Ventilation Filter Testin'h Program (VFTP). The CREVS filter tests are in accordance with portions of Reference 2. The VFTP includes testing
~SR HEPA3.7.8.2 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 3.7.8.3 This SR verifies each CREVS train starts and operates on ai actual or simulated actuation signal (CRRS) .y h.-~~
." CY
-- U . ... 1I ~L 4 U~l y . I11[ I I Zo I F-j U 1L 1 I 3j adqut to enzurc thc CR-EV-S is eapable of 3+&t~in ~i am aperating em an aetual or sifflulated CRR .
SR 3.7.8.4 This SR verifies the OPERABILITY of the CRE boundary by testing for unfiltered air inleakage past the CRE boundary and into the CRE. The details of the testing are specified in the Control Room Envelope Habitability Program.
The CRE is considered habitable when the radiological dose to the CRE occupants calculated in the licensing basis analysis of DBA consequences is no more than 5 rem TEDE and the CRE occupants are protected from hazardous chemicals and smoke. This SR verifies that the unfiltered air inleakage into the CRE is no greater than the flow rate assumed in the licensing basis analysis of DBA consequences. When unfiltered air inleakage is greater than the assumed flow rate, Condition E must be entered. Options for restoring the CRE boundary to OPERABLE status include changing the CALVERT CLIFFS - UNITS 1 & 2 B 3.7.8-9 Revision 42
CRETS B 3.7.9 BASES B.1 and B.2 If the Required Actions and associated Completion Times of Condition A are not met in MODEs 1, 2, 3, or 4, the unit must be placed in a MODE that minimizes the accident risk.
To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 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.
C.1 If both CRETS trains are inoperable in MODEs 1, 2, 3, or 4, or during movement of irradiated fuel assemblies, the CRETS may not be capable of performing the intended function and the unit is in a condition outside the accident analysis.
Therefore, LCO 3.0.3 must be entered immediately and movement of irradiated fuel must be suspended immediately.
This does not preclude the movement of fuel assemblies to a safe condition.
SURVEILLANCE SR 3.7.9.1 REQUIREMENTS This SR verifies each required CRETS train has the capability to maintain Control Room temperature
- 104 0F for
> 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> in the recirculation mode. During this test, the backup Control Room air conditioner is to be de-energized. This SR consists of a combination of testing./*
thisne the 4cs cp-f r-' r-1 -; and i-iz noitigifexpcctcd Aver 4IT
'LlIN dogr-adatin REFERENCES 1. UFSAR, Section 9.8.2.3, "Auxiliary Building Ventilating Systems" CALVERT CLIFFS - UNITS 1 & 2 B 3.7.9-3 Revision 31
SFPEVS B 3.7.11 BASES SURVEILLANCE SR 3.7.11.1 REQUIREMENTS The SR requires verification A that the SFPEVS is in operation. Verification includes verifying that one exhaust fan is operatin and discharging into the
- ventilation stc. Rej 1- IIu SR3.7.11.2-,e~r-4--R h Deleted.
SR 3.7.11.3 This SR verifies the integrity of the spent fuel storage 3pool area. The ability of the spent fuel storage pool area Ato maintain negative pressure with respect to potentially uncontaminated adjacent areas is periodically tested to verify proper function of the SFPEVS. During operation, the spent fuel storage pool area is designed to maintain a
~slight negative pressure in the spent fuel storage pool area, with respect to adjacent stensure that REFERENCES 1. UFSAR
- 2. Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000 CALVERT CLIFFS - UNITS 1 & 2 B 3.7. 11-3 Revision 41
PREVS B 3.7.12 BASES during this time period, and the consideration that the remaining train can provide the required capability.
B.1 and B.2 If the inoperable train cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. 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.
SURVEILLANCE SR 3.7.12.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly. As the environment and normal operating conditions on this system are not severe, testing each train once every month provides an adequate check on this system.
The test is performed by initiating the system from the Control Room, ensuring flow through the HEPA filter and charcoal adsorber train, and verifyin this system operates SR 3.7.12.2 This SR verifies the performance of PREVS filter testing in accordance with the VFTP. The PREVS filter tests are in accordance with portions of Reference 2. The VFTP includes testing the performance of the HEPA filter, 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.
Revision 41 UNITS 1 CLIFFS - UNITS CALVERT CLIFFS -
1&& 2 2 B 3.7.12-3 B 3.7.12-3 Revision 41
PREVS B 3.7.12 BASES SR 3.7.12.3 This SR verifies that each PREVS train starts and operates on an actual or simulated actuation signal (Containment REFERENCES 1. UFSAR
- 2. Regulatory Guide 1.52, Revision 2, "Design, Testing, and Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants," March 1978
- 3. Regulatory Guide 1.194, Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants, June 2003
- 4. Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000 Revision 41 CALVERT CLIFFS -
& 2 UNITS 11 &
CLIFFS - UNITS 2 B 3.7.12-4 B 3.7.12-4 Revision 41
SFP Water Level B 3.7.13 BASES The SFP water level satisfies 10 CFR 50.36(c)(2)(ii),
Criteria 2 and 3.
LCO The specified water level preserves the assumptions of the fuel handling accident analysis (Reference 1, Section 14.18). As such, it is the minimum required for fuel storage, reconstitution, and movement within the fuel storage pool.
APPLICABILITY This LCO applies during movement of irradiated fuel assemblies in the SFP since the potential for a release of fission products exists.
ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.
When the initial conditions for an accident cannot be met, steps should be taken to preclude the accident from occurring. When the SFP water level is lower than the required level, the movement of irradiated fuel assemblies in the SFP is immediately suspended. This effectively precludes a spent fuel handling accident from occurring.
This does not preclude moving a fuel assembly to a safe position.
If moving irradiated fuel assemblies while in MODEs 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODEs 1, 2, 3, and 4, the fuel movement is independent of reactor operations.
Therefore, in either case, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.
SURVEILLANCE SR 3.7.13.1 REQUIREMENTS This SR verifies sufficient SFP water is available in the event of a fuel handling accident. The water level in the SFP must be checked periodically. T zeendyFrqnc se
,tn
,iie-1 he~ - v-nv' 11" CALVERT CLIFFS - UNITS 1 & 2 B 3.7.13-2 Revision 41
Secondary Specific Activity B 3.7.14 BASES secondary specific activity cannot be restored to within limits in the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 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.
SURVEILLANCE SR 3.7.14.1 REQUIREMENTS This SR ensures that the secondary specific activity is within the limits of the accident analysis. A gamma isotope analysis of the secondary coolant, which determines DOSE EQUIVALENT 1-131, confirms the validity of the safety analysis assumptions as to the source terms in post-accident releases. It also serves to identify and trend any unusual isotopic concentrations that might ind* ate chn e n REFERENCES 1. UFSAR, Chapter 14, "Safety Analysis"
- 2. Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000 2 B 3.7.14-3 Revision 41 CALVERT CLIFFS UNITS 11 &
CLIFFS - UNITS
& 2 B 3.7.14-3 Revision 41
SFP Boron Concentration B 3.7.16 BASES ACTIONS A.1 and A.2 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply. If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation. Therefore, inability to suspend movement of fuel assemblies is not a sufficient reason to require a reactor shutdown.
When the concentration of boron in the SFPs is less than required, immediate action must be taken to preclude an accident from happening or to mitigate the consequences of an accident in progress. This is most efficiently achieved by immediately suspending the movement of fuel assemblies.
This does not preclude the movement of fuel assemblies to a safe position. In addition, action must be immediately initiated to restore boron concentration to within limits.
SURVEILLANCE SR 3.7.16.1 REQUIREMENTS This SR verifies that the concentration of boron in the SFPs is within the required limit. As long as this SR is met,
- 4) the analyzed incidents are full addressed. The -daY pool water is expeeted to take ver a shert pcriod of elc REFERENCES None CALVERT CLIFFS - UNITS I & 2 B 3.7.16-2 Revision 23 1
AC Sources-Operating B 3.8.1 BASES SR 3.8.1.1 and SR 3.8.1.2 These SRs assure proper circuit continuity for the offsite AC electrical power supply to the onsite distribution network and availability of offsite AC electrical power.
The breaker alignment verifies that each breaker is in its correct position to ensure that distribution buses and loads are connected to their preferred power source, and that appropriate independence of offsite circuits is maintained.
The Frequency of once within one hour after substitution for a 500 kV circuit and thereafter, for SR 3.8.1.1 was established to ensure that the breaker alignment for the SMECO circuit (which does not have Control Room indication) is in its correct position alth ugh breaker
//euit breake-r Surveillance Requirement 3.8.1.1 is modified by a Note which states that this SR is only required when SMECO is being credited for an offsite source. This SR will prevent unnecessary testing on an uncredited circuit.
SR 3.8.1.3 and SR 3.8.1.9 These SRs help to ensure the availability of the standby electrical power supply to mitigate DBAs and transients and to maintain the unit in a safe shutdown condition.
To minimize the wear on moving parts that do not get lubricated when the engine is not running, these SRs are modified by a Note (Note 2 for SR 3.8.1.3) to indicate that all DG starts for these surveillance tests may be preceded by an engine prelube period and followed by a warmup period prior to loading by an engine prelube period.
For the purposes of SR 3.8.1.9 testing, the DGs are required to start frbm standby conditions only for SR 3.8.1.9.
Standby conditions for a DG mean the diesel engine coolant and oil are being continuously circulated and temperature is being maintained consistent with manufacturer recommendations.
Revision 26 UNITS 1 CLIFFS - UNITS CALVERT CLIFFS -
& 2 1 & 2 B 3.8.1-24 B 3.8. 1-24 Revision 26
AC Sources-Operating B 3.8.1 BASES In order to reduce stress and mechanical wear on diesel engines, the DG manufacturers recommend a modified start in which the starting speed of DGs is limited, warmup is limited to this lower speed, and the DGs are gradually accelerated to synchronous speed prior to loading. This is the intent of Note 3, which is only applicable when such modified start procedures are recommended by the manufacturer.
Survejjjlnce Requirement 3.8.1.9 requires thattý
' E the DG starts from standby conditions and achieves required voltage and frequency within 10 seconds.
The minimum voltage and frequency stated in the SR are those necessary to ensure the DG can accept DBA loading while maintaining acceptable voltage and frequency levels. The 10 second start requirement supports the assumptions of the design basis loss of coolant accident analysis in Reference 2, Chapter 14.
Since SR 3.8.1.9 requires a 10 second start, it is more restrictive than SR 3.8.1.3, and it may be performed in lieu of SR 3.8.1.3.
SR 3.8.1.4 This SR verifies that the DGs are capable of synchronizing with the offsite electrical system and accepting loads greater than or equal to 4000 kW for No. 1A DG and greater than or equal to 90% of the continuous duty rating for the remaining DGs. The 90% minimum load limit is consistent with Reference 3 and is acceptable because testing of these DGs at post-accident load values is performed by SR 3.8.1.11. A minimum run time of 60 minutes is required to stabilize engine temperatures, while minimizing the time that the DG is connected to the offsite source.
CALVERT CLIFFS - UNITS 1 & 2 B 3.8. 1-25 Revision 26
AC Sources-Operating B 3.8.1 BASES Although no power factor requirements are established by this SR, the DG is normally operated at a power factor between 0.8 lagging and 1.0. The 0.8 value is the design rating of the machine, while 1.0 is an operational This SR is modified by four Notes. Note 1 indicates that the diesel engine runs for this surveillance test may include gradual loading, as recommended by the manufacturer, so that mechanical stress and wear on the diesel engine are minimized. Note 2 states that momentary transients because of changing bus loads do not invalidate this test. Note 3 indicates that this surveillance test shall be conducted on only one DG at a time in order to prevent routinely paralleling multiple DGs and to minimize the potential for effects from offsite circuit or grid perturbations. Note 4 stipulates a prerequisite requirement for performance of this SR. A successful DG start must precede this test to credit satisfactory performance.
SR 3.8.1.5 This SR provides verification that the level of fuel oil in the day tank is at or above the level at which fuel oil is automatically added. The level required by the SR is expressed as an equivalent volume in gallons, and is selected to ensure adequate fuel oil for a minimum of one hour of DG operation at full load plus 10%.
SR 3.8.1.6 Microbiological fouling is a major cause of fuel oil degradation. There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a water environment in order to survive. Removal of water from the fuel oil day tanks -eliminates the necessary environment for bacterial survival. This is the most effective means of controlling microbiological fouling.
CALVERT CLIFFS - UNITS 1 & 2 B 3.8. 1-26 Revision 26
AC Sources-Operating B 3.8.1 BASES In addition, it eliminates the potential for water entrainment in the fuel oil during DG operation. Water may come from any of several sources, including condensation, ground water, rain water, contaminated fuel oil, and from breakdown of the fuel oil by bacteria. Frequent checking for and removal of accumulated water minimizes fouling and provides data regardin the waterti ht integrit of the fuel PP P Maipte-A&A- The presence of water does not necessarily represent failure of this SR provided the accumulated water is removed during the
/ performance of this surveillance test.
SR 3.8.1.7 This SR demonstrates that one fuel oil transfer pump 73
- operates and transfers fuel oil from its associated storage tank to its associated day tank. This is required to support continuous operation of standby power sources. This SR provides assurance that the fuel oil transfer pump is OPERABLE, the fuel oil piping system is intact, the fuel delivery piping is not obstructed, and the controls and control systems for automatic fuel transfer systems are OPERABLE.
TheF cy for th-i R 31de day Fruny arrespeqnd to thc de4nnt of the fuel transfer lyoaemd.The sdeqiun o fel transRfver systems 1-such 5). Thi sperate autmatlically conrolus hbeptartmd anually inordar to mabintain a padrev verloi goff tel D n theday taohg dmriogor fsllvin c g e. thestin. Inl su eacae, a 31 dat SR 3.8.1.8 Under accident and loss of offsite power conditions loads are sequentially connected to the bus by the automatic load sequencer (this SR verifies steps 1 through 5). The sequencing logic controls the permissive and closing signals to breakers to prevent overloading of the DGs due to high motor starting currents. The 10% load sequence time interval tolerance ensures that sufficient time exists for the DG to restore frequency and voltage prior to applying the next load, and that safety analysis assumptions CALVERT CLIFFS - UNITS 1 & 2 B 3.8. 1-27 Revision 26
AC Sources-Operating B 3.8.1 BASES regarding ESF equipment time delays are not violated. The UFSAR provides a summary of the automatic loading of ESF buses.
/Tho-P4-eeny of 31 days is consistent with BGmntl aid ; suficict to~ ens~ure the load sequenrce LZ~ti~
SR 3.8.1.9 See SR 3.8.1.3.
SR 3.8.1.10 Transfer of each 4.16 kV ESF bus power supply from the normal offsite circuit to the alternate offsite circuit tat94aeres judgmnt, o tkign conditin" datsosibte, ite odemosratoensueta the OPRAILT of thesaternateoa circuitin Thst SR faeperfomied ab4iliy the-frq~
DC~sverfication to erormthir h saeyfnto.
reatthe e ane opreatedI at calload eate ta accident redicted loades ftor at .ea4s 60h inues p wer Operation athe greuater ohsRdertoviensurerthatathDGisn teste under lGcneoadraconditin 60miustbepromduigaO Oadereater a than oreqaltor thncalculated tat areasler accidentt loads anwuing desig accidentios clarlpowermfactorate0.84 adspossibe ltsti thanescalcuatued acciDen loads wuill cleal demonstrted theig for No. 1A DG and *ý0.83 for Nos. 1B, 2A, and 2B DGs.
These power factors are chosen based on the calculated highest kW value of DG loads during the postulated design basis accidents.
In addition, the post-accident load for No. 1A DG is significantly lower than the continuous rating of No. 1A DG.
B 3.8.1-28 Revision 46 CALVERT - UNITS 1 CLIFFS - UNITS CALVERT CLIFFS & 2 1 & 2 B 3.8. 1-28 Revision 46
AC Sources-Operating B 3.8.1 BASES To ensure No. 1A DG performance is not degraded, routine monitoring of engine parameters should be performed during the performance of this SR for No. IA DG (Reference 9).
This SR is modified by two Notes, Note 1 states that momentary transients due to changing bus loads do not invalidate this test. Similarly, momentary power factor transients above the limit will not invalidate the test.
Note 2 ensures that the DG is tested under load conditions that are as close to design basis conditions as practicable.
When synchronized with offsite power, testing should be performed at a power factor of
- 0.84 for No. 1A DG and
< 0.83 for Nos. IB, 2A, and 2B DGs. These power factors are representative of the actual inductive loading a DG would see under design basis accident conditions. Under certain conditions, however, Note 2 allows the surveillance to be conducted at a power factor other than
- 0.84 for No.
1A DG and
- 0.83 for Nos. lB, 2A, and 2B DGs. These conditions occur when grid voltage is high, and the additional field excitation needed to get the power factor to
- 0.84 for No. 1A DG and
- 0.83 for Nos. IB, 2A, and 2B DGs results in voltages on the emergency busses that are too high. Conditions can also occur that could result in emergency bus voltages which are too low. Under these conditions, the power factor shall be maintained as close a practicable to 0.84 for No. 1A DG and 0.83 for Nos. IB, 2A, and 2B DGs while maintaining acceptable voltages on the emergency busses.
SR 3.8.1.12 Each DG is provided with an engine overspeed trip to prevent damage to the engine. Recovery from the transient caused by the loss of a large load could cause diesel engine overspeed, which, if excessive, might result in a trip of the engine. This SR demonstrates the DG load response characteristics. This SR is accomplished by tripping the DG output breaker with the DG carrying greater than or equal to CALVERT CLIFFS - UNITS 1 & 2 B 3.8.1-29 Revision 46
AC Sources-Operating B 3.8.1 BASES its associated single largest post-accident load while paralleled to offsite power.
Consistent with References 10, 3, and 4, the load rejection test is acceptable if the increase in diesel speed does not exceed 75% of the difference between synchronous speed and the overspeed trip setpoint, or 15% above synchronous speed, whichever is lower.
SR 3.8.1.13 This SR demonstrates that DG non-critical protective functions are bypassed on a required actuation signal. This SR is accomplished by verifying the bypass contact changes to the correct state which prevents actuation of the non-critical function. The non-critical protective functions are consistent with References 3 and 4, and Institute of Electrical and Electronic Engineers (IEEE)-387 and are listed in Reference 2, Chapter 8. Verifying the non-critical trips are bypassed will ensure DG operation during a required actuation. The non-critical trips are bypassed during DBAs and provide an alarm on an abnormal engine condition. A failure of the electronic governor results in the diesel generator operating in hydraulic mode. This alarm provides the operator with sufficient time to react appropriately. The DG availability to mitigate the DBA is more critical than protecting the engine against minor problems that are not immediately detrimental to emergency operation of the DG.
experience haszcw th_4a4+t thQzc eEompoAnen~ctsicllv pass thn SR whem p..rformed at- th...4 lLotII Freqt Tlerczr, h
..~cl~edw1~o b~accptole roma rliailiy Fr~geIIy
,tda~dputfL.tL3 r~ 1 y; U~tr ihRfrcc2 CALVERT CLIFFS - UNITS 1 & 2 B 3.8. 1-30 Revision 46
AC Sources-Operating B 3.8.1 BASES SR 3.8.1.14 This SR ensures that the manual synchronization and load transfer from the DG to the offsite source can be made and that the DG can be returned to ready-to-load status when offsite power is restored. The DG is considered to be in ready-to-load status when the DG is at rated speed and voltage, the output breaker is open and can receive an auto-close signal on bus undervoltage, and the load sequence timers are reset.
SR 3.8.1.15 In the event of a DBA coincident with a loss of offsite power, the DGs are required to supply the necessary power to ESF systems so that the fuel, RCS, and containment design limits are not exceeded.
This SR demonstrates the DG operation during a loss of offsite power actuation test signal in conjunction with an ESF (i.e., safety injection) actuation signal. In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the DG system to perform these functions is acceptable. This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.
It is not necessary to energize loads which are dependent on temperature to load (i.e., heat tracing, switchgear HVAC compressor, computer room HVAC compressor). Also, it is acceptable to transfer the instrument AC bus to the non tested train to maintain safe operation of the plant during testing. Loads (both permanent and auto connect) < 15 kW do not require loading onto the diesel since these are insignificant loads for the DG.
Revision 46 UNITS 1 CLIFFS - UNITS CALVERT CLIFFS -
& 2 1 & 2 B 3.8.1-31 B 3.8. 1-31 Revision 46
AC Sources-Operating B 3.8.1 BASES Permanently- and auto-connected loads to the emergency diesel generators are defined as follows:
Permanently-Connected Load - Equipment that is not shed by an undervoltage or safety injection actuation signal and is normally operating, i.e., loads that are manually started, selected, or process signal controlled are not considered permanently-connected loads.
Auto-Connected Loads - Emergency equipment required for mitigating the events described in UFSAR Chapter 14 that are energized by loss-of-coolant incident sequencer actions after step zero and within the first minute of emergency diesel generator operation after the initiation of an undervoltage signal.
This SR is modified by a Note. The reason for the Note is to minimize mechanical wear and stress on the DGs during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant and oil continuously circulated and temperature maintained consistent with manufacturer recommendations for DGs.
SR 3.8.1.16 This SR lists the SRs that are applicable to the LCO 3.8.1.c (SRs 3.8.1.1, 3.8.1.2, 3.8.1.3, 3.8.1.5, 3.8.1.6, and 3.8.1.7). Performance of any SR for the LCO 3.8.1.c will satisfy both Unit 1 and Unit 2 requirements for those SRs.
Surveillance Requirements 3.8.1.4, 3.8.1.8, 3.8.1.9, 3.8.1.10, 3.8.1.11, 3.8.1.12, 3.8.1.13, 3.8.1.14, and 3.8.1.15, are not required to be performed for the LCO 3.8.1.c. Surveillance Requirement 3.8.1.10 is not required because this SR verifies manual transfer of AC power sources from the normal offsite circuit to the alternate offsite circuit, but only one qualified offsite circuit is necessary for the LCO 3.8.1.c. Surveillance Requirements 3.8.1.4, 3.8.1.11, and 3.1.8.12 are not CALVERT CLIFFS - UNITS 1 & 2 B 3.8.1-32 Revision 46
Diesel Fuel Oil B 3.8.3 BASES F.1 With a Required Action and associated Completion Time not met, or one or more DGs with diesel fuel oil not within limits for reasons other than addressed by Conditions A through E, the associated DG may be incapable of performing its intended function and must be immediately declared inoperable. "Associated DG(s)" are identified in the Applicability Bases.
SURVEILLANCE SR 3.8.3.1 REQUIREMENTS This SR provides verification that there is an adequate inventory of fuel oil in the DG FOSTs to support one unit on accident loads and one unit on shutdown loads for seven days. The seven day period is sufficient time to place the unit in a safe shutdown condition and to bring in replenishment fuel from an offsite location.
I The day Fr,,*-,,~ i*s* adqaetoesr
- hat a swf- +
SR 3.8.3.2 The tests listed below are a means of determining whether new fuel oil is of the appropriate grade (i.e., 2D and 2D low sulfur) and has not been contaminated with substances that would have an immediate, detrimental impact on diesel engine combustion. Note that further references to American Society for Testing Materials (ASTM) 2D fuel oil include both 2D and 2D low sulfur. If results from these tests are within acceptable limits, the fuel oil may be added to the storage tanks without concern for contaminating the entire volume of fuel oil in the storage tanks. These tests are to be conducted prior to adding the new fuel to the storage tank(s), but in no case is the time between receipt of new fuel and conducting the tests to exceed 31 days. The tests, limits, and applicable ASTM Standards are as follows:
- a. Sample the new fuel oil in accordance with Reference 3, ASTM D4057-1995; CALVERT CLIFFS - UNITS 1 & 2 B 3.8.3-7 Revision 2
Diesel Fuel Oil B 3.8.3 BASES The Frequency of this test takes into consideration fuel oil degradation trends that indicate that particulate concentration is unlikely to change significantly between Frequency intervals.
SR 3.8.3.3 Microbiological fouling is a major cause of fuel oil degradation. There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a water environment in order to survive. Removal of water from the fuel storage tanks - eliminates the necessary environment for bacterial survival. This is the most effective means of controlling microbiological fouling.
In addition, it eliminates the potential for water entrainment in the fuel oil during DG operation. Water may come from any of several sources, including condensation, ground water, rain water, and contaminated fuel oil, and from breakdown of the fuel oil by bacteria. Frequent checking for and removal of accumulated water minimizes fouling and provides data regarding the watertight integrity T hepresence o water does not necessarily represent ailure of this SR provided the accumulated water is removed during performance of the surveillance test.
REFERENCES 1. UFSAR
- 2. ASTM Standards
- 3. Regulatory Guide 1.137, "Fuel-Oil Systems for Standby Diesel Generators," January 1978 CALVERT CLIFFS - UNITS 1 & 2 B 3.8.3-9 Revision 2
DC Sources-Operating B 3.8.4 BASES Visual inspection to detect corrosion of the battery cells and connections, or measurement of the resistance of each cell to cell and terminal connection, provides an indication of physical damage or abnormal deterioration that could potentially degrade battery performance.
The limits established for this SR must be no more than 20%
above the resistance as measured during installation or not above the ceiling value established by the manufacturer.
- r-'x ~Visual inspection of the battery cells, cell plates, and battery racks provides an indication of physical damage or abnormal deterioration that could potentially degrade battery performance.
The presence of physical damage or deterioration does not necessarily represent a failure of this SR, provided an
~evaluation determines that the physical damage or
~deterioration does not affect the OPERABILITY of the batsther (is abiltyt performe its18mhfun'Jctin.
dhesg Oporatinj experience bag shown t-hat these eomponents, usually Therefore,. the r .......... c... m li..e- to I-- - -
SR 3.8.4.4 and SR 3.8.4.5 Visual inspection and resistance measurements of cell to cell and terminal connections provide an indication of CALVERT CLIFFS - UNITS 1 & 2 B 3.8.4-5 Revision 45
DC Sources-Operating B 3.8.4 BASES physical damage or abnormal deterioration that could indicate degraded battery condition. The anti-corrosion material is used to help ensure good electrical connections and to reduce terminal deterioration. The visual inspection for corrosion is not intended to require removal of and inspection under each terminal connection. The removal of visible corrosion is a preventive maintenance SR. The presence of visible corrosion does not necessarily represent a failure of this SR provided visible corrosion is removed during performance of SR 3.8.4.4.
The connection resistance limits for SR 3.8.4.5 shall be no more than 20% above the resistance as measured during installation, or not above the ceiling value established by the manufacturer.
The 18 month Fr nyfr these SR9 is based zirei§ judgment. Ope-ratinA9xperi emee has shw tha thege comp@ncnts usually pass the SR9 witen perfurmcd at the 18month Fr-equency'. Tiherle forc, the Frzguccypr conluJ ttLe aceeptable fromii etreliabili;ty stanpoint~.
SR 3.8.4.6 This SR requires that each battery charger be capable of supplying 400 amps and 125 V for Ž 30 minutes. These requirements are based on the output rating of the chargers (Reference 1, Chapter 8). According to Reference 7, the battery charger supply is required to be based on the largest combined demands of the various steady state loads and the charging capacity to restore the battery from the design minimum charge state to the fully charged state, irrespective of the status of the unit during these demand occurrences. The minimum required amperes and duration cqisprfomdto il superfgorma nd~ DC~othradmsnioratin l~~
c~ontrols existing to oni"ue adequatee eharger- perfer-manee F1 CgUCIIy 'r9 itended to be eansistent with expeeted fuel eyl in CALVERT CLIFFS - UNITS 1 & 2 B 3.8.4-6 Revision 45
DC Sources-Operating B 3.8.4 BASES SR 3.8.4.7 A battery service test is a special test of battery capability, as found and with the associated battery charger disconnected, to satisfy the design requirements (battery duty cycle) of the DC source. The test duration must be 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and battery terminal voltage must be maintained 105 volts during the test. The discharge rate and test length should correspond to the design accident load (duty) cycle requirements as specified in Reference 1, Chapter 8.
A dummy load simulating the emergency loads of the design duty cycle may be used in lieu of the actual emergency loads.
This SR is modified by a Note. The Note allows the performance of a modified performance discharge test in lieu of a service test. This substitution is acceptable because a modified performance discharge test represents a more severe test of battery capacity than SR 3.8.4.7.
SR 3.8.4.8 A battery performance discharge test is a test of constant current capacity of a battery after having been in service, to detect any change in the capacity determined by the acceptance test. The test is intended to determine overall battery degradation due to age and usage.
A battery modified performance discharge test is a simulated duty cycle consisting of just two rates; the one minute rate published for the battery or the largest current load of the duty cycle, followed by the test rate employed for the performance discharge test, both of which envelope the duty cycle of the service test. Since the ampere-hours removed by a rated one minute discharge represents a very small portion of the battery capacity, the test rate can be changed to that for the performance test without compromising the results of the performance discharge test.
The battery terminal voltage for the modified performance discharge test should remain above the minimum battery CALVERT CLIFFS - UNITS 1 & 2 B 3.8.4-7 Revision 45
DC Sources-Operating B 3.8.4 BASES terminal voltage specified in the battery performance discharge test for the duration of time equal to that of the performance discharge test.
A modified performance discharge test is a test of the battery capacity and its ability to provide a high rate, short duration load (usually the highest rate of the duty cycle). This will often confirm the battery's ability to meet the critical period of the load duty cycle, in addition to determining its percentage of rated capacity. Initial conditions for the modified performance discharge test should be identical to those specified for a service test.
Either the battery performance discharge test or the modified performance discharge test is acceptable for satisfying SR 3.8.4.8; however, only the modified performance discharge test may be used to satisfy SR 3.8.4.8 while satisfying the requirements of SR 3.8.4.7 at the same time.
The acceptance criteria for this SR are consistent with References 6 and 4. These references recommend that the battery be replaced if its capacity is below 80% of the manufacturer rating. A capacity of 80% shows that the battery rate of deterioration is increasing, even if there is ample capacity to meet the load requirements.
If a er s ows egrada Ion, or t e battery has reached 85% of its expected life and capacity is < 100% of the manufacturer's rating, the SR Frequency is reduced to 12 months. However, if the battery shows no degradation but has reached 85% of its expected life, the SR Frequency is only reduced to 24 months for batteries that retain capacity
> 100% of the manufacturer's rating. Degradation is indicated, according to Reference 6, when the battery capacity drops by more than 10% relative to its capacity on the previous performance test or when it is Ž 10% below the manufacturer's rating. These Frequencies are consistent with the recommendations in Reference 6.
CALVERT CLIFFS - UNITS 1 & 2 B 3.8.4-8 Revision 45
Battery Cell Parameters B 3.8.6 BASES Continued operation prior to declaring the affected batteries inoperable is permitted for 31 days before battery cell parameters must be restored to within Category A and B limits. With the consideration that, while battery capacity is degraded, sufficient capacity exists to perform the intended function and to allow time to fully restore the battery cell parameters to normal limits, this time is acceptable prior to declaring the battery inoperable.
B.1 With one or more batteries with one or more battery cell parameters outside the Category C limit for any connected cell, sufficient capacity to supply the maximum expected load requirement is not assured and the corresponding DC channel must be declared inoperable. Additionally, other potentially extreme conditions, such as any Required Action of Condition A and associated Completion Time not met, or average electrolyte temperature of representative cells
< 691F, are also cause for immediately declaring the associated DC channel inoperable.
SURVEILLANCE SR 3.8.6.1 REQUIREMENTS This SR verifies that Category A battery cell parameters are consistent with Reference 2 which recommends regular battery inspections ý_'_ - 0_"_r- men" including voltage, specific gravity, and e ectro Fyt* emperature of pilot cells.
SR3.8.6.2 The i inspection of specific gravity and voltage is consistent with Reference 2.
SR 3.8.6.3 This Surveillance verification that the average temperature of representative cells is > 69 0 F is consistent with a recommendation of Reference 2, which states that the temperature of electrolytes in representative cells should be determined* The temperature is also high enough to supply the required capacity.
B 3.8.6-3 Revision 2 CALVERT CLIFFS -
UNITS 1 CLIFFS - UNITS & 2 1 & 2 B 3.8.6-3 Revision 2
Inverters-Operating B 3.8.7 BASES from its 120 VAC bus powered by an ESF motor control center through a regulating transformer.
Required Action A.1 is modified by a Note, which states to enter the applicable conditions and Required Actions of LCO 3.8.9, when Condition A is entered with one AC vital bus de-energized. This ensures the vital bus is re-energized within two hours.
Required Action A.1 allows 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to fix the inoperable inverter and return it to service. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> limit is based upon engineering judgment, taking into consideration the time required to repair an inverter and the additional risk to which the unit is exposed because of the inverter inoperability. This has to be balanced against the risk of an immediate shutdown, along with the potential challenges to safety systems such a shutdown might entail. When the AC vital bus is powered from its constant voltage source, it is relying upon interruptible AC electrical power sources (offsite and onsite). The uninterruptible inverter source to the AC vital buses is the preferred source for powering instrumentation trip setpoint devices.
B.1 and B.2 If the inoperable devices or components cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within six hours and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 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.
SURVEILLANCE SR 3.8.7.1 REQUIREMENTS This SR verifies that the inverters are functioning properly with all required circuit breakers closed and AC vital buses energized from the inverter. The verification of proper voltage output ensures that the required power is readily available for the instrumentation of the RPS and ESFAS connected to the AC vital buses.
CALVERT CLIFFS - UNITS 1 & 2 B 3.8.7-3 Revision 2
Inverters-Operati ng B 3.8.7 BASES RtRa UoFtrA blE 4E the Room tht REFERENCES 1. UFSAR CALVERT CLIFFS - UNITS 1 & 2 B 3.8.7-4 Revision 2
Inverters-Shutdown B 3.8.8 BASES voltage output ensures that the required power is readily available for the instrumentation connected to the AC vital REFERENCES 1. UFSAR Revision 19 CLIFFS - UNITS CALVERT CLIFFS -
UNITS 11&& 2 2 B 3.8.8-4 B 3.8.8-4 Revision 19
Distribution Systems-Operating B 3.8.9 BASES loads connected to these buses. The sevenday Frequenc and other irndications,3Vdiulable inte Cuntolroom th REFERENCES 1. UFSAR
- 2. Regulatory Guide 1.93, "Availability of Electric Power Sources," December 1974 Revision 2 CLIFFS - UNITS CALVERT CLIFFS -
UNITS 11&& 22 B 3.8.9-9 B 3.8.9-9 Revision 2
Distribution Systems-Shutdown B 3.8.10 BASES Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition.
These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required AC and DC electrical power distribution subsystems and to continue this action until restoration is accomplished in order to provide the necessary power to the unit safety systems.
Notwithstanding performance of the above conservative Required Actions, a required shutdown cooling (SDC) subsystem may be inoperable. In this case, Required Actions A.2.1 through A.2.3 do not adequately address the concerns relating to coolant circulation and heat removal.
Pursuant to LCO 3.0.6, the SDC ACTIONS would not be entered.
Therefore, Required Action A.2.4 is provided to direct declaring SDC inoperable, which results in taking the appropriate SDC actions. The SDC subsystem(s) declared inoperable and not in operation as a result of not meeting this LCO, may be used if needed. However, the appropriate actions are still required to be taken.
The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required distribution subsystems should be completed as quickly as possible in order to minimize the time the unit safety systems may be without power.
SURVEILLANCE SR 3.8.10.1 REQUIREMENTS This SR verifies that the AC, DC, and AC vital bus Electrical Power Distribution System is functioning properly, with all the buses energized. The verification of proper voltage availability on the buses ensures that the required power is readily available for motive as well as control functions for critical s stem loads connected to these buses.th
+uz~czabilit t 1 ....
Ftc triepal -ler din trhu ot B 3.8.10-4 Revision 38 CALVERT CLIFFS - UNITS 1 CLIFFS - UNITS 1&& 2 2 B 3.8. 10-4 Revision 38
Boron Concentration B 3.9.1 BASES depends on the amount of boron that must be injected to reach the required concentration.
SURVEILLANCE SR 3.9.1.1 REQUIREMENTS This Surveillance Requirement (SR) ensures the coolant boron concentration in the RCS and the refueling pool is within the COLR limits. The coolant boron concentration in each volume is determined periodically by chemical analysis.
4EminiUmt Fma S AsqueUcyo7ARf
~ 3 oasonabl- amount of time to ;criaify the berem comeentration4 of--represent-ative s-am~plzs. Thc Frzguoemey is based en 3praig xpren, WhiChaEhcn7 heburs to be REFERENCES 1. Updated Final Safety Analysis Report (UFSAR)
CALVERT CLIFFS - UNITS 1 & 2 B 3.9. 1-4 Revision 10
Nuclear Instrumentation B 3.9.2 BASES The Completion Time of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient to obtain and analyze a reactor coolant sample for boron concentration and ensures that unplanned changes in boron concentration would be identified. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is reasonable, considering the low probability of a change in core reactivity during this period.
SURVEILLANCE SR 3.9.2.1 REQUIREMENTS Surveillance Requirement 3.9.2.1 is the performance of a CHANNEL CHECK, which is a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that the two indication channels should be consistent with core conditions. Changes in fuel loading and core geometry can result in significant differences between source range channels, but each channel should be consistent with its local conditions.
ThrvelquncRey ef 12 hur i niste pcfe itht similacly for the sam ae oANNL inFrumcnt CHECK Frequec SR 3.9.2.2 Surveillance Requirement_3.9.2.2 is the performance of a CHANNEL CALIBRATION r. This SR is modified by a Note stating that neutron etctors are excluded from the CHANNEL CALIBRATION. This is because generating a meaningful test signal is difficult; the detectors are of simple construction, and any failures in the detectors will be apparent as a change in channel output. rq-REFERENCES 1. UFSAR CALVERT CLIFFS - UNITS 1 & 2 B 3.9.2-3 Revision 19
Containment Penetrations B 3.9.3 BASES containment atmosphere to the outside atmosphere through a filtered or unfiltered pathway not in the required status, (including the Containment Purge and Exhaust Isolation System not capable of automatic actuation when the purge and exhaust valves are open) the unit must be placed in a condition in which the isolation function is not needed.
This is accomplished by immediately suspending movement of irradiated fuel assemblies within the Containment Structure.
Performance of these actions shall not preclude completion of movement of a component to a safe position.
SURVEILLANCE SR 3.9.3.1 REQUIREMENTS This SR demonstrates that each of the containment penetrations required to be in its closed position, is in that position. The surveillance test on the open purge and exhaust valves will demonstrate that the valves are not blocked from closing. Also, the surveillance test will demonstrate that each purge and exhaust valve operator has motive power, which will ensure each valve is capable of being closed by an OPERABLE automatic Containment Purge Valve Isolation System.
The survei!Ia~ee-4est is performed aur cvndy54iR emt of irradiated f..el asse-bl^is
..... the Ge...ai...t Strueture. The 5utvs . . .!c e titt-r (e
r~1e.ed te e--eefflmem 9tirate -with the no~rntal duratien of time.
La thref mplete fuel handling. epeaticncs. A survei'llanace test-the.. 3tf.t . e-Fue;im ope, a
.F. tw +w1o )r t ^ ^.voif1ti_
m.e . . ...- d -
. . . .-,-'-- Lhe p* i::i 1 tit ulbh
........... ,_1d . . ...
.';. flrthi LCO. R. 3uwL, tLTTT SR Lensures tzat a pu*Lulatl: f hR4i.... .. e.d..t, that. reeasees fi.ssion.prd*t..
..a...ti.ty witHn the euILMlrerent Struclule, aiil met esult in-- a release f fissiin product rFad . "etivity to %ho environment in r nf thnm described in Reference 1 SR 3.9.3.2 This SR demonstrates that each containment purge and exhaust valve actuates to its isolation position on manual initiation or on an actual or simulated high radiation CALER CLIFFS - 2o UNISei~ Bg393- 4A9 Revisiont 41 CALVERT CLIFFS - UNITS 1 & 2 B 3.9.3-6 Revision 41
Containment Penetrations B 3.9.3 BASES r~girccnt. Hwzvrin order to einzurz the SR. Frequcncýy on per
.. refiieling
- Utagll*pI at -Ms. .
stovement f irradiated fuel assembli;es within Ce 1 1tainmentr-4i LCO 3. 3. , tI - -6etiniinieiiT Rdit a CHANNEL CH'ECK eyerY, 12 hatir and a CHANNEL FUTNCTI~I~ TEST_1 WeAry 92 days tc e I LII LIal l OEA . ITI dunin.
..... l..ig upetonl.t Every 24 months d CMANNEL CALIBRATION p 8- -rformed. The system ac....tia. respcn timei
...... vthr 24rtII ta,2 i n ref tie! img on,, STAGGERE TE.ST BASIS. S-uS*" e.c iIlela R liI i it 36r.3
_3.I6i.ii.j dcmoezrt.ra Pr=
t* tc*
tJ~t thc iso!atio, timz of each valve isin accorda.cz *4ith urvei ance ests per orme during DE will ensure that the valves are capable of closing after a postulated fuel handling accident to limit a release of fission product radioactivity from the Containment Structure.
REFERENCES 1. UFSAR, Section 14.18, "Fuel Handling Incident" CALVERT CLIFFS - UNITS 1 & 2 B 3.9.3-7 Revision 41
SDC and Coolant Circulation-High Water Level B 3.9.4 BASES With SDC loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere. Performing the actions described above ensure that all containment penetrations are either closed or can be closed so that the dose limits are not exceeded.
The Completion Time of four hours allows fixing of most SDC problems and is reasonable, based on the low probability of the coolant boiling in that time.
The emergency air lock temporary closure device cannot be credited for containment closure for a loss of shutdown cooling event. At least one door in the emergency air lock must be closed to satisfy this action statement.
SURVEILLANCE SR 3.9.4.1 REQUIREMENTS This SR demonstrates that the SDC loop is in operation and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability, and to prevent thermal and boron stratifcation in the core.
Q +r t perator, pm tontroel
- nd alarm inhdiratinnq avail~ble to the operator in h REFERENCES 1. UFSAR, Section 9.2, "Shutdown Cooling System" Revision 43 UNITS 1 CLIFFS - UNITS CALVERT CLIFFS -
1 && 2 2 B 3.9.4-5 B 3.9.4-5 Revision 43
SDC and Coolant Circulation-Low Water Level B 3.9.5 BASES The Completion Time of four hours allows fixing of most SDC problems and is reasonable, based on the low probability of the coolant boiling in that time.
The emergency air lock temporary closure device cannot be credited for containment closure for a loss of shutdown cooling event. At least one door in the emergency air lock must be closed to satisfy this action statement.
SURVEILLANCE SR 3.9.5.1 REQUIREMENTS This SR demonstrates that one SDC loop is operating and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core. This SR also demonstrates that the other SDC loop is OPERABLE.
In addition, during operation of the SDC loop with the water level in the vicinity of the reactor vessel nozzles, the SDC loop flow rate determination must also consider the SDC pump sucrfation rqientha sh.eurdlo: r PRBEadi opcrtuie~srastha 1cps ean bpacdin eperation a-S ncccdto aint-ain d~ecay heat and retain forccd cirre ul;Partion The Frequency-of hours is considered ea-senable, 9inee ether adffministr-ativc controls, are avil pH and have pre e to be aeeepta-bl y prai This SR demonstrates that the SOC loop is in operation and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron strati- icat-im in the core.
and lar indcatons vaiableforthc pertoor in th Ccntrol Roomr fer men iterin h OCS4c CALVERT CLIFFS - UNITS 1 & 2 B 3.9.5-5 Revision 41
SDC and Coolant Circulation-Low Water Level B 3.9.5 BASES SR 3.9.5.3 Verification that the required pump and valves are OPERABLE ensures that an additional SDC loop can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to REFERENCES 1. UFSAR, Section 9.2, "Shutdown Cooling System" Revision 41 UNITS I CLIFFS - UNITS CALVERT CLIFFS - 1&&22 B 3.9.5-6 B 3.9.5-6 Revision 41
Refueling Pool Water Level B 3.9.6 BASES Structure are within the acceptable limits given in Reference 2.
APPLICABILITY LCO 3.9.6 is applicable when moving irradiated fuel assemblies in the Containment Structure. The LCO minimizes the possibility of a fuel handling accident in the Containment Structure that is beyond the assumptions of the safety analysis. If irradiated fuel is not present in the Containment Structure, there can be no significant radioactivity release as a result of a postulated fuel handling accident. Requirements for fuel handling accidents in the spent fuel pool are covered by LCO 3.7.13.
ACTIONS A.1 With a water level of < 23 ft above the top of the irradiated fuel assemblies seated in the reactor vessel, all operations involving movement of irradiated fuel assemblies, shall be suspended immediately to ensure that a fuel handling accident cannot occur.
The suspension of fuel movement shall not stop the movement of a component to a safe position.
SURVEILLANCE SR 3.9.6.1 REQUIREMENTS Verification of a minimum water level of 23 ft above the top of the irradiated fuel assemblies seated in the reactor vessel ensures that the design basis for the postulated fuel handling accident analysis during refueling operations is met. Water at the required level above the top of the irradiated fuel assemblies seated in the reactor vessel limits the consequences of damaged fuel rods, that are postulated to result from a fuel handling accident inside the Containment Structure (Reference 2).
CALVRT UITS1
- CIFF & B .9.62 Rvisin 4 CALVERT CLIFFS - UNITS I & 2 B 3.9.6-2 Revision 41
ATTACHMENT (5)
COMPARISON MATRIX Calvert Cliffs Nuclear Power Plant, LLC May 1, 2014
ATTACHMENT (5)
COMPARISON MATRIX Technical Specification Title/Surveillance Description TSTF-425 CCNPP Shutdown Margin (SDM) 3.1.1 3.1.1 Verify SDM within limits SR 3.1.1.1 SR 3.1.1.1 Reactivity Balance 3.1.2 CEA Alignment 3.1.4 3.1.4 Verify indicated position within 7 inches SR 3.1.4.1 SR 3.1.4.1 Verify motion inhibit is Operable SR 3.1.4.2 SR 3.1.4.2 Verify deviation circuit is Operable SR 3.1.4.3 SR 3.1.4.3 Verify CEA freedom of movement SR 3.1.4.4 SR 3.1.4.4 Perform Channel Functional Test SR 3.1.4.5 SR 3.1.4.5 Shutdown CEA Insertion Limits 3.1.5 3.1.5 Verify CEA is withdrawn SR 3.1.5.1 SR 3.1.5.1 Regulating CEA Insertion Limits 3.1.6 3.1.6 Verify CEA group position is within limits SR 3.1.6.1 SR 3.1.6.1 Verify CEA insertion times SR 3.1.6.2 SR 3.1.6.2 Verify PDIL alarm circuit is Operable SR 3.1.6.3 SR 3.1.6.3 STE-SDM 3.1.7 3.1.7 Verify CEA insertion is within acceptance criteria SR 3.1.7.1 SR 3.1.7.1 STE-Modes I and 2 3.1.8 3.1.8 Verify Thermal Power is within test power plateau SR 3.1.8.1 SR 3.1.8.1 LHR 3.2.1 3.2.1 Verify ASI alarm setpoints SR 3.2.1.1 SR 3.2.1.2 Verify incore detector local power density alarms SR 3.2.1.2 SR 3.2.1.3 Verify incore local power density alarm setpoints SR 3.2.1 3 SR 3.2.1.4 Fxy 3.2.2 Fr 3.2.3 3.2.3 Verify value of Fr SR 3.2.3.1 SR 3.2.3.1 Tq 3.2.4 3.2.4 Verify value of Tq SR 3.2.4.1 SR 3.2.4.1 ASI 3.2.5 3.2.5 Verify ASI is within limits SR 3.2.5.1 SR 3.2.5.1 RPS Instrumentation -Operating 3.3.1 3.3.1 Perform Channel Check SR 3.3.1.1 SR 3.3.1.1 Perform calibration of excore and dT power channels SR 3.3.1.2 SR 3.3.1.2 Calibrate power range excores using incore detectors SR 3.3.1.3 SR 3.3.1.3 Perform Channel Functional Test SR 3.3.1.4 SR 3.3.1.4 Perform Channel Calibration on excore power range channels SR 3.3.1.5 SR 3.3.1.5 Perform Channel Functional Test on automatic bypass removal SR 3.3.1.7 Perform Channel Calibration of each RPS channel SR 3.3.1.8 SR 3.3.1.8 Verify RPS response times SR 3.3.1.9 SR 3.3.1.9 RPS Instrumentation - Shutdown 3.3.2 3.3.2 Perform Channel Check of wide range power channel SR 3.3.2.1 SR 3.3.2.1 Perform Channel Functional Test of power rate of change trip SR 3.3.2.2 Perform Channel Functional Test of automatic bypass removal SR 3.3.2.3 SR 3.3.2.3 Perform Channel Calibration, including bypass functions SR 3.3.2.4 SR 3.3.2.4 RPS Logic and Trip Initiation 3.3.3 3.3.3 Perform Channel Functional Test on RTCB channel SR 3.3.3.1 SR 3.3.3.1 1
ATTACHMENT (5)
COMPARISON MATRIX Technical Specification Title/Surveillance Description TSTF-425 CCNPP Perform Channel Functional Test on RPS logic SR 3.3.3.2 SR 3.3.3.2 Perform Channel Functional Test with undervoltage and shunt trips SR 3.3.3.4 ESFAS Instrumentation 3.3.4 3.3.4 Perform Channel Check on each ESFAS channel SR 3.3.4.1 SR 3.3.4.1 Perform Channel Functional Test on each ESFAS channel SR 3.3.4.2 SR 3.3.4.2 Perform Channel Functional Test on automatic block removal SR 3.3.4.3 Perform Channel Calibration of ESFAS channels, including block SR 3.3.4.4 SR 3.3.4.4 removal Verify ESF response time in limits SR 3.3.4.5 SR 3.3.4.5 ESFAS Logic and Manual Trip/Actuation 3.3.5 3.3.5 Perform Channel Functional Test on ESFAS Logic channel SR 3.3.5.1 SR 3.3.5.1 Perform Channel Functional Test on ESFAS trip/actuation channel SR 3.3.5.2 SR 3.3.5.2 DG-LOVS 3.3.6 3.3.6 Perform Channel Check SR 3.3.6.1 Perform Channel Functional Test SR 3.3.6.2 SR 3.3.6.1 Perform Channel Calibration SR 3.3.6.3 SR 3.3.6.2 CPISICRS 3.3.7 3.3.7 Perform Channel Check on radiation monitors SR 3.3.7.1 SR 3.3.7.1 Perform Channel Functional Test on radiation monitor channel SR 3.3.7.2 SR 3.3.7.3 Perform Channel Functional Test on actuation logic channel SR 3.3.7.3 SR 3.3.7.2 Perform Channel Calibration on radiation monitor channel SR 3.3.7.4 SR 3.3.7.4 Perform Channel Functional Test on manual trip/actuation channel SR 3.3.7.5 SR 3.3.7.5 Verify response time is within limits SR 3.3.7.6 SR 3.3.7.6 CRIS/CRRS 3.3.8 3.3.8 Perform Channel Check on radiation monitor channel SR 3.3.8.1 SR 3.3.8.1 Perform Channel Functional Test on the radiation monitor channel SR 3.3.8.2 SR 3.3.8.2 Perform Channel Functional Test on actuation logic channel SR 3.3.8.3 Perform Channel Calibration on the radiation monitor channel SR 3.3.8.4 SR 3.3.8.3 Perform Channel Functional Test on the manual trip channel SR 3.3.8.5 Verify response times are within limits SR 3.3.8.6 CVCS Isolation Signal 3.3.9 3.3.9 Perform Channel Check SR 3.3.9.1 SR 3.3.9.1 Perform Channel Functional Test on CVCS isolation channels SR 3.3.9.2 SR 3.3.9.2 Perform Channel Calibration on CVCS sensor channels SR 3.3.9.3 SR 3.3.9.3 Verify response time is within limits SR 3.3.9.4 Shield Building Filtration Actuation Signal 3.3.10 PAM Instrumentation 3.3.11 3.3.10 Perform Channel Check for normally energized channels SR 3.3.11.1 SR 3.3.10.1 Perform Channel Calibration SR 3.3.11.2 SR 3.3.10.3 Remote Shutdown Systemllnstrumentation 3.3.12 3.3.11 Perform Channel Check for normally energized channels SR 3.3.12.1 SR 3.3.11.1 Verify control circuit and transfer switch works SR 3.3.12.2 Perform Channel Calibration for each channel SR 3.3.12.3 SR 3.3.11.2 2
ATTACHMENT (5)
COMPARISON MATRIX Technical Specification Title/Surveillance Description TSTF-425 CCNPP Perform Channel Functional Test SR 3.3.12.4-Power Monitoring Channels/Wide Range Logarithmic Neutron 3.3.13 3.3.12 Flux Monitor Channels Perform Channel Check SR 3.3.13.1 SR 3.3.12.1 Perform Channel Functional Test SR 3.3.13.2-Perform Channel Calibration SR 3.3.13.3 SR 3.3.12.3 RCS Pressure, Temperature and Flow DNBR Limits 3.4.1 3.4.1 Verify pressurizer pressure is within limits SR 3.4.1.1 SR 3.4.1.1 Verify cold leg temperatures are within limits SR 3.4.1.2 SR 3.4.1.2 Verify RCS total flow SR 3.4.1.3 SR 3.4.1.3 Verify heat balance/measured RCS flow is within limits SR 3.4.1.4 SR 3.4.1.4 RCS Minimum Temperature for Criticality 3.4.2 RCS P/T Limits 3.4.3 3.4.3 Verify RCS temperature, pressure, rates are within limits SR 3.4.3.1 SR 3.4.3.1 RCS Loops - Modes 1 and 2 3.4.4 3.4.4 Verify RCS loops in operation SR 3.4.4.1 SR 3.4.4.1 RCS Loops - Mode 3 3.4.5 3.4.5 Verify RCS loop is in operation SR 3.4.5.1 SR 3.4.5.1 Verify secondary side SG water level SR 3.4.5.2 SR 3.4.5.2 Verify correct breaker alignment SR 3.4.5.3 SR 3.4.5.3 RCS Loops - Mode 4 3.4.6 3.4.6 Verify one RCS/SDC in operation SR 3.4.6.1 SR 3.4.6.1 Verify secondary side SG water level SR 3.4.6.2 SR 3.4.6.2 Verify correct breaker alignment SR 3.4.6.3 SR 3.4.6.3 RCS Loops - Mode 5, Loops Filled 3.4.7 3.4.7 Verify SDC train is in operation SR 3.4.7.1 SR 3.4.7.1 Verify secondary side SG water level SR 3.4.7.2 SR 3.4.7.2 Verify correct breaker alignment SR 3.4.7.3 SR 3.4.7.3 RCS Loops - Mode 5, Loops not Filled 3.4.8 3.4.8 Verify SDC train is in operation SR 3.4.8.1 SR 3.4.8.1 Verify correct breaker alignment SR 3.4.8.2 SR 3.4.8.2 Pressurizer 3.4.9 3.4.9 Verify pressurizer water level SR 3.4.9.1 SR 3.4.9.1 Verify capacity of heaters SR 3.4.9.2 SR 3.4.9.2 Verify heaters are powered from emergency power SR 3.4.9.3 Pressurizer PORVs 3.4.11 3.4.11 Perform Channel Functional Test of PORVs SR 3.4.11.1 Perform cycle of block valves SR 3.4.11.1 SR 3.4.11.2 Perform cycle of PORVs SR 3.4.11.2 SR 3.4.11.3 Perform Channel Calibration of PORVs SR 3.4.11.4 Perform cycle of control valve and check valve on PORV SR 3.4.11.3-3
ATTACHMENT (5)
COMPARISON MATRIX Technical Specification Title/Surveillance Description TSTF-425 CCNPP Verify PORVs and block valves are powered from emergency power SR 3.4.11.4 LTOP System 3.4.12 3.4.12 Verify one HPSI capable of RCS injection SR 3.4.12.1 SR 3.4.12.1 Verify one charging pump capable of RCS injection SR 3.4.12.2 Verify SITs are isolated SR 3.4.12.3 Verify HPSI MOVs are capable of manual alignment SR 3.4.12.2 Verify RCS vent is open SR 3.4.12.4 SR 3.4.12.3 Verify PORV block valve is open SR 3.4.12.5 SR 3.4.12.4 Perform Channel Functional Test for PORVs SR 3.4.12.6 SR 3.4.12.5 Perform Channel Calibration on PORVs SR 3.4.12.7 SR 3.4.12.6 RCS Operational Leakage 3.4.13 3.4.13 Verify leakage is within limits SR 3.4.13.1 SR 3.4.13.1 Verify primary to secondary leakage is within limits SR 3.4.13.2 SR 3.4.13.2 RCS PIV Leakage 3.4.14 RCS Leakage Detection Instrumentation 3.4.15 3.4.14 Perform Channel Check of radiation monitor SR 3.4.15.1 SR 3.4.14.1 Perform Channel Functional Test of radiation monitor SR 3.4.15.2 SR 3.4.14.2 Perform Channel Calibration of sump monitor SR 3.4.15.3 SR 3.4.14.3 Perform Channel Calibration of radiation monitor SR 3.4.15.4 SR 3.4.14.4 Perform Channel Calibration of condensate flow monitor SR 3.4.15.5 RCS Specific Activity 3.4.16 3.4.15 Verify gross reactor coolant activity SR 3.4.16.1 SR 3.4.15.1 Verify reactor coolant specific activity SR 3.4.16.2 SR 3.4.15.2 Determine E SR 3.4.16.3 SR 3.4.15.3 STE-RCS Loops, Mode 2 3.4.17 3.4.16 Verify thermal power SR 3.4.17.1 SR 3.4.16.1 STE-RCS Loops, Modes 4 & 5 3.4.17 Verify charging pumps are de-energized SR 3.4.17.2 Verify charging flow path is isolated SR 3.4.17.3 4
ATTACHMENT (5)
COMPARISON MATRIX Technical Specification Title/Surveillance Description TSTF-425 CCNPP Perform SR 3.1.1.1 SR 3.4.17.4 SITs 3.5.1 3.5.1 Verify isolation valves are open SR 3.5.1.1 SR 3.5.1.1 Verify borated water volume in SITs SR 3.5.1.2 SR 3.5.1.2 Verify nitrogen pressure in SITs SR 3.5.1.3 SR 3.5.1.3 Verify boron concentration in SITs SR 3.5.1.4 SR 3.5.1.4 Verify power is removed from SIT isolation valves SR 3.5.1.5 SR 3.5.1.5 ECCS-Operating 3.5.2 3.5.2 Verify specified valve position SR 3.5.2.1 SR 3.5.2.1 Verify valves are in the correct position SR 3.5.2.2 SR 3.5.2.2 Verify ECCS piping is full of water SR 3.5.2.3 Verify valves actuate to correct position SR 3.5.2.6 SR 3.5.2.5 Verify ECCS pumps start on signal SR 3.5.2.7 SR 3.5.2.6 Verify LPSI pump stops on signal SR 3.5.2.8 SR 3.5.2.7 Verify throttle valves stop in correct position SR 3.5.2.9 Verify containment sump is not blocked by debris SR 3.5.2.10 SR 3.5.2.8 Verify SDC interlock works SR 3.5.2.9 RWT 3.5.4 3.5.4 Verify water temperature SR 3.5.4.1 SR 3.5.4.1 Verify water temperature SR 3.5.4.1 SR 3.5.4.2 Verify water volume SR 3.5.4.2 SR 3.5.4.3 Verify boron concentration SR 3.5.4.3 SR 3.5.4.4 TSP/STB 3.5.5 3.5.5 Verify baskets contain TSP/STB SR 3.5.5.1 SR 3.5.5.1 Verify pH adjustment of water SR 3.5.5.2 SR 3.5.5.2 Containment Air Locks 3.6.2 3.6.2 Verify only one door open at a time SR 3.6.2.2 SR 3.6.2.2 Containment Isolation Valves 3.6.3 3.6.3 Verify purge valves are closed SR 3.6.3.1 Verify vent valves are closed SR 3.6.3.2 SR 3.6.3.1 Verify isolation valves are closed SR 3.6.3.3 SR 3.6.3.2 Verify isolation time of valves SR 3.6.3.5 Perform leak rate test of purge valves SR 3.6.3.6 Verify valves actuate to correct position SR 3.6.3.7 SR 3.6.3.5 Verify purge valve is blocked SR 3.6.3.8 Containment Pressure 3.6.4 3.6.4 Verify pressure is within limits SR 3.6.4.1 SR 3.6.4.1 Containment Air Temperature 3.6.5 3.6.5 Verify temperature is within limits SR 3.6.5.1 SR 3.6.5.1 Containment Spray and Cooling System 3.6.6A 3.6.6 Verify containment spray valves are in the correct position SR 3.6.6A.1 SR 3.6.6.1 Operate cooling train fan SR 3.6.6A.2 SR 3.6.6.2 Verify cooling water flow rate SR 3.6.6A.3 SR 3.6.6.3 Verify containment spray pipe is full of water SR 3.6.6A.4-Verify valves actuate to their correct position SR 3.6.6A.6 SR 3.6.6.5 5
ATTACHMENT (5)
COMPARISON MATRIX Technical Specification Title/Surveillance Description TSTF-425 CCNPP Verify containment spray pump starts SR 3.6.6A.7 SR 3.6.6.6 Verify cooling train starts SR 3.6.6A.8 SR 3.6.6.7 Verify spray nozzle is unobstructed SR 3.6.6A.9-Spray Additive System 3.6.7 Shield Building Exhaust Air Cleanup System 3.6.8 HMS 3.6.9 Iodine Cleanup System/Iodine Removal System 3.6.10 3.6.8 Operate each train SR 3.6.10.1 SR 3.6.8.1 Verify each train actuates SR 3.6.10.3 SR 3.6.8.3 Verify bypass damper can be opened SR 3.6.10.4 Shield Building 3.6.11 MSIVs 3.7.2 3.7.2 MFIVs 3.7.3 3.7.15 ADVs 3.7.4 AFW System 3.7.5 3.7.3 Verify valves are in the correct position SR 3.7.5.1 SR 3.7.3.1 Verify valves actuate to the correct position SR 3.7.5.3 SR 3.7.3.4 Verify the AFW pumps start automatically SR 3.7.5.4 SR 3.7.3.5 Verify the AFW system can provide minimum flow SR 3.7.3.6 Condensate Storage Tank 3.7.6 3.7.4 Verify CST level/volume SR 3.7.6.1 SR 3.7.4.1 CCWlCC System 3.7.7 3.7.5 Verify valves are in the correct position SR 3.7.7.1 SR 3.7.5.1 Verify valves actuate to the correct position SR 3.7.7.2 SR 3.7.5.2 Verify pumps start automatically SR 3.7.7.3 SR 3.7.5.3 SWS/SRW 3.7.8 3.7.6 Verify valves are in the correct position SR 3.7.8.1 SR 3.7.6.1 Verify valves actuate to the correct position SR 3.7.8.2 SR 3.7.6.2 Verify pumps start automatically SR 3.7.8.3 SR 3.7.6.3 SWS/SW 3.7.8 3.7.7 Verify valves are in the correct position SR 3.7.8.1 SR 3.7.7.1 Verify valves actuate to the correct position SR 3.7.8.2 SR 3.7.7.2 Verify pumps start automatically SR 3.7.8.3 SR 3.7.7.3 UHS 3.7.9 ECW 3.7.10 CREACS/CREVS 3.7.11 3.7.8 Operate filter train SR 3.7.11.1 SR 3.7.8.1 Verify each train actuates on a signal SR 3.7.11.3 SR 3.7.8.3 Verify positive pressure can be maintained SR 3.7.11.4-CREATS/CRETS 3.7.12 3.7.9 Verify heat load can be removed/temperature can be maintained SR 3.7.12.1 SR 3.7.9.1 ECCS PREACS 3.7.13 FBACS/SFPEVS 3.7.14 3.7.11 Operate each train SR 3.7.14.1 Verify a train is in operation SR 3.7.11.1 6
ATTACHMENT (5)
COMPARISON MATRIX Technical Specification Title/Surveillance Description TSTF-425 CCNPP Verify each train actuates on a signal SR 3.7.14.3 Verify a train can maintain a negative pressure SR 3.7.14.4 SR 3.7.11.3 Verify the bypass damper can be opened SR 3.7.14.5 PREACS/PREVS 3.7.15 3.7.12 Operate each train SR 3.7.15.1 SR 3.7.12.1 Verify each train actuates on a signal SR 3.7.15.3 SR 3.7.12.3 Verify a train can maintain a negative pressure SR 3.7.15.4 Verify the bypass damper can be opened SR 3.7.15.5 Fuel Storage Pool Water Level/ SFP Water Level 3.7.16 3.7.13 Verify water level in pool SR 3.7.16.1 SR 3.7.13.1 FSP Boron Concentration/SFP Boron Concentration 3.7.17 3.7.16 Verify boron concentration in pool SR 3.7.17.1 SR 3.7.16.1 Secondary Specific Activity 3.7.19 3.7.14 Verify secondary specific activity is within limits SR 3.7.19.1 SR 3.7.14.1 AC Sources-Operating 3.8.1 3.8.1 Verify correct breaker alignment SR 3.8.1.1 SR 3.8.1.1 Verify correct breaker alignment SR 3.8.1.1 SR 3.8.1.2 Verify DG standby start SR 3.8.1.2 SR 3.8.1.3 Verify DG is loaded and operate for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> SR 3.8.1.3 SR 3.8.1.4 Verify day tank volume SR 3.8.1.4 SR 3.8.1.5 Check for water in the day tank SR 3.8.1.5 SR 3.8.1.6 Verify fuel oil transfer system operates SR 3.8.1.6 SR 3.8.1.7 Verify DG standby start (fast start) SR 3.8.1.7 SR 3.8.1.9 Verify manual transfer of AC sources SR 3.8.1.8 SR 3.8.1.10 Verify single load reject SR 3.8.1.9 SR 3.8.1.12 Verify full load reject SR 3.8.1.10 Verify DG start and operation on a loss of offsite power SR 3.8.1.11 Verify DG start and operation on an ESF signal SR 3.8.1.12 Verify non-critical trips are bypassed SR 3.8.1.13 SR 3.8.1.13 Verify DG operation for 24/1 hour SR 3.8.1.14 SR 3.8.1.11 Verify hot restart of the DG SR 3.8.1.15 Verify DG synchronizes to offsite power SR 3.8.1.16 SR 3.8.1.14 Verify ESF signal overrides DG test mode SR 3.8.1.17 Verify load sequencer operation SR 3.8.1.18 SR 3.8.1.8 Verify DG start and operation on an ESF/LOOP signal SR 3.8.1.19 SR 3.8.1.15 7
ATTACHMENT (5)
COMPARISON MATRIX Technical Specification TitlelSurveillance Description TSTF-425 CCNPP Verify DG operation during simultaneous start SR 3.8.1.20-Diesel Fuel Oil, Lube Oil and Starting Air/Diesel Fuel Oil 3.8.3 3.8.3 Verify volume of diesel fuel oil SR 3.8.3.1 SR 3.8.3.1 Verify lube oil inventory SR 3.8.3.2 Verify air start receiver pressure SR 3.8.3.4 Remove accumulated water from storage tank SR 3.8.3.5 SR 3.8.3.3 DC Sources - Operating 3.8.4 3.8.4 Verify battery terminal voltage SR 3.8.4.1 SR 3.8.4.1 Verify battery charger provides adequate voltage SR 3.8.4.2 SR 3.8.4.6 Verify battery capacity SR 3.8.4.3 SR 3.8.4.7 Verify no visible corrosion SR 3.8.4.2 Verify no physical damage SR 3.8.4.3 Remove visible corrosion SR 3.8.4.4 Verify battery connection resistance SR 3.8.4.5 Battery Parameters/Battery Cell Parameters 3.8.6 3.8.6 Verify float current/voltage SR 3.8.6.1 SR 3.8.6.1 Verify battery cell voltage SR 3.8.6.2 SR 3.8.6.2 Verify electrolyte level SR 3.8.6.3 SR 3.8.6.1 Verify battery temperature SR 3.8.6.4 SR 3.8.6.3 Verify cell voltage SR 3.8.6.5 SR 3.8.6.2 Verify battery capacity - discharge test SR 3.8.6.6 SR 3.8.4.8 Inverters-Operating 3.8.7 3.8.7 Verify correct voltage SR 3.8.7.1 SR 3.8.7.1 Inverters-Shutdown 3.8.8 3.8.8 Verify correct voltage SR 3.8.8.1 SR 3.8.8.1 Distribution Systems-Operating 3.8.9 3.8.9 Verify correct breaker alignments SR 3.8.9.1 SR 3.8.9.1 Distribution Systems-Shutdown 3.8.10 3.8.10 Verify correct breaker alignments SR 3.8.10.1 SR 3.8.10.1 Boron Concentration 3.9.1 3.9.1 Verify boron concentration SR 3.9.1.1 SR 3.9.1.1 Nuclear Instrumentation 3.9.2 3.9.2 Perform Channel Check SR 3.9.2.1 SR 3.9.2.1 Perform Channel Calibration SR 3.9.2.2 SR 3.9.2.2 Containment Penetrations 3.9.3 3.9.3 Verify containment penetrations are in correct status SR 3.9.3.1 SR 3.9.3.1 Verify containment purge actuates on a signal SR 3.9.3.2 SR 3.9.3.2 SDC and Coolant Circulation-High Water Level 3.9.4 3.9.4 Verify one SDC loop is in operation SR 3.9.4.1 SR 3.9.4.1 SDC and Coolant Circulation-Low Water Level 3.9.5 3.9.5 Verify SDC loops are operable SR 3.9.5.1 SR 3.9.5.1 Verify correct breaker alignment SR 3.9.5.2 SR 3.9.5.3 Verify flow rate of SDC loop SR 3.9.5.2 Refueling Water Level/Refueling Pool Water Level 3.9.6 3.9.6 Verify water level SR 3.9.6.1 SR 3.9.6.1 8